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Sample records for decay-heat removal analyses

  1. CRBRP decay heat removal systems

    International Nuclear Information System (INIS)

    Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

    1977-01-01

    The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

  2. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  3. Advances in technologies for decay heat removal

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Berkovich, V.; Bianchi, A.; Chen B.; Meseth, J.; Vecchiarelli, J.; Vidard, M.

    1999-01-01

    The various decay heat removal concepts that have been used for the evolutionary water reactor plant designs developed worldwide are examined and common features identified. Although interesting new features of the 'classical' plants are mentioned, the emphasis is on passive core and containment decay heat removal systems. The various systems are classified according to the function they have to accomplish; they often share common characteristics and similar equipment. (author)

  4. Study on diverse passive decay heat removal approach and principle

    International Nuclear Information System (INIS)

    Lin Qian; Si Shengyi

    2012-01-01

    Decay heat removal in post-accident is one of the most important aspects concerned in the reactor safety analysis. Passive decay heat removal approach is used to enhance nuclear safety. In advanced reactors, decay heat is removed by multiple passive heat removal paths through core to ultimate heat sink by passive residual heat removal system, passive injection system, passive containment cooling system and so on. Various passive decay heat removal approaches are summarized in this paper, the common features and differences of their heat removal paths are analyzed, and the design principle of passive systems for decay heat removal is discussed. It is found that. these decay heat removal paths is combined by some basic heat transfer processes, by the combination of these basic processes, diverse passive decay heat removal approach or system design scheme can be drawn. (authors)

  5. Decay heat removal for the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Zemanick, P.P.; Brown, N.W.

    1975-01-01

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. A statement of the high reliability of the Clinch River Breeder reactor Plant decay heat removal systems and a summary of the supporting arguments is presented. (U.S.)

  6. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zemanick, P. P.; Brown, N. W.

    1975-10-15

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  7. Study on diverse passive decay heat removal approach

    International Nuclear Information System (INIS)

    Lin Qian; Si Shengyi

    2012-01-01

    One of the most important principles for nuclear safety is the decay heat removal in accidents. Passive decay heat removal systems are extremely helpful to enhance the safety. In currently design of many advanced nuclear reactors, kinds of passive systems are proposed or developed, such as the passive residual heat removal system, passive injection system, passive containment cooling system. These systems provide entire passive heat removal paths from core to ultimate heat sink. Various kinds of passive systems for decay heat removal are summarized; their common features or differences on heat removal paths and design principle are analyzed. It is found that, these passive decay heat removal paths are similarly common on and connected by several basic heat transfer modes and steps. By the combinations or connections of basic modes and steps, new passive decay heat removal approach or diverse system can be proposed. (authors)

  8. AEA studies on passive decay heat removal in advanced reactors

    International Nuclear Information System (INIS)

    Lillington, J.N.

    1994-01-01

    The main objectives of the UK study were: to identify, describe and compare different types of systems proposed in current designs; to identify key scenarios in which passive decay heat removal systems play an important preventative or mitigative role; to assess the adequacy of the relevant experimental database; to assess the applicability and suitability of current generation models/codes for predicting passive decay heat removal; to assess the potential effectiveness of different systems in respect of certain key licensing questions

  9. Method for removal of decay heat of radioactive substances

    International Nuclear Information System (INIS)

    Hesky, H.; Wunderer, A.

    1981-01-01

    In this process, the decay heat from radioactive substances is removed by means of a liquid carried in the coolant loop. The liquid is partially evaporated by the decay heat. The steam is used to drive the liquid through the loop. When a static pressure level equivalent to the pressure drop in the loop is exceeded, the steam is separated from the liquid, condensed, and the condensate is reunited with the return flow of liquid for partial evaporation. (orig.) [de

  10. Tests for removal of decay heat by natural convection

    International Nuclear Information System (INIS)

    Kashiwagi, E.; Wataru, M.; Gomi, Y.; Hattori, Y.; Ozaki, S.

    1993-01-01

    Interim storage technology for spent fuel by dry storage casks have been investigated. The casks are vertically placed in a storage building. The decay heat is removed from the outer cask surface by natural convection of air entering from the building wall to the roof. The air flow pattern in the storage building was governed by the natural driving pressure difference and circulating flow. The purpose of this study is to understand the mechanism of the removal of decay heat from casks by natural convection. The simulated flow conditions in the building were assumed as a natural and forced combined convection and were investigated by the turbulent quantities near wall. (author)

  11. Summary report of RAMONA investigations into passive decay heat removal

    International Nuclear Information System (INIS)

    Hoffmann, H.; Marten, K.; Weinberg, D.; Frey, H.H.; Rust, K.; Ieda, Y.; Kamide, H.; Ohshima, H.; Ohira, H.

    1995-07-01

    An important safety feature of an advanced sodium-cooled reactor (e.g. European Fast Reactor, EFR) is the passive decay heat removal. This passive concept is based on several direct reactor cooling systems operating independently from each other. Each of the systems consists of a sodium/sodium decay heat exchanger immersed in the primary vessel and connected via an intermediate sodium loop to a heat sink formed by a sodium/air heat exchanger installed in a stack with air inlet and outlet dampers. The decay heat is removed by natural convection on the sodium side and natural draft on the air side. To demonstrate the coolability of the pool-type primary system by buoyancy-driven natural circulation, tests were performed under steady-state and transient conditions in facilities of different scale and detail. All these investigations serve to understand the physical processes and to verify computer codes used to transfer the results to reactor conditions. RAMONA is the three-dimensional 1:20-scaled apparatus equipped with all active components. Water is used as simulant fluid for sodium. The maximum core power is 75 kW. The facility is equipped with about 250 thermocouples to register fluid temperatures. Velocities and mass flows are measured by Laser Doppler Anemometers and magneto-inductive flowmeters. Flow paths are visualized by tracers. The conclusion of the investigations is that the decay heat can be removed from the primary system by means of natural convection. Always flow paths develop, which ensure an effective cooling of all regions. This is even proved for extreme conditions, e.g. in case of delays of the decay heat exchanger startup, failures of several DHR chains, and a drop of the fluid level below the inlet windows of the IHXs and decay heat exchangers. (orig.) [de

  12. A decay heat removal methodology for reuseable orbital transfer vehicles

    Science.gov (United States)

    McDaniel, Patrick J.; Perkins, David R.

    1992-07-01

    Operation of a nuclear thermal rocket(NTR) as the propulsion system for a reusable orbital transfer vehicle has been considered. This application is the most demanding in terms of designing a multiple restart capability for an NTR. The requirements on a NTR cooling system associated with the nuclear decay heat stored during operation have been evaluated, specifically for a Particle Bed Reactor(PBR) configuration. A three mode method of operation has been identified as required to adequately remove the nuclear decay heat.

  13. Control of the ASTRA decay heat removal system

    International Nuclear Information System (INIS)

    Nedelik, A.

    1982-11-01

    To ensure a minimum of core cooling even under severest accident conditions (loss of reactor pool water) a core spray system for decay heat removal has been installed at the ASTRA-reactor. The automatic and manual control of the system, its power supply and test procedures are shortly described. (Author)

  14. Passive decay heat removal by natural circulation

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Venkat Raj, V.; Kakodkar, A.; Mehta, S.K.

    1990-01-01

    The standardised 235 MWe PHWRs being built in India are the pressure tube type, heavy water moderated, heavy water cooled and natural uranium fuelled reactors. Several passive safety features are incorporated in these reactors. These include: (1) Containment pressure reduction and fission product trapping with the help of suppression pool following LOCA. (2) Emergency coolant injection by means of accumulators. (3) Large heat sink provided by the low temperature moderator under accident conditions. (4) Low excess reactivity, through the use of natural uranium fuel and on power fuelling. (5) Residual heat removal by means of natural circulation, etc. of which the last item is the subject matter of this report. (author). 8 refs, 10 figs

  15. Passive decay heat removal from the core region

    International Nuclear Information System (INIS)

    Hichen, E.F.; Jaegers, H.

    2002-01-01

    The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

  16. Analysis of decay heat removal by natural convection in PFBR

    International Nuclear Information System (INIS)

    Kasinathan, N.; Vaidyanathan, G.; Chetal, S.C.; Bhoje, S.B.

    1993-01-01

    PFBR is a 500 MWe, 1200 MWt pool type LMFBR. In order to assure reliable decay heat removal, four totally independent Safety Grade Decay Heat Removal Systems (SGDHRS) which removes heat directly from the hot pool, is provided. Each of the SGDHRS comprises of a hot pool dipped decay heat exchanger (DHX), a sodium - air heat exchanger (AHX) at a suitable elevation and associated piping and circuits. This paper brings out the step by step approach that have been taken to decide on the preliminary sizing of the SGDHRS components, and static and transient analysis to assess the adequacy of the Decay Heat Removal capacity of the SGDHRS during the worst of the foreseen design basis conditions. The maximum values the important safety related temperatures viz., clad hotspot, hot pool top surface, reactor inlet, fuel subassembly outlets etc., would reach, can be obtained only through a comprehensive transient analysis. In order to get quick and reasonably meaningful results, one dimensional thermal-hydraulics models for the core, hot and cold pools, IHX, DHX, AHX and various pipings were developed and a code DHDYN formulated. With this a total power failure situation followed by initiations of DHR half an hour later was studied and the results revealed the following: (i) clad hotspot temperature in the in-vessel stored spent fuel subassemblies could be held below 800 deg. C only if primary sodium flow through these subassemblies are increased up to three times the originally allocated flow in the design, (ii) hotpool top zone temperature reaches 572 deg. C, (iii) reactor inlet temperature reaches 482 deg. C, (iv) the hot pool top zone temperature cools down to 450 deg. C in about 25 h. Thus these results satisfactorily established the adequacy of the sizing and the capability of the SGDHRS. DHDYN code is also used to study the RAMONA water experiments conducted in Germany. Initial results available has brought out the conservative nature of the DHDYN predictions as compared

  17. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  18. A decay heat removal system requiring no external energy

    International Nuclear Information System (INIS)

    Costes, D.; Fermandjian, J.

    1983-12-01

    A new Decay heat Removal System is described for PWR's with dry containment, i.e. a containment building which encloses no permanent reserve of cooling water. This new system is intended to provide a high level of safety since it uses no external energy, but only the thermodynamic energy of the air-steam-liquid water mixture generated in the containment after the failure of the primary circuit (''LOCA'') or of the secondary circuit. Thermodynamics of the system is evaluated first: after some design considerations, the use of the system for protecting actual PWR's is addressed

  19. Passive safety systems for decay heat removal of MRX

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

    1996-12-01

    The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

  20. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  1. Analysis of decay heat removal following loss of RHR

    International Nuclear Information System (INIS)

    Naff, S.A.; Ward, L.W.

    1991-01-01

    Recent plant experience has included many events occurring during outages at pressurized water reactors. A recent example is the loss of residual heat removal system event that occurred March 20, 1990 at the Vogtle-1 plant following refueling. Plant conditions during outages differ markedly from those prevailing at normal full-power operation on which most past research has concentrated. Specifically, during outages the core power is low, the coolant system may be in a drained state with air or nitrogen present, and various reactor coolant system closures may be unsecured. With the residual heat removal system operating, the core decay heat is readily removed. However, if the residual heat removal system capability is lost and alternative heat removal means cannot be established, heat up of the coolant could lead to core coolant boil-off, fuel rod heat up, and core damage. A study was undertaken by the Nuclear Regulatory Commission to identify what information was needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that might be used, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain into the reactor coolant system, core water boil-off, and reflux condensation cooling processes

  2. PANDA passive decay heat removal transient test results

    International Nuclear Information System (INIS)

    Bandurski, Th.; Dreier, J.; Huggenberger, M.

    1997-01-01

    PANDA is a large scale facility for investigating the long-term decay heat removal from the containment of a next generation of 'passive' Advanced Light Water Reactors (ALWR). PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The first PANDA test series had the dual objectives of demonstrating the performance of the SBWR PCCS and extending the data base available for containment analysis code qualification. The test objectives also include the study of the effects of mixing and stratification of steam and noncondensible gases in the drywell (DW) and in the suppression chamber or wetwell (WW). Ten tests were conducted in the course of the PANDA SBWR Program. The tests demonstrated a favorable and robust overall PCCS performance under different conditions. The present paper focuses on the main phenomena observed during the tests with respect to PCCS operation and DW gas mixing. (author)

  3. Passive decay heat removal by sump cooling after core meltdown

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1996-01-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is 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. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

  4. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  5. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  6. Decay heat removal and transient analysis in accidental conditions in the EFIT reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Meloni, P.; Polidori, M.; Casamirra, M.; Castiglia, F.; Giardina, M.

    2007-01-01

    The development of a conceptual design of an industrial scale transmutation facility (EFIT) of several 100 MW thermal power based on Accelerator Driven System (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which lead to the Loss of Heat Sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1-D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios. (author)

  7. Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

    Directory of Open Access Journals (Sweden)

    Giacomino Bandini

    2008-01-01

    Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

  8. Possible design of PBR for passive decay heat removal

    International Nuclear Information System (INIS)

    Sambuu, Odmaa; Obara, Toru

    2016-01-01

    Conditions for design parameters of above-ground and underground, prismatic high-temperature gas-cooled reactor (HTGR)s for passive decay heat removal based on fundamental heat transfer mechanisms were obtained in the previous works. In the present study, analogous conditions were obtained for pebble bed reactors by performing the same procedure using the model for heat transfer in porous media of COMSOL 4.3a software, and the results were compared. For the power density profile, several approximated distributions together with original one throughout the 10-MWt high-temperature gas-cooled reactor-test module (HTR-10) were used, and it was found that an HTR-10 with a uniform power density profile has the higher safety margin than those with other profiles. In other words, the safety features of a PBR can be enhanced by flattening the power density profile. We also found that a prismatic HTGR with a uniform power density profile throughout the core has a greater safety margin than a PBR with the same design characteristics. However, when the power density profile is not flattened during the operation, the PBR with the linear power density profile has more safety margin than the prismatic HTGR with the same design parameters and with the power density profile by cosine and Bessel functions. (author)

  9. Performance of ALMR passive decay heat removal system

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hunsbedt, A.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the small (471 MWt) modular reactor to the environmental air by natural convection heat transfer. The system has no active components, requires no operator action to initiate, and is inherently reliable. The RVACS can perform its function under off-normal or degraded operating conditions without significant loss in performance. Several such events are described and the RVACS thermal performance for each is given and compared to the normal operation performance. The basic RVACS performance as well as the performance during several off-normal events have been updated to reflect design changes for recycled fuel with minor actinides for end of equilibrium cycle conditions. The performance results for several other off-normal events involving various degrees of RVACS air flow passage blockages are presented. The results demonstrated that the RVACS is unusually tolerant to a wide range of postulated faults. (author)

  10. Modelling of decay heat removal using large water pools

    International Nuclear Information System (INIS)

    Munther, R.; Raussi, P.; Kalli, H.

    1992-01-01

    The main task for investigating of passive safety systems typical for ALWRs (Advanced Light Water Reactors) has been reviewing decay heat removal systems. The reference system for calculations has been represented in Hitachi's SBWR-concept. The calculations for energy transfer to the suppression pool were made using two different fluid mechanics codes, namely FIDAP and PHOENICS. FIDAP is based on finite element methodology and PHOENICS uses finite differences. The reason choosing these codes has been to compare their modelling and calculating abilities. The thermal stratification behaviour and the natural circulation was modelled with several turbulent flow models. Also, energy transport to the suppression pool was calculated for laminar flow conditions. These calculations required a large amount of computer resources and so the CRAY-supercomputer of the state computing centre was used. The results of the calculations indicated that the capabilities of these codes for modelling the turbulent flow regime are limited. Output from these codes should be considered carefully, and whenever possible, experimentally determined parameters should be used as input to enhance the code reliability. (orig.). (31 refs., 21 figs., 3 tabs.)

  11. A study on the characteristics of the decay heat removal capacity for a large thermal rated LMR design

    International Nuclear Information System (INIS)

    Uh, J. H.; Kim, E. K.; Kim, S. O.

    2003-01-01

    The design characteristics and the decay heat removal capacity according to the type of DHR (Decay Heat Removal) system in LMR are quantitatively analyzed, and the general relationship between the rated core thermal power and decay heat removal capacity is created in this study. Based on these analyses results, a feasibility of designing a larger thermal rating KALIMER plant is investigated in view of decay heat removal capacity, and DRC (Direct Reactor Cooling) type DHR system which rejects heat from the reactor pool to air is proper to satisfy the decay heat removal capacity for a large thermal rating plant above 1,000 MWth. Some defects, however, including the heat loss under normal plant operation and the lack of reliance associated with system operation should be resolved in order to adopt the total passive concept. Therefore, the new concept of DHR system for a larger thermal rating KALIMER design, named as PDRC (passive decay heat removal circuit), is established in this study. In the newly established concept of PDRC, the Na-Na heat exchanger is located above the sodium cold pool and is prevented from the direct sodium contact during normal operation. This total passive feature has the superiority in the aspect of the minimizing the normal heat loss and the increasing the operation reliance of DHR system by removing either any operator action or any external operation signal associated with system operation. From this study, it is confirmed that the new concept of PDRC is useful to the designing of a large thermal rating power plant of KALIMER-600 in view of decay heat removal capability

  12. Study on decay heat removal capability of reactor vessel auxiliary cooling system

    International Nuclear Information System (INIS)

    Nishi, Y.; Kinoshita, I.

    1991-01-01

    The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

  13. Shutdown decay heat removal analysis: Plant case studies and special issues: Summary report

    International Nuclear Information System (INIS)

    Ericson, D.M. Jr.; Cramond, W.R.; Sanders, G.A.; Hatch, S.W.

    1989-04-01

    Shutdown Decay Heat Removal Requirements has been designated as Unresolved Safety Issue (USI) A-45. The overall objectives of the USI A-45 program were to evaluate the safety adequacy of decay heat removal (DHR) systems in existing light water reactor nuclear power plants and to assess the value and impact (benefit-cost) of alternative measures for improving the overall reliability of the DHR function. To provide the technical data required to meet these objectives a program was developed that examined the state of DHR system reliability in a sample of existing plants. This program identified potential vulnerabilities and identified and established the feasibility of potential measures to improve the reliability of the DHR function. A value/impact (V/I) analysis of the more promising of such measures was conducted and documented. This report summarizes those studies. In addition, because of the evolving nature of V/I analyses in support of regulation, a number of supporting studies related to appropriate procedures and measures for the V/I analyses were also conducted. These studies are also summarized herein. This report only summarizes findings of technical studies performed by Sandia National Laboratories as part of the program to resolve this issue. 46 refs., 7 figs., 124 tabs

  14. Application of the PSA method to decay heat removal systems in a large scale FBR design

    International Nuclear Information System (INIS)

    Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

    1993-01-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  15. Application of the PSA method to decay heat removal systems in a large scale FBR design

    Energy Technology Data Exchange (ETDEWEB)

    Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

    1993-02-01

    The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

  16. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

    2016-01-01

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  17. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  18. A PRA case study of extended long term decay heat removal for shutdown risk assessment

    International Nuclear Information System (INIS)

    Roglans, J.; Ragland, W.A.; Hill, D.J.

    1992-01-01

    A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL). The results of this PRA have shown that the decay heat removal system for EBR-II is extremely robust and reliable. In addition, the methodology used demonstrates how the actions of other systems not normally used for actions of other systems not normally used for decay heat removal can be used to expand the mission time of the decay heat removal system and further increase its reliability. The methodology may also be extended to account for the impact of non-safety systems in enhancing the reliability of other dedicated safety systems

  19. Strategy of experimental studies in PNC on natural convection decay heat removal

    International Nuclear Information System (INIS)

    Ieda, Y.; Kamide, H.; Ohshima, H.; Sugawara, S.; Ninokata, H.

    1993-01-01

    Experimental studies have been and are being carried out in PNC to establish the design and safety evaluation methods and the design and safety evaluation guide lines for decay heat removal by natural convection. A strategy of the experimental studies in PNC is described in this paper. The sphere of studies in PNC is to develop the evaluation methods to be available to DRACS as well as PRACS and IRACS for the plant where decay heat is removed by natural convection in some cases of loss of station service power. Similarity parameters related to natural convection are derived from the governing equations. The roles of both sodium and water experiments are defined in consideration of the importance of the similarity parameters and characteristics of scale model experiments. The experimental studies in PNC are reviewed. On the basis of the experimental results, recommended evaluation methods are shown for decay heat removal feature by natural convection. Future experimental works are also proposed. (author)

  20. Transient testing of the FFTF for decay-heat removal by natural convection

    International Nuclear Information System (INIS)

    Beaver, T.R.; Johnson, H.G.; Stover, R.L.

    1982-06-01

    This paper reports on the series of transient tests performed in the FFTF as a major part of the pre-operations testing program. The structure of the transient test program was designed to verify the capability of the FFTF to safely remove decay heat by natural convection. The series culminated in a scram from full power to complete natural convection in the plant, simulating a loss of all electrical power. Test results and acceptance criteria related to the verification of safe decay heat removal are presented

  1. Probabilistic analysis of the loss of the decay heat removal function for Creys-Malville reactor

    International Nuclear Information System (INIS)

    Lanore, J.M.; Villeroux-Lombard, C.; Bouscatie, F.; Pavret de la Rochefordiere, A.

    1982-01-01

    The classical fault tree/event tree methods do not take into account the dependence in time of the systems behaviour during the sequences, and that is quite unrealistic for the decay heat removal function. It was then necessary to use a new methodology based on functional states of the whole system and on transition laws between these states. Thus, the probabilistic analysis of the decay heat removal function for Creys-Malville plant is performed in a global way. The main accident sequences leading to the loss of the function are then determined a posteriori. The weak points are pointed out, in particular the importance of common mode failures

  2. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  3. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu

    2016-01-01

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely

  4. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  5. Investigation on natural convection decay heat removal for the EFR status of the program

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, F [Kernforschungszentrum Karlsruhe (Germany); Essig, C [Siemens AG, Bergisch Gladbach (Germany); Georgeoura, S [AEA Reactor Service, Dounreay (United Kingdom); Tenchine, D [CEA Grenoble (France)

    1993-02-01

    The European Research and Development (R+D) Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes within the primary system and the direct reactor cooling circuits and include reactor experiments. (author)

  6. Investigation on natural convection decay heat removal for the EFR: Status of the program

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H; Weinberg, D [Kernforschungszentrum Karlsruhe GmbH, IATF, Karlsruhe (Germany); Webster, R [AEA Reactor Services, Dounreay (United Kingdom)

    1991-07-01

    The European Research and Development Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes withinthe primary system and the direct reactor cooling circuits and include fundamental tests as well as reactor experiments. (author)

  7. Investigation on natural convection decay heat removal for the EFR status of the program

    International Nuclear Information System (INIS)

    Hofmann, F.; Essig, C; Georgeoura, S.; Tenchine, D.

    1993-01-01

    The European Research and Development (R+D) Program on decay heat removal by natural convection for the European Fast Reactor (EFR) covers the calculational methods and the model experiments performed for code validation. The studies concentrate on important physical effects of the cooling modes within the primary system and the direct reactor cooling circuits and include reactor experiments. (author)

  8. Application of optimal estimation techniques to FFTF decay heat removal analysis

    International Nuclear Information System (INIS)

    Nutt, W.T.; Additon, S.L.; Parziale, E.A.

    1979-01-01

    The verification and adjustment of plant models for decay heat removal analysis using a mix of engineering judgment and formal techniques from control theory are discussed. The formal techniques facilitate dealing with typical test data which are noisy, redundant and do not measure all of the plant model state variables directly. Two pretest examples are presented. 5 refs

  9. The status of thermal-hydraulic studies on the decay heat removal by natural convection using RAMONA and NEPTUN models

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hain, K.; Marten, K.; Rust, K.; Weinberg, D.; Ohira, H.

    2004-01-01

    Thermal-hydraulic experiments were performed with water in order to simulate the decay heat removal by natural convection in a pool-type sodium-cooled reactor. Two test rigs of different scales were used, namely RAMONA (1:20) and NEPTUN (1:5). RAMONA served to study the transition from nominal operation by forced convection to decay heat removal operation by natural convection. Steady-state similarity tests were carried out in both facilities. The investigations cover nominal and non-nominal operation conditions. These data provide a broad basis for the verification of computer programs. Numerical analyses performed with the three-dimensional FLUTAN code indicated that the thermal-hydraulic processes can be quantitatively simulated even for the very complex geometry of the NEPTUN test rig. (author)

  10. Experimental and analytical studies for the validation of HTR-VGD and primary cell passive decay heat removal. Supplement. Calculations

    International Nuclear Information System (INIS)

    Geiss, M.; Giannikos, A.; Hejzlar, P.; Kneer, A.

    1993-04-01

    The alternative concept for a modular HTR-reactor design by Siempelkamp, Krefeld, using a prestressed cast iron vessel (VGD) combined with a cast iron/concrete module for the primary cell with integrated passive decay heat removal system was fully qualified with respect to operational and accidental thermal loads. The main emphasis was to confirm and validate the passive decay heat removal capability. An experimental facility (INWA) was designed, instrumented and operated with an appropriate electrical heating system simulating steady-state operational and transient accidental thermal loads. The experiments were accompanied by extensive computations concerning the combination of conductive, radiative and convective energy transport mechanisms in the different components of the VGD/primary cell structures, as well as elastic-plastic stress analyses of the VGD. In addition, a spectrum of potential alternatives for passive energy removed options have been parametrically examined. The experimental data clearly demonstrate that the proposed Siempelkamp-design is able to passively and safely remove the decay heat for operational and accidental conditions without invalidating technological important thermal limits. This also holds in case of failures of both the natural convection system and ultimate heat sink by outside concrete water film cooling. (orig./HP) [de

  11. A passive decay-heat removal system for an ABWR based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2017-01-15

    Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

  12. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

    2014-01-01

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  13. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  14. Development of a new decay heat removal system for a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sim, Yoon Sub; Park, Rae Young; Kim, Seyun

    2007-01-01

    The heat removal capacity of a RCCS is one of the major parameters limiting the capacity of a HTGR based on a passive safety system. To improve the plant economy of a HTGR, the decay heat removal capacity needs to be improved. For this, a new analysis system of an algebraic method for the performance of various RCCS designs was set up and the heat transfer characteristics and performance of the designs were analyzed. Based on the analysis results, a new passive decay heat removal system with a substantially improved performance, LFDRS was developed. With the new system, one can have an expectation that the heat removal capacity of a HTGR could be doubled

  15. Experience with after-shutdown decay heat removal - BWRs and PWRs

    International Nuclear Information System (INIS)

    Haugh, J.J.; Mollerus, F.J.; Booth, H.R.

    1992-01-01

    Boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) make use of residual heat removal systems (RHRSs) during reactor shutdown. RHRS operational events involving an actual loss or significant degradation of an RHRS during shutdown heat removal are often prompted or aggravated by complex, changing plant conditions and by concurrent maintenance operations. Events involving loss of coolant inventory, loss of decay heat removal capability, or inadvertent pressurization while in cold shutdown have occurred. Because fewer automatic protective fetures are operative during cold shutdowns, both prevention and termination of events depend heavily on operator action. The preservation of RHRS cooling should be an important priority in all shutdown operations, particularly where there is substantial decay heat and a reduced water inventory. 13 refs., 3 figs., 4 tabs

  16. Removal of decay heat by specially designed isolation condensers for advanced heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dhawan, M L; Bhatia, S K [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    For Advanced Heavy Water Reactor (AHWR), removal of decay heat and containment heat is being considered by passive means. For this, special type of isolation condensers are designed. Isolation condensers when submerged in a pool of water, are the best choice because condensation of high temperature steam is an extremely efficient heat transfer mechanism. By the use of isolation condensers, not only heat is removed but also pressure and temperature of the system are automatically controlled without losing the coolant and without using conventional safety relief valves. In this paper, design optimisation studies of isolation condensers of different types with natural circulation for the removal of core decay heat for AHWR is presented. (author). 8 refs., 2 figs.

  17. Overview report of RAMONA-NEPTUN program on passive decay heat removal

    International Nuclear Information System (INIS)

    Weinberg, D.; Rust, K.; Hoffmann, H.

    1996-03-01

    The design of the advanced sodium-cooled European Fast Reactor provides a safety graded decay heat removal concept which ensures the coolability of the primary system by natural convection when forced cooling is lost. The findings of the RAMONA and NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The operation of the decay heat exchangers being installed in the upper plenum causes the formation of a thermal stratification associated with a pronounced temperature gradient. The vertical extent of the stratification and the qualitity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. A delayed startup time of the decay heat exchangers leads only to a slight increase of the temperatures in the upper plenum. A complete failure of half of the decay heat exchangers causes a higher temperature level in the primary system, but does not alter the global temperature distribution. The transient development of the temperatures is faster going on in a three-loop model than in a four-loop model due to the lower amount of heat stored in the compacter primary vessel. If no coolant reaches the core inlet side via the intermediate heat exchangers, the core remains coolable. In this case, cold water of the upper plenum penetrates into the subassemblies (thermosyphon effects) and the interwrapper spaces existing in the NEPTUN core. The core coolability from above is feasible without any difficulty though the temperatures increase to a minor degree at the top end of the core. The thermal hydraulic computer code FLUTAN was applied for the 3D numerical simulation of the majority of the steady state RAMONA and NEPTUN tests as well as for selected transient RAMONA tests. (orig./HP) [de

  18. Specialists' meeting on evaluation of decay heat removal by natural convection

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-02-01

    Decay heat removal by natural convection (DHRNC) is essential to enhancing the safety of liquid metal fast reactors (LMFRs). Various design concepts related to DHRNC have been proposed and experimental and analytical studies have been carried out in a number of countries. The purpose of this Specialists' Meeting on 'Decay Heat Removal by Natural Convection' organized by the International Working Group on Fast Reactors IAEA, is to exchange information about the state of the art related to methodologies on evaluation of DHRNC features (experimental studies and code developments) and to discuss problems which need to be solved in order to evaluate DHRNC properly and reasonably. The following main topical areas were discussed by delegates: Overview; Experimental studies and code validation; Design study. Two main DHR systems for LMFR are under consideration: (i) direct reactor auxiliary cooling system (DRACS) with immersed DFIX in main vessel, intermediate sodium loop and sodium-air heat exchanger; and (ii) auxiliary cooling system which removes heat from the outside surface of the reactor vessel by natural convection of air (RVACS). The practicality and economic viability of the use of RVACS is possible up to a modular type reactor or a middle size reactor based on current technology. For the large monolithic plant concepts DRACS is preferable. The existing experimental results and the codes show encouraging results so that the decay heat removal by pure natural convection is feasible. Concerning the objective, 'passive safety', the DHR by pure natural convection is essential feature to enhance the reliability of DHR.

  19. Specialists' meeting on evaluation of decay heat removal by natural convection

    International Nuclear Information System (INIS)

    1993-02-01

    Decay heat removal by natural convection (DHRNC) is essential to enhancing the safety of liquid metal fast reactors (LMFRs). Various design concepts related to DHRNC have been proposed and experimental and analytical studies have been carried out in a number of countries. The purpose of this Specialists' Meeting on 'Decay Heat Removal by Natural Convection' organized by the International Working Group on Fast Reactors IAEA, is to exchange information about the state of the art related to methodologies on evaluation of DHRNC features (experimental studies and code developments) and to discuss problems which need to be solved in order to evaluate DHRNC properly and reasonably. The following main topical areas were discussed by delegates: Overview; Experimental studies and code validation; Design study. Two main DHR systems for LMFR are under consideration: (i) direct reactor auxiliary cooling system (DRACS) with immersed DFIX in main vessel, intermediate sodium loop and sodium-air heat exchanger; and (ii) auxiliary cooling system which removes heat from the outside surface of the reactor vessel by natural convection of air (RVACS). The practicality and economic viability of the use of RVACS is possible up to a modular type reactor or a middle size reactor based on current technology. For the large monolithic plant concepts DRACS is preferable. The existing experimental results and the codes show encouraging results so that the decay heat removal by pure natural convection is feasible. Concerning the objective, 'passive safety', the DHR by pure natural convection is essential feature to enhance the reliability of DHR

  20. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  1. ALPHA - The long-term passive decay heat removal and aerosol retention program

    International Nuclear Information System (INIS)

    Guentay, S.; Varadi, G.; Dreier, J.

    1996-01-01

    The Paul Scherrer Institute initiated the major new experimental and analytical program ALPHA in 1990. The program is aimed at understanding the long-term decay heat removal and aerosol questions for the next generation of Passive Light Water Reactors. The ALPHA project currently includes four major items: the large-scale, integral system behaviour test facility PANDA, which will be used to examine multidimensional effects of the SBWR decay heat removal system; an investigation of the thermal hydraulics of natural convection and mixing in pools and large volumes (LINX); a separate-effects study of aerosols transport and deposition in plenum and tubes (AIDA); while finally, data from the PANDA facility and supporting separate effects tests will be used to develop and qualify models and provide validation of relevant system codes. The paper briefly reviews the above four topics and current status of the experimental facilities. (author). 3 refs, 12 figs

  2. ALPHA - The long-term passive decay heat removal and aerosol retention program

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, S; Varadi, G; Dreier, J [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-12-01

    The Paul Scherrer Institute initiated the major new experimental and analytical program ALPHA in 1990. The program is aimed at understanding the long-term decay heat removal and aerosol questions for the next generation of Passive Light Water Reactors. The ALPHA project currently includes four major items: the large-scale, integral system behaviour test facility PANDA, which will be used to examine multidimensional effects of the SBWR decay heat removal system; an investigation of the thermal hydraulics of natural convection and mixing in pools and large volumes (LINX); a separate-effects study of aerosols transport and deposition in plenum and tubes (AIDA); while finally, data from the PANDA facility and supporting separate effects tests will be used to develop and qualify models and provide validation of relevant system codes. The paper briefly reviews the above four topics and current status of the experimental facilities. (author). 3 refs, 12 figs.

  3. Decay heat removal plan of the SNR-300: a licensed concept

    International Nuclear Information System (INIS)

    Morgenstern, F.H.; Gyr, W.; Stoetzel, H.; Vossebrecker, H.

    1976-01-01

    The report describes how the decay heat removal plan of the SNR-300 has been established in 3 essential licensing steps, thus giving a very significant example for the slow but steady progress in the overall licensing process of the plant. (1) Introduction of an ECCS in addition to the 3 main heat transfer chains as a back-up for rather unlikely and undefined occurrences, 1970; (2) Experimental and computational demonstration of a reliable functioning of the in-vessel natural convection of the fluid flow, 1974; and (3) Proof of fulfilling the general safety and specific reliability criteria for the overall decay heat removal plan; i.e., the 3 main heat transfer chains with specific installations on the steam/water system side and the ECCS, 1976. Some special problem areas, for instance the cavity concept provided for the pipe fracture accident, have still to be licensed, but they do not contribute considerably to the overall risk

  4. Reliability assessment on decay heat removal system of a fast reactor

    International Nuclear Information System (INIS)

    Hioki, Kazumasa

    1991-01-01

    The reliability of a decay heat removal system (DHRS) is influenced by the success criteria, the components which constitute the system, the support systems configuration, and the mission time. Assessments were performed to investigate quantitatively the effects of these items. Failure probabilities of DHRS under forced or natural circulation modes were calculated and then components and systems of large importance for each mode were identified. (author)

  5. Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

    International Nuclear Information System (INIS)

    Stehle, H.; Damm, G.; Jansing, W.

    1994-01-01

    Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA. (orig.)

  6. Analysis of a convection loop for GFR post-LOCA decay heat removal

    International Nuclear Information System (INIS)

    Williams, W.C.; Hejzlar, P.; Saha, P.

    2004-01-01

    A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

  7. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

  8. Preliminary study of the decay heat removal strategy for the gas demonstrator allegro

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Gusztáv, E-mail: gusztav.mayer@energia.mta.hu [Hungarian Academy of Sciences, Centre for Energy Research, P.O. Box 49, H-1525 Budapest (Hungary); Bentivoglio, Fabrice, E-mail: fabrice.bentivoglio@cea.fr [CEA/DEN/DM2S/STMF/LMES, F-38054, Grenoble (France)

    2015-05-15

    Highlights: • Improved decay heat removal strategy was adapted for the 75 MW ALLEGRO MOX core. • New nitrogen injection strategy was proposed for the DEC LOCA transients. • Preliminary CATHARE study shows that most of the investigated transients fulfill criteria. • Further improvements and optimizations are needed for nitrogen injection. - Abstract: The helium cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV International Forum. Since no gas cooled fast reactor has ever been built, a medium power demonstrator reactor – named ALLEGRO – is necessary on the road towards the 2400 MWth GFR power reactor. The French Commissariat à l’Energie Atomique (CEA) completed a wide range of studies during the early stage of development of ALLEGRO, and later the ALLEGRO reactor concept was developed in several European Union projects in parallel with the GFR2400. The 75 MW thermal power ALLEGRO is currently developed in the frame of the European ALLIANCE project. As a result of the collaboration between CEA and the Hungarian Academy of Sciences Centre for Energy Research (MTA EK) new improvements were done in the safety approach of ALLEGRO. A complete Decay Heat Removal (DHR) strategy was devised, relying on the primary circuits as a first way to remove decay heat using pony-motors to drive the primary blowers, and on the secondary and tertiary circuits being able to work in forced or natural circulation. Three identical dedicated loops circulating in forced convection are used as a second way to remove decay heat, and these loops can circulate in natural convection for pressurized transients, providing a third way to remove decay heat in case of accidents when the primary circuit is still under pressure. The possibility to use nitrogen to enhance both forced and natural circulation is discussed. This DHR strategy is supported by a wide range of accident transient simulations performed using the CATHARE2 code

  9. Meeting of Specialists on the Reliability of Decay Heat Removal Systems for Fast Reactors. Summary Report

    International Nuclear Information System (INIS)

    1975-10-01

    The Specialists Meeting on Reliability of Decay Heat Removal Systems proposed for Fast Reactors was sponsored by the UKAEA Safety & Reliability Directorate and held at Harwell between 28th April and 1st May, 1975. The meeting was attended by delegates from six countries - (USA, Federal Republic of Germany, France, Japan, USSR and the UK). A list of participants is included in an Appendix to this report. The subject matter of the meeting was concerned with the degree to which the ability to maintain decay heat removal from a fast reactor after shutdown in normal and abnormal circumstances could be guaranteed by design provisions and substantiated by reliability analysis techniques, operational testing etc. Consideration of conditions prevailing after a hypothetical core melt down incident were not included in the subject matter. The deliberations of the meeting were focussed at each working session on a defined theme and its dependant topics as shown in the detailed Agenda included in this report. Although provision had been made in the Agenda for a limited amount of discussion of the decay heat rejection problems of Gas Cooled Fast Reactors, delegates had no contributions to offer on this subject. During each session a Recording Secretary prepared a summary of the main points made by national delegates and of the resulting recommendations and conclusions. These draft summaries were made available to delegates during subsequent sessions of the meeting and approved by them for inclusion in the Summary, General Conclusions and Recommendations provided under Table of Contents (item 3 and 4)

  10. Uncertainty correlation in stochastic safety analysis of natural circulation decay heat removal of liquid metal reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira

    2009-01-01

    Since various uncertainties of input variables are involved and nonlinearly-correlated in the Best Estimate (BE) plant dynamics code, it is of importance to evaluate the importance of input uncertainty to the computational results and to estimate the accuracy of the confidence level of the results. In order to estimate the importance and the accuracy, the authors have applied the stochastic safety analysis procedure using the Latin Hypercube sampling method to Liquid Metal Reactor (LMR) natural circulation Decay Heat Removal (DHR) phenomenon in the present paper. 17 input variables are chosen for the analyses and 5 influential variables, which affect the maximum coolant temperature at the core in a short period of time (several tens seconds), are selected to investigate the importance by comparing with the full-scope parametric analysis. As a result, it has been demonstrated that a comparative small number of samples is sufficient enough to estimate the dominant input variable and the confidence level. Furthermore, the influence of the sampling method on the accuracy of the upper tolerance limit (confidence level of 95%) has been examined based on the Wilks' formula. (author)

  11. Preliminary Analysis on Decay Heat Removal Capability of Helium Cooled Solid Breeder Test Blanket Module

    International Nuclear Information System (INIS)

    Ahn, Mu Young; Cho, Seung Yon; Kim, Duck Hoi; Lee, Eun Seok; Kim, Hyung Seok; Suh, Jae Seung; Yun, Sung Hwan; Cho, Nam Zin

    2007-01-01

    One of the main ITER goals is to test and validate design concepts of tritium breeding blankets relevant to DEMO or fusion power plants. Korea Helium-Cooled Solid Breeder (HCSB) Test Blanket Module (TBM) has been developed with overall objectives of achieving this goal. The TBM employs high pressure helium to cool down the First Wall (FW), Side Wall (SW) and Breeding Zone (BZ). Therefore, safety consideration is a part of the design process. Each ITER Party performing the TBM program is requested to reach a similar level of confidence in the TBM safety analysis. To meet ITER's request, Failure Mode and Effects Analysis (FMEA) studies have been performed on the TBM to identify the Postulated Initial Event (PIE). Although FMEA on the KO TBM has not been completed, in-vessel, in-box and ex-vessel Loss Of Coolant Accident (LOCA) are considered as enveloping cases of PIE in general. In this paper, accidental analyses for the three selected LOCA were performed to investigate the decay heat removal capability of the TBM. To simulate transient thermo-hydraulic behavior of the TBM for the selected scenarios, RELAP5/MOD3.2 code was used

  12. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

  13. Passive decay heat removal by natural air convection after severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

    1995-09-01

    The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

  14. RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Polidori, Massimiliano; Meloni, Paride; Tarantino, Mariano; Di Piazza, Ivan

    2015-01-01

    Highlights: • The CIRCE DHR experiments simulate LOHS+LOF transients in LFR systems. • Decay heat removal by natural circulation through immersed heat exchangers is investigated. • The RELAP5 simulation of DHR experiments is presented. • The SIMMER-III simulation of DHR experiments is presented. • The focus is on the transition from forced to natural convection and stratification in a large pool. - Abstract: In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool. The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient. In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal. Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend

  15. A value/impact assessment for alternative decay heat removal systems

    International Nuclear Information System (INIS)

    Cave, L.; Kastenberg, W.E.; Lin, K.Y.

    1984-01-01

    A Value/Impact assessment for several alternative decay heat removal systems has been carried out using several measures. The assessment is based on an extension of the methodology presented in the Value/Impact Handbook and includes the effects of uncertainty. The assessment was carried out as a function of site population density, existing plant features, and new plant features. Value/Impact measures based on population dose are shown to be sensitive to site, while measures which monetize and aggregate risk are less so. The latter are dominated by on-site costs such as replacement power costs. (orig.)

  16. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  17. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  18. Analysis of Decay Heat Removal by Natural Convection in LMR with a Combined Steam Generator

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Eoh, Jae Hyuk; Han, Ji Woong; Lee, Tae Ho

    2011-01-01

    Liquid metal reactors (LMRs) conventionally employ an intermediate heat transport system (IHTS) to protect the nuclear core during a sodium-water reaction (SWR) event. However these SWR-related components increase plant construction costs. In order to eliminate the need for an IHTS, a combined steam generator, which is an integrated heat exchanger of a steam generator and intermediate heat exchanger (IHX), was proposed by the Korea Atomic Energy Research Institute (KAERI). The objective of this work is to analyze the natural circulation heat removal capability of the rector system using a combined steam generator. As a means of decay heat removal, a normal heat transport path is composed of a primary sodium system, intermediate lead-bismuth circuit combined with SG and steam/water system. This paper presents the results of the possible temperature and natural circulation flows in all circuits during a steady state for a given reactor power level varied as a function of time

  19. Design of Passive Decay Heat Removal System using Mercury Thermosyphon for SFR

    Energy Technology Data Exchange (ETDEWEB)

    You, Byung Hyun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, thermosyphon application is suggested to accomplish the fully passive safety grade system and compactness of components via enhance the heat removal performance. A two-phase evaporating thermosyphon operates when the evaporator is heated, the working fluid start boiling, the vapor that is formed moves to the condenser, where it is condensed on the walls, giving up the heat of phase change to the cooling fluid. Gravity forces cause the condensate to condensed liquid flow to the evaporator again. These processes occur continuously, which causes transfer of heat from evaporator to condenser vice versa. After the thermal design and performance evaluation, the results were compared with the performance of conventional DRACS system. For the same amount of decay heat removal performance of PDRC system of KALIMER-600 mercury thermosyphon system can archive around 30∼50% of compactness. For the detailed design, improved analytical model and experimental data for the validation will be required to specify the new DHR system.

  20. Evaluation of Heat Removal Performance of Passive Decay Heat Removal system for S-CO{sub 2} Cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The modular systems is able to be transported by large trailer. Moreover, dry cooling system is applied for waste heat removal. The characteristics of MMR takes wide range of construction area from coast to desert, isolated area and disaster area. In MMR, Passive decay heat removal system (PDHRS) is necessary for taking the advantage on selection of construction area where external support cannot be offered. The PDHRS guarantees to protect MMR without external support. In this research, PDHRS of MMR is introduced and decay heat removal performance is analyzed. The PDHRS guarantees integrity of reactor coolant system. The high level of decay heat (2 MW) can be removed by PDHRS without offsite power.

  1. Improved Design Concept for ensuring the Passive Decay Heat Removal Performance of an SFR

    International Nuclear Information System (INIS)

    Eoh, Jae Hyuk; Lee, Tae Ho; Han, Ji Woong; Kim, Seong O

    2011-01-01

    In order to enhance the operational reliability of a purely passive decay heat removal system in KALIMER, which is named as PDRC, three design options to prevent a sodium freezing in an intermediate decay heat removal circuit were proposed, and their feasibilities was quantitatively evaluated. For all the options, more specific design considerations were made to confirm their feasibility to properly materialize their concepts in a practical system design procedure, and the general definitions for a purely passive concept and its design features have been discussed. A numerical study to evaluate the coastdown flow effect of the primary pump was performed to figure out the early stage DHR capability inside reactor pool during a loss of normal heat sink accident. The thermal-hydraulic calculations have been made by using the COMMIX-1AR/P code, and it was found that the initiation of heat removal by DHX could be accelerated by the increase of the coastdown time but it needs a large-sized flywheel. For the demonstration of the innovative concept, a large scale sodium thermal-hydraulic test facility is currently being designed. It is very difficult to reproduce both a hydrodynamic and a thermodynamic similarity to the prototype plant if the thermal driving head is determined by structure-to-fluid heat transfer under natural circulation flow. Hence the similitude requirements for the sodium thermal-hydraulic test facility employing natural convection heat transfer were developed, and the preliminary design data of the test facility by implementing proper scaling methodologies was produced. The design restrictions imposed on the test facility and the scaling distortions of the design data to the full-scale system were also discussed

  2. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  3. Experimental investigations on scaled models for the SNR-2 decay heat removal by natural convection

    International Nuclear Information System (INIS)

    Hoffmann, H.; Weinberg, D.; Tschoeke, H.; Frey, H.H.; Pertmer, G.

    1986-01-01

    Scaled water models are used to prove the mode of function of the decay heat removal by natural convection for the SNR-2. The 2D and 3D models were designed to reach the characteristic numbers (Richardson, Peclet) of the reactor. In the experiments on 2D models the position of the immersed cooler (IC) and the power were varied. Temperature fields and velocities were measured. The IC installed as a separate component in the hot plenum resulted in a very complex flow behavior and low temperatures. Integrating the IC in the IHX showed a very simple circulating flow and high temperatures within the hot plenum. With increasing power only slightly rising temperature differences within the core and IC were detected. Recalculations using the COMMIX 1B code gave qualitatively satisfying results. (author)

  4. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  5. Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    1984-01-01

    A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

  6. Studies related to emergency decay heat removal in EBR-II

    International Nuclear Information System (INIS)

    Singer, R.M.; Gillette, J.L.; Mohr, D.; Tokar, J.V.; Sullivan, J.E.; Dean, E.M.

    1979-01-01

    Experimental and analytical studies related to emergency decay heat removal by natural circulation in the EBR-II heat transport circuits are described. Three general categories of natural circulation plant transients are discussed and the resultant reactor flow and temperature response to these events are presented. these categories include the following: (1) loss of forced flow from decay power and low initial flow rates; (2) reactor scram with a delayed loss of forced flow; and (3) loss of forced flow with a plant protective system activated scram. In all cases, the transition from forced to natural convective flow was smooth and the peak in-core temperature rises were small to moderate. Comparisons between experimental measurements in EBR-II and analytical predictions of the NATDEMO code are included

  7. Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1997-01-01

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

  8. Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U.; Mueller, U. [Forschungszentrum Karlsruhe - Technik und Umwelt Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF), Karlsruhe (Germany)

    1997-12-31

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

  9. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  10. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A.S.

    2010-01-01

    Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs - of the so-called second generation - will soon need replacement, and a third generation is now being made available, offering increased safety while still based on light water technology. For the longer-term future, viz. beyond the year 2030, R and D is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For de-pressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure

  11. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A. S.

    2010-09-01

    The majority of NPPs worldwide are currently light water reactors, using ordinary water as both coolant and moderator. (...) For the longer-term future, viz. beyond the year 2030, Research and Development is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be

  12. Summary report of NEPTUN investigations into the steady state thermal hydraulics of the passive decay heat removal

    International Nuclear Information System (INIS)

    Rust, K.; Weinberg, D.; Hoffmann, H.; Frey, H.H.; Baumann, W.; Hain, K.; Leiling, W.; Hayafune, H.; Ohira, H.

    1995-12-01

    During the course of steady state NEPTUN investigations, the effects of different design and operating parameters were studied; in particular: The shell design of the above core sturcture, the core power, the number of decay heat exchangers put in operation, the complete flow path blockage at the primary side of the intermediate heat exchangers, and the fluid level in the primary vessel. The findings of the NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The interwrapper flow makes an essential contribution to that behavior. The decay heat exchangers installed in the upper plenum cause a thermal stratification associated with a pronounced gradient. The vertical extent of the stratification and the quantity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. An increase of the core power or a reduction of the number of decay heat exchangers being in operation leads to a higher temperature level in the primary system but does not alter the global temperature distribution. In the case that no coolant enters the inlet windows at the primary side of the intermediate and decay heat exchangers, the core remains coolable as far as the primary vessel is filled with fluid up to a minimum level. Cold water penetrates from the upper plenum into the core and removes the decay heat. The thermal hydraulic computer code FLUTAN was applied for the three-dimensional numerical simulation of the majority of NEPTUN tests reported here. The comparison of computed against experimental data indicates a qualitatively and quantitatively satisfying agreement of the findings with respect to the field of isotherms as well as the temperature profiles in the upper plenum and within the core region of very complex geometry. (orig./HP) [de

  13. Design of DC Conduction Pump for PGSFR Active Decay Heat Removal System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dehee; Hong, Jonggan; Lee, Taeho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A DC conduction pump has been designed for the ADHRS of PGSFR. A VBA code developed by ANL was utilized to design and optimize the pump. The pump geometry dependent parameters were optimized to minimize the total current while meeting the design requirements. A double-C type dipole was employed to produce the calculated magnetic strength. Numerical simulations for the magnetic field strength and its distribution around the dipole and for the turbulent flow under magnetic force will be carried out. A Direct Current (DC) conduction Electromagnetic Pump (EMP) has been designed for Active Decay Heat Removal System (ADHRS) of PGSFR. The PGSFR has active as well as passive systems for the DHRS. The passive DHRS (PDHRS) works by natural circulation head and the ADHRS is driven by an EMP for the DHRS sodium loop and a blower for the finned-tube sodium-to-air heat exchanger (FHX). An Annular Linear Induction Pump (ALIP) can be also considered for the ADHRS, but DC conduction pump has been chosen. Selection basis of DHRS EMP is addressed and EMP design for single ADHRS loop with 1MWt heat removal capacity is introduced.

  14. Shutdown decay heat removal analysis of a Babcock and Wilcox pressurized water reactor: Case study

    International Nuclear Information System (INIS)

    Cramond, W.R.; Ericson, D.M. Jr.; Sanders, G.A.

    1987-03-01

    This is one of six case studies for USI A-45 Decay Heat Removal (DHR) Requirements. The purpose of this study is to identify any potential vulnerabilities in the DHR systems of a typical Babcock and Wilcox PWR, to suggest possible modifications to improve the DHR capability, and to assess the value and impact of the most promising alternatives to the existing DHR systems. The systems analysis considered small LOCAs and transient internal initiating events, and seismic, fire, extreme wind, internal and external flood, and lightning external events. A full-scale systems analysis was performed with detailed fault trees and event trees including support system dependencies. The system analysis results were extrapolated into release categories using applicable past PRA phenomenological results and improved containment failure mode probabilities. Public consequences were estimated using site specific CRAC2 calculations. The Value-Impact (VI) analysis of possible alternatives considered both onsite and offsite impacts arriving at several risk measures such as averted population dose out to a 50-mile radius and dollars per person rem averted. Uncertainties in the VI analysis are discussed and the issues of feed and bleed and secondary blowdown are analyzed

  15. Experimental and numerical simulation of passive decay heat removal by sump cooling after cool melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.; Kuhn, D.; Mueller, U.

    1997-01-01

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase and two-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is 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. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software package Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a first statement with regard to the feasibility of the sump cooling concept. 11 refs., 9 figs., 3 tabs

  16. Analysis of removal of residual decay heat from interim storage facilities by means of the CFD program FLUENT

    International Nuclear Information System (INIS)

    Stratmann, W.; Hages, P.

    2004-01-01

    Within the scope of nuclear licensing procedures of on-site interim storage facilities for dual purpose casks it is necessary, among other things, to provide proof of sufficient removal of the residual decay heat emitted by the casks. The results of the analyses performed for this purpose define e.g. the boundary conditions for further thermal analyses regarding the permissible cask component temperatures or the maximum permissible temperatures of the fuel cladding tubes of the fuel elements stored in the casks. Up to now, for the centralized interim storage facilities in Germany such analyses were performed on the basis of experimental investigations using scaled-down storage geometries. In the engineering phase of the Lingen on-site interim storage facility, proof was furnished for the first time using the CFD (computational fluid dynamics) program FLUENT. The program FLUENT is an internationally recognized and comprehensively verified program for the calculation of flow and heat transport processes. Starting from a brief discussion of modeling and the different boundary conditions of the computation, this contribution presents various results regarding the temperatures of air, cask surfaces and storage facility components, the mass flows through the storage facility and the heat transfer at the cask surface. The interface point to the cask-specific analyses is defined to be the cask surface

  17. Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

    2009-01-01

    An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

  18. Design of passive decay heat removal system using thermosyphon for low temperature and low pressure pool type LWR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; You, Byung Hyun; Jung, Yong Hun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In seawater desalination process which doesn't need high temperature steam, the reactor has profitability. KAIST has be developing the new reactor design, AHR400, for only desalination. For maximizing safety, the reactor requires passive decay heat removal system. In many nuclear reactors, DHR system is loop form. The DHR system can be designed simple by applying conventional thermosyphon, which is fully passive device, shows high heat transfer performance and simple structure. DHR system utilizes conventional thermosyphon and its heat transfer characteristics are analyzed for AHR400. For maximizing safety of the reactor, passive decay heat removal system are prepared. Thermosyphon is useful device for DHR system of low pressure and low temperature pool type reactor. Thermosyphon is operated fully passive and has simple structure. Bundle of thermosyphon get the goal to prohibit boiling in reactor and high pressure in reactor vessel.

  19. Development of margin assessment methodology of decay heat removal function against external hazards. (2) Tornado PRA methodology

    International Nuclear Information System (INIS)

    Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

    2014-01-01

    Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method after the TEPCO's Fukushima Daiichi nuclear power station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also the PRA methodology should be developed for the other external hazards (e.g. tornado). In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, tornado hazard curve was estimated by using data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, boards and cars) caused by a tornado. Main damage cause for important structures and components is the missiles and the tornado missiles that can reach those components and structures placed on high elevations were identified, and the failure probabilities of the components and structures against the tornado missiles were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, the event tree was quantified. As a result, the core damage frequency was enough lower than 10 -10 /ry. (author)

  20. Summary report of NEPTUN investigations into transient thermal hydraulics of the passive decay heat removal

    International Nuclear Information System (INIS)

    Weinberg, D.; Hoffmann, H.; Rust, K.; Frey, H.H.; Hain, K.; Leiling, W.; Hayafune, H.

    1995-12-01

    The results corroborate the findings of tests with the RAMONA model. With the core power reduction at scram and the start of the decay heat exchangers operation cold fluid is delivered into the prevailing upper plenum. A temperature stratification develops with distinct large temperature gradients. The onset of natural convection is mainly influenced by two effects, namely, the temperature increase on the intermediate heat exchangers primary sides as a result of which the downward pressures are reduced, and the startup of the decay heat exchangers which leads to a decrease of the buoyancy forces in the core. The temperatures of the upper plenum are systematically reduced as soon as the decay heat exchangers are in operation. Then mixed fluid in the hot plenum reaches the intermediate heat exchangers inlet windows and causes an increase in the core flow rate. The primary pump coastdown curve influences the primary system thermal hydraulics only during the first thousand seconds after scram. The longer the pumps operate the more cold fluid is delivered via the core to the upper plenum. The delay of the start of the decay heat exchangers operation separates the two effects which influence the core mass flow, namely the heatup of the intermediate heat exchangers as well as the formation of the stratification in the upper plenum. Increasing the power as well as the operation of only half of the available decay heat exchangers increase the system temperatures. A permeable above core structure produces a temperature stratification along the total upper plenum, and therefore a lower temperature gradient in the region between core outlet and lower edge of the above core structure, in comparison to the impermeable design. A complete flow path blockage of the primary fluid through the intermediate heat exchangers leads to an enhanced cooling effect of the interstitial flow and gives rise to a thermosiphon effect inside the core elements. (orig./GL) [de

  1. A portable backup power supply to assure extended decay heat removal during natural phenomena-induced station blackout

    International Nuclear Information System (INIS)

    Proctor, L.D.; Merryman, L.D.; Sallee, W.E.

    1989-01-01

    The High Flux Isotope Reactor (HFIR) is a light water cooled and moderated flux-trap type research reactor located at Oak Ridge National Laboratory (ORNL). Coolant circulation following reactor shutdown is provided by the primary coolant pumps. DC-powered pony motors drive these pumps at a reduced flow rate following shutdown of the normal ac-powered motors. Forced circulation decay heat removal is required for several hours to preclude core damage following shutdown. Recent analyses identified a potential vulnerability due to a natural phenomena-induced station blackout. Neither the offsire power supply nor the onsite emergency diesel generators are designed to withstand the effects of seismic events or tornadoes. It could not be assured that the capacity of the dedicated batteries provided as a backup power supply for the primary coolant pump pony motors is adequate to provide forced circulation cooling for the required time following such events. A portable backup power supply added to the plant to address this potential vulnerability is described

  2. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  3. Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

    International Nuclear Information System (INIS)

    Schubert, B.K.

    1984-07-01

    A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

  4. A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A., E-mail: aaron@epiney.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland); Alpy, N., E-mail: nicolas.alpy@cea.fr [CEA, DEN, Service d' Etudes des Systemes Innovants, F-13108 Saint Paul Lez Durance (France); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Chawla, R., E-mail: rakesh.chawla@psi.ch [Paul Scherrer Institute PSI, Villigen (Switzerland); Ecole Polytechnique Federale EPFL, Lausanne (Switzerland)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer An analytical model predicting Brayton cycle off-design steady states, is developed. Black-Right-Pointing-Pointer The model is used to design an autonomous decay heat removal system for the GFR. Black-Right-Pointing-Pointer Predictions of the analytical model are verified using CATHARE. Black-Right-Pointing-Pointer CATHARE code is used to simulate a set of GFR safety depressurization transients using this device. Black-Right-Pointing-Pointer Convenient turbo-machine designs exist for the targeted autonomous decay heat removal for a wide pressure range. - Abstract: This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping model', is described first. This is based on simplified thermodynamic and aerodynamic equations, and was developed to highlight design choices. Two different machine designs are analyzed: a Brayton loop turbo-machine working with helium, and a second one working with nitrogen, since nitrogen is the heavy gas foreseen to be injected into the primary system to enhance the natural convection under loss-of-coolant-accident (LOCA) conditions. Simulations of the steady-state and transient behavior of the proposed device have then been carried out using the CATHARE code. These serve to confirm the insights obtained from usage of the

  5. Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

    Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

  6. Experiences with on line fault detection system for protection system logic and decay heat removal system logic in Dhruva

    International Nuclear Information System (INIS)

    Ramkumar, N.; Dutta, P.K.; Darbhe, M.D.; Bharadwaj, G.

    2001-01-01

    Dhruva is a 100 MW (Thermal) natural uranium fuelled, vertical core, tank type multi purpose research reactor with heavy water acting as moderator, coolant and reflector. Helium is used as cover gas for heavy water system. Reactor Protection System and Decay Heat Removal System (DHRS) have triplicated instrumented channels. The logic for these systems are hybrid in nature with a mixture of relay logic and solid state logic. Fine Impulse Technique(FIT) is employed for On-line fault detection in the solid state logics of these systems. The FIT systems were designed in the early eighties. Operating experiences over the past 15 years has revealed certain deficiencies. In view of this, a microcomputer based state of the art FIT systems for logics of Reactor Protection System and DHRS are being implemented with improved functionalities built into them. This paper describes the operating experience of old FIT systems and improved features of the proposed new FIT systems. (author)

  7. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  8. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  9. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  10. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  11. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

    2015-12-15

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  12. Analytical studies on the impact of using repeated-rib roughness in LMR [Liquid Metal Reactor] decay heat removal systems

    International Nuclear Information System (INIS)

    Obot, N.T.; Tessier, J.H.; Pedersen, D.R.

    1988-01-01

    A numerical study was carried out to determine the effects of roughness on the thermal performance of Liquid Metal Reactor (LMR) decay heat removal systems for a range of possible design configurations and operating conditions. The ranges covered for relative rib height (e/D/sub h/), relative pitch (p/e) and flow attack angle were 0.026--0.103, 5--20 and 0--90 degrees, successively. The heat flux was varied between 1.1 and 21.5 kW/m 2 (0.1 and 2.0 kW/ft 2 ). Calculations were made for three cases: smooth duct with no ribs, ribs on both the guard vessel and collector wall, and ribs on the collector wall only. The results indicate that significant benefits, amounting to nearly two-fold reductions in guard vessel and collector wall temperatures, can be realized by placing repeated ribs on both the guard vessel and the collector wall. The magnitudes of the reduction in the reactor vessel temperature are considerably smaller. In general, the level of improvement, be it with respect to temperature or heat flux, is only mildly affected by changes in rib height or pitch but exhibits greater sensitivity to the assumed value for the system form loss. When the ribs are placed only on the collector wall, the heat removal capability is substantially reduced

  13. Evaluation of the decay heat removal capability using the concept of a thermosyphon in the liquid metal reactor

    International Nuclear Information System (INIS)

    Kim, Y. S.; Sim, Y. S.; Kim, W. K.

    2000-01-01

    A study related to understand the characteristics of the heat pipe and thermosyphon was performed to evaluate their applicabilities to the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced LIquid MEtal Reactor) design. The possible heat transfer rate by the heat pipe and thermosyphon was reviewed to compare the required capability in the PSDRS. A quantitative comparison was done between the current PSDRS and the modified PSDRS with the thermosyphon. The result showed the dominant heat transfer rate in the air channel, e.g. radiation or convection, is different from each other. The total heat transfer rate is not sensitive to the operating temperature of the thermosyphon. The heat removal by the air in the modified case is relatively reduced and the resultant outlet temperature appears less than above 10 .deg. C. A reversal heat transfer between the air and the thermosyphon may exist near the exit of the active heat transfer region. The total heat transfer rate by the modified case showed about 20∼40% increase relative to the reference one

  14. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  15. Development of a steady-state calculation model for the KALIMER PDRC(Passive Decay Heat Removal Circuit)

    International Nuclear Information System (INIS)

    Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Kwon, Young Min; Eoh, Jae Hyuk; Lee, Yong Bum

    2003-06-01

    A sodium circuit has usually featured for a Liquid Metal Reactor(LMR) using sodium as coolant to remove the decay heat ultimately under accidental conditions because of its high reliability. Most of the system codes used for a Light Water Reactor(LWR) analysis is capable of calculating natural circulation within such circuit, but the code currently used for the LMR analysis does not feature stand alone capability to simulate the natural circulation flow inside the circuit due to its application limitation. To this end, the present study has been carried out because the natural circulation analysis for such the circuit is realistically raised for the design with a new concept. The steady state modeling is presented in this paper, development of a transient model is also followed to close the study. The incompressibility assumption of sodium which allow the circuit to be modeled with a single flow, makes the model greatly simplified. Models such as a heat exchanger developed in the study can be effectively applied to other system analysis codes which require such component models

  16. Method and device to remove the decay heat produced in the core of a nuclear reactor

    International Nuclear Information System (INIS)

    Loimann, E.; Reutler, H.

    1977-01-01

    For decay haet removal of the HTGR the heat absorbed by the top reflector is discharged by means of heat exchangers. For this purpose the heat exchangers are arranged between the top bricks consisting of graphite blocks. By convection or forced circulation with the aid of pumps the liquid coolant is flowing in a cycle between the individual heat exchangers connected in parallel and a heat sink arranged outside the containment. The distributing and collection pipes are mounted between the upper and lower thermal shield. The heat exchanger compartments themselves consist of double-walled hollow bodies with a disc-shaped section and a columnar part extending from there to one side respectively. (RW) [de

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

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

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

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  19. Studies on the characteristics of the separated heat pipe system with non-condensible gas for the use of the passive decay heat removal in reactor systems

    International Nuclear Information System (INIS)

    Hayashi, Takao; Ishi, Takayuki; Hayakawa, Hitoshi; Ohashi, Kazutaka

    1997-01-01

    Experiments on the separated heat pipe system of variable conductance type, which enclose non-condensible gas, have been carried out with intention of applying such system to passive decay heat removal of the modular reactors such as HTR plant. Basic experiments have been carried out on the experimental apparatus consisting of evaporator, vapor transfer tube, condenser tube and return tube which returns the condensed liquid back to the evaporator. Water and methanol were examined as the working fluids and nitrogen gas was enclosed as the non-condensible gas. The behaviors of the system were examined for the parametric changes of the heat input under the various pressures of nitrogen gas initially enclosed, including the case without enclosing N 2 gas for the comparison. The results of the experiments shows very clear features of self control characteristics. The self control mechanism was made clear, that is, in such system in which the condensing area in the condenser expands automatically in accordance with the increase of the heat input to keep the system temperature nearly constant. The working temperature of the system are clearly dependent on the pressure of the non-condensable gas initially enclosed, with higher system working temperature with higher initial gas pressure enclosed. The analyses were done on water and methanol as the working fluids, which show very good agreement with the experimental results. A lot of attractive applications are expected including the self switching feature with minimum heat loss during normal operation with maintaining the sufficient heat removal at accidents. (author)

  20. Preliminary design of a Brayton cycle as a standalone Decay Heat Removal system for the Gas-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Epiney, A.; Mikityuk, K.; Chawla, R.; Alpy, N.; Haubensack, D.; Malo, J.Y.

    2009-01-01

    This paper reports a preliminary design study of a Brayton cycle which would be a dedicated, standalone Decay Heat Removal (DHR) loop of the Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed during the GFR pre-conceptual design phase (which was completed by the CEA at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - to the efficiency of the forced convection process which is foreseen for operation down to very low pressures. An analytical model, the so-called 'Brayton scoping' model, is described in the paper. This is based on simplified thermodynamical and aerodynamical equations and was developed to highlight design choices. First simulations of the proposed device's performance during loss-of-coolant-accident (LOCA) transients have been performed using the CATHARE code, and these are also reported. Analysis of the simulation results are consistent with the first insights obtained from usage of the 'Brayton scoping' model, e.g. the turbomachine accelerates during the depressurization process to tend towards a steady rotational speed value which is inversely proportional to the pressure. For small break LOCA events, the device operates successfully as regards its safety function and delivers to the core a relatively unperturbed cooling mass flowrate as a function of pressure change. However, further studies are required for medium to large break sizes, since certain stability concerns have been met in such cases. For example, an unexpected turbomachine stoppage was induced during the transients, resulting in loss of the necessary core cooling mass flow. (author)

  1. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

    2011-01-01

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  2. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  3. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  4. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  5. Analysis of non simultaneous common mode failures. Application to the reliability assessment of the decay heat removal of the RNR 1500 project

    International Nuclear Information System (INIS)

    Natta, M.; Bloch, M.

    1991-01-01

    The experience with the LMFBR PHENIX has shown many cases of failures on identical and redundant components, which were close in time but not simultaneous and due to the same causes such as a design error, an unappropriate material, corrosion, ... Since the decay heat removal (DHR) must be assured for a long period after shutdown of the reactor, the overall reliability of the DHR system depends much on this type of successive failures by common mode causes, for which the usual β factor methods are not appropriate since they imply that the several failures are simultaneous. In this communication, two methods will be presented. The first one was used to assess the reliability of the DHR system of the RNR 1500 project. In this method, one modelize the occurrence of successive failures on n identical files by a sudden jump of the failure rate from the value λ attributed to the first failure to the value λ' attributed to the (n-1) still available files. This method leads to a quite natural quantification of the interest of diversity for highly redundant systems. For the RNR 1500 project where, in case of the loss of normal DHR path through the steam generators, the decay heat is removed by four separated sodium loops of 26 MW unit capacity in forced convection, the probabilistic assessment shows that it is necessary to diversify the sodium-sodium heat exchanger in order to fullfil the upper limit of 10 -7 /year for the probability of failure of DHR. A separate assessment for the main sequence leading to DHR loss was performed using a different method in which the successive failures are interpreted as a premature end of life, the lifetimes being directly used as random variables. This Monte-Carlo type method, which can be applied to any type of lifetime distribution, leads to results consistent to those obtained with the first one

  6. 3D CFD simulations to study the effect of inclination of condenser tube on natural convection and thermal stratification in a passive decay heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Minocha, Nitin [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400 019 (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2016-08-15

    Highlights: • Investigation of three-dimensional natural convection and thermal stratification inside large water pool. • Effect of inclination (α) of condenser tube on fluid flow and heat transfer. • The heat transfer was found to be maximum for α = 90° and minimum for α = 15°. • Laminar-turbulent natural convection and heat transfer in the presence of longitudinal vortices. - Abstract: Many advanced nuclear reactors adopt methodologies of passive safety systems based on natural forces such as gravity. In one of such system, the decay heat generated from a reactor is removed by isolation condenser (ICs) submerged in a large water pool called the Gravity Driven Water Pool (GDWP). The objective of the present study was to design an IC for the passive decay heat removal system (PDHRS) for advanced nuclear reactor. First, the effect of inclination of IC tube on three dimensional temperature and flow fields was investigated inside a pilot scale (10 L) GDWP. Further, the knowledge of these fields has been used for the quantification of heat transfer and thermal stratification phenomenon. In a next step, the knowledge gained from the pilot scale GDWP has been extended to design an IC for real size GDWP (∼10,000 m{sup 3}). Single phase CFD simulation using open source CFD code [OpenFOAM-2.2] was performed for different tube inclination angles (α) (w.r.t. to vertical direction) in the range 0° ⩽ α ⩽ 90°. The results indicate that the heat transfer coefficient increases with increase in tube inclination angle. The heat transfer was found to be maximum for α = 90° and minimum for α = 15°. This behavior is due to the interaction between the primary flow (due to pressure gradient) and secondary flow (due to buoyancy force). The primary flow enhanced the fluid sliding motion at the tube top whereas the secondary flow resulted in enhancement in fluid motion along the circumference of tube. As the angle of inclination (α) of the tube was increased, the

  7. Application study of the heat pipe to the passive decay heat removal system of the modular HTR

    International Nuclear Information System (INIS)

    Ohashi, K.; Okamoto, F.; Hayakawa, H.; Hayashi, T.

    2001-01-01

    To investigate the applicability of the heat pipe to the decay hat removal (DHR) system of the modular HTRs, preliminary study of the Heat Pipe DHR System was performed. The results show that the Heat Pipe DHR System is applicable to the modular HTRs and its heat removal capability is sufficient. Especially by applying the variable conductance heat pipe, the possibility of a fully passive DHR system with lower heat loss during normal operation is suggested. The experiments to obtain the fundamental characteristics data of the variable conductance heat pipe were carried out. The experimental results show very clear features of self-control characteristics. The experimental results and the experimental analysis results are also shown. (author)

  8. Studies on the characteristics of the separated type heat pipe system with non-condensible gas for the use of the passive decay heat removal in reactor systems

    International Nuclear Information System (INIS)

    Hayashi, Takao; Iigaki, Kazuhiko; Ohashi, Kazutaka; Hayakawa, Hitoshi; Yamada, Masao.

    1995-01-01

    This study is the fundamental research by experiments to aim at the development of the complete passive decay heat removal system on the modular reactor systems by the form of the separated type of heat pipe system utilizing the features of both the big latent heat for vaporization from water to steam and easy transportation characteristics. Special intention in our study on the fundamental experiments is to look for the effects in such a separated type of heat pipe system to introduce non-condensible gas such as nitrogen gas together with the working fluid of water. Many interesting findings have been obtained so far on the experiments for the variable conductance heat pipe characteristics from viewpoint of the actual application on the aim said above. This study has been carried out by the joint study between Tokai University and Fuji Electric Co., Ltd. and this paper is made up from the several papers presented so far at both the national and international symposiums under the name of joint study of the both bodies. (author)

  9. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    International Nuclear Information System (INIS)

    Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

    1993-01-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  10. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

    1993-02-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  11. Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung

    2008-01-01

    A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the service

  12. Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the

  13. Development of risk assessment methodology of decay heat removal function against external hazards for sodium-cooled fast reactors. (3) Numerical simulations of forest fire spread and smoke transport as an external hazard assessment methodology development

    International Nuclear Information System (INIS)

    Okano, Yasushi; Yamano, Hidemasa

    2015-01-01

    As a part of a development of the risk assessment methodologies against external hazards, a new methodology to assess forest fire hazards is being developed. Frequency and consequence of the forest fire are analyzed to obtain the hazard intensity curve and then Level 1 probabilistic safety assessment is performed to obtain the conditional core damage probability due to the challenges by the forest fire. 'Heat', 'flame', 'smoke' and 'flying object' are the challenges to a nuclear power plant. For a sodium-cooled fast reactor, a decay heat removal under accident conditions is operated with an ultimate heat sink of air, then, the challenge by 'smoke' will potentially be on the air filter of the system. In this paper, numerical simulations of forest fire propagation and smoke transport were performed with sensibility studies to weather conditions, and the effect by the smoke on the air filter was quantitatively evaluated. Forest fire propagation simulations were performed using FARSITE code. A temporal increase of a forest fire spread area and a position of the frontal fireline are obtained by the simulation, and 'reaction intensity' and 'frontal fireline intensity' as the indexes of 'heat' are obtained as well. The boundary of the fire spread area is shaped like an ellipse on the terrain, and the boundary length is increased with time and fire spread. The sensibility analyses on weather conditions of wind, temperature, and humidity were performed, and it was summarized that 'forest fire spread rate' and 'frontal fireline intensity' depend much on wind speed and humidity. Smoke transport simulations were performed by ALOFT-FT code where three-dimensional spatial distribution of smoke density, especially of particle matters of PM2.5 and PM10, are evaluated. The snapshot outputs, namely 'reaction intensity' and 'position of frontal fireline', from the sensibility studies of the FARSITE were directly utilized as the input data for ALOFT-FT, whereas it is assumed that the

  14. Uncertainties on decay heat power due to fission product data uncertainties; Incertitudes sur la puissance residuelle dues aux incertitudes sur les donnees de produits de fission

    Energy Technology Data Exchange (ETDEWEB)

    Rebah, J

    1998-08-01

    Following a reactor shutdown, after the fission process has completely faded out, a significant quantity of energy known as 'decay heat' continues to be generated in the core. The knowledge with a good precision of the decay heat released in a fuel after reactor shutdown is necessary for: residual heat removal for normal operation or emergency shutdown condition, the design of cooling systems and spent fuel handling. By the summation calculations method, the decay heat is equal to the sum of the energies released by individual fission products. Under taking into account all nuclides that contribute significantly to the total decay heat, the results from summation method are comparable with the measured ones. Without the complete covariance information of nuclear data, the published uncertainty analyses of fission products decay heat summation calculation give underestimated errors through the variance/covariance analysis in consideration of correlation between the basic nuclear data, we calculate in this work the uncertainties on the decay heat associated with the summation calculations. Contribution to the total error of decay heat comes from uncertainties in three terms: fission yields, half-lives and average beta and gamma decay energy. (author)

  15. Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Morris, D.G.

    1989-01-01

    The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab

  16. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors. Proceedings of a specialists meeting held in Juelich, Germany, 6-8 July 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-15

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs.

  17. An experimental study on natural draft-dry cooling tower as part of the passive system for the residual decay heat removal

    International Nuclear Information System (INIS)

    Caruso, G.; Fatone, M.; Naviglio, A.

    2007-01-01

    An experimental apparatus has been built in order to perform sensitivity analysis on the performance of a natural draft-dry cooling tower. This component plays an important role in the passive system for the residual heat decay removal foreseen in the MARS reactor and in the GCFR of the Generation IV reactors. The sensitivity analysis has investigated: 1) the heat exchanger arrangement; two different arrangements have been considered: a horizontal arrangement, in which a system of electrical heaters are placed at the inlet cross section of the tower, and a vertical arrangement, with the heaters distributed vertically around the circumference of the tower. 2) The shape of the cooling tower; by varying the angle of the shell inclination it is possible to obtain a different shape for the tower itself. An upper and a lower angle inclination were modified and by a calculation procedure eleven different configuration were selected. 3) The effect of cross wind on the tower performance. An equation-based procedure to design the dry-cooling tower is presented. In order to evaluate the influence of the shape and the heat exchanger arrangement on the performance of the cooling tower, a geometrical factor (FG) and a thermal factor (FT) are introduced. By analyzing the experimental results, engineering design relations are obtained to model the cooling tower performance. The comparison between the experimental heat transfer coefficient and the heat transfer coefficient obtained by the mathematical procedure shows that there is a good agreement. The obtained results show that it is possible to evaluate the shape and the heat exchanger arrangement to optimize the performance of the cooling tower either in wind-less condition either in presence of cross wind. (authors)

  18. Decay heat experiment and validation of calculation code systems for fusion reactor

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

  19. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

  20. Decay heat uncertainty quantification of MYRRHA

    Directory of Open Access Journals (Sweden)

    Fiorito Luca

    2017-01-01

    Full Text Available MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay heat. Radioactive decay data, independent fission yield and cross section uncertainties/covariances were propagated using two nuclear data sampling codes, namely NUDUNA and SANDY. According to the results, 238U cross sections and fission yield data are the largest contributors to the MYRRHA decay heat uncertainty. The calculated uncertainty values are deemed acceptable from the safety point of view as they are well within the available regulatory limits.

  1. Total decay heat estimates in a proto-type fast reactor

    International Nuclear Information System (INIS)

    Sridharan, M.S.

    2003-01-01

    Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

  2. Decay heat uncertainty quantification of MYRRHA

    OpenAIRE

    Fiorito Luca; Buss Oliver; Hoefer Axel; Stankovskiy Alexey; Eynde Gert Van den

    2017-01-01

    MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS) currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay hea...

  3. Development of limiting decay heat values

    International Nuclear Information System (INIS)

    Khotylev, V.A.; Thompson, J.W.; Gibb, R.A.

    1999-01-01

    A number of tools are used in the assessment of decay heat during an outage of the CANDU-6. Currently, the technical basis for all of these tools is 'CANDU Channel Decay Power', Reference 1. The methods used in that document were limited to channel decay powers. However, for most outage support analysis, decay heat limits are based on bundle heats. Since the production of that document in 1977, new versions of codes, and updates of general-purpose and CANDU-specific libraries have become available. These tools and libraries have both a more formal technical basis than Reference 1, and also a more formal validation base. Using these tools it is now possible to derive decay heat with more specific input parameters, such as fuel composition, heat per unit of fuel, and irradiation history, and to assign systematically derived uncertainty allowances to such decay heat values. In particular, we sought to examine a broad range of likely bundle histories, and thus establish a set of limiting bundle decay beat values, that could serve as a bounding envelope for use in Nuclear Safety Analysis. (author)

  4. Status of the Japanese decay heat standard

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1992-01-01

    Fission product decay heat power plays an important role in the safety evaluation of nuclear power plants, especially for the analysis of hypothetical reactor accident scenarios. The ANS-5.1 decay heat standard for safety evaluation issued in 1979 has been used widely, even in Japan. Since the issuance of the standard, several improvements have been made to measurements and summation calculations. Summation calculations, in particular, have improved because of the adoption of theoretically calculated decay energies for nuclides with incomplete decay data. Taking into consideration those improvements, the Atomic Energy Society of Japan (AESJ) organized a research committee on a standard for decay heat power in nuclear reactors in 1987. The committee issued its recommendation after more than 2 yr discussion. After the AESJ recommendation, the Nuclear Safety Commission of Japan also began to discuss whether the recommendation should be included in its regulatory guide. The commission concluded in 1992 that the recommendation should be approved for licensing analysis of reactors if three times the uncertainties attached to the recommendation are included in the analysis. The AESJ recommendation may now be used for the safety evaluation of reactors in Japan in addition to the standards already used, which include ANS-5.1 (1973), General Electric Corporation (GE) curve, and ANS-5.1 (1979)

  5. Microscopic beta and gamma data for decay-heat needs

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1983-01-01

    Microscopic beta and gamma data for decay-heat needs are defined as absolute-intensity spectral distributions of beta and gamma rays following radioactive decay of radionuclides created by, or following, the fission process. Four well-known evaluated data files, namely the US ENDF/B-V, the UK UKFPDD-2, the French BDN (for fission products), and the Japanese JNDC Nuclear Data Library, are reviewed. Comments regarding the analyses of experimental data (particularly gamma-ray data) are given; the need for complete beta-ray spectral measurements is emphasized. Suggestions on goals for near-term future experimental measurements are presented. 34 references

  6. Decay Heat Calculations for Reactors: Development of a Computer Code ADWITA

    International Nuclear Information System (INIS)

    Raj, Devesh

    2015-01-01

    Estimation of release of energy (decay heat) over an extended period of time after termination of neutron induced fission is necessary for determining the heat removal requirements when the reactor is shutdown, and for fuel storage and transport facilities as well as for accident studies. A Fuel Cycle Analysis Code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) which can generate inventory based on irradiation history and calculate radioactivity and decay heat for extended period of cooling, has been written. The method and data involved in Fuel Cycle Analysis Code ADWITA and some results obtained shall also be presented. (author)

  7. Calculational tracking of decay heat for FFTF plant

    International Nuclear Information System (INIS)

    Cillan, T.F.; Carter, L.L.

    1985-01-01

    A detailed calculational monitoring of decay heat for each assembly on the Fast Flux Test Facility (FFTF) plant is obtained by utilizing a decay heat data base and user friendly computer programs to access the data base. Output includes the time-dependent decay heat for an assembly or a specific set of assemblies, and optional information regarding the curies of activated nuclides along the axial length of the assembly. The decay heat data base is updated periodically, usually at the end of each irradiation cycle. 1 ref., 2 figs

  8. Application of least-squares method to decay heat evaluation

    International Nuclear Information System (INIS)

    Schmittroth, F.; Schenter, R.E.

    1976-01-01

    Generalized least-squares methods are applied to decay-heat experiments and summation calculations to arrive at evaluated values and uncertainties for the fission-product decay-heat from the thermal fission of 235 U. Emphasis is placed on a proper treatment of both statistical and correlated uncertainties in the least-squares method

  9. Jeff-3 and decay heat calculations

    International Nuclear Information System (INIS)

    Huynh, T.D.

    2009-07-01

    The decay heat power, i.e. the residual heat generated by irradiated nuclear fuels, is a significant parameter to define the power of a reactor. A good evaluation of this power depends both on the accuracy of the processing algorithm and on the quality of the physical data used. This report describes the steps carried out, ranging from tests of consistency to the validation by calculations - experiments comparisons, allowing to choose the validated nuclear data. We have compared the Jeff-3 evaluation (only the file 8 containing decay data) with the Jeff-2.2 and Endf/B7.O evaluations through the computation of residual power. It appears that the residual powers computed by the DARWIN code from Jeff-3.1.1 data for short times agree more with experimental data. There is a slight discrepancy (∼ 2%) between Jeff-3.1 and Jeff-3.1.1 on the total residual power computed for PWR UO 2 fuel. For long decay times the discrepancy is more significant between Jeff-3.1.1 and Jeff-2 on the computation of detailed residual powers because some prevailing isotopes have more formation channels taken into account in Jeff-3 and Jeff-3.1.1 than in Jeff-2

  10. Consistency among integral measurements of aggregate decay heat power

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, H.; Sagisaka, M.; Oyamatsu, K.; Kukita, Y. [Nagoya Univ. (Japan)

    1998-03-01

    Persisting discrepancies between summation calculations and integral measurements force us to assume large uncertainties in the recommended decay heat power. In this paper, we develop a hybrid method to calculate the decay heat power of a fissioning system from those of different fissioning systems. Then, this method is applied to examine consistency among measured decay heat powers of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U and {sup 239}Pu at YAYOI. The consistency among the measured values are found to be satisfied for the {beta} component and fairly well for the {gamma} component, except for cooling times longer than 4000 s. (author)

  11. Safety technology qualification of the prestressed cast iron pressure vessel (PCIV) and of the primary cell of the HTR-modul for the passive removal of decay heat, phase 1 (INHR)

    International Nuclear Information System (INIS)

    Warnke, E.P.

    1990-02-01

    During this development program the thermodynamic behaviour of a system was investigated, consisting of a hot working Prestressed Cast Iron Pressure Vessel and an inactive heat sink in the surrounding cavern cell. It could be shown, that the inactive heat removal system designed as a natural circuit can remove the maximum amount of heat of 890 kW during emergency conditions via a natural-draught air cooling tower even under very conservative assumptions and for a 50% loss of cooling pipes. Further it could be shown, that the hot working Prestressed Cast Iron Pressure Vessel has a very safe load carrying behaviour during all normal and upset conditions. (orig.) With 10 tabs., 38 figs., 43 refs [de

  12. An Operators View of Reliability Testing and Decay Heat Rejection Systems

    International Nuclear Information System (INIS)

    Henderson, J.D.C.

    1975-01-01

    The object of this paper is to review the in-situ testing of DHR systems, and to convey policy rather than to indicate a definitive test programme. The test policy is aimed primarily at commissioning the plant and secondly at providing such support for reliability predictions as is practical. Provisions for removal of decay heat from the core and from the reactor tank are described in papers by Broadley and Davies

  13. In-calandria retention of corium in Indian PHWR - experimental simulations with decay heat

    International Nuclear Information System (INIS)

    Nayak, A.K.

    2015-01-01

    The severe accident at Fukushima has compelled the nuclear community to relook at the safety of existing nuclear power plants (NPP) against natural origin events of beyond design basis and prolonged station black out (SBO). A major lesson learned is to assess the capability of the safety systems to cool the reactor core and spent fuel storage facilities in the event of a prolonged station black out (SBO). Similar safety review is planned for the Indian Pressurized Heavy Water Reactors (PHWRs) considering a prolonged SBO. The Indian PHWR is a heavy water-moderated and cooled, natural uranium-fuelled reactor in which the horizontal fuel channels are submerged in a pool of heavy water moderator located inside the calandria vessel. The calandria vessel is surrounded by a calandria vault having large volume of light water. Concerns are raised that in the event of an unmitigated SBO, it may result into a low probable severe accident leading to core melt down. The core melt may further fail the calandria vessel in case the melt is not quenched. If the calandria vessel fails, the corium shall interact with the cold calandria vault water and concrete resulting in generation of large amount of non-condensable gases and steam which will lead to over pressurization of containment and may cause its failure. Therefore, in-calandria corium retention via external cooling using vault water can be considered as an important accident management program in PHWR. In this strategy, the core melt retains inside the calandria vessel by continually removing the stored heat and decay heat through outer surface of the vessel by cooling water and maintaining the integrity of the vessel. The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat by using the calandria vault water. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics

  14. Uncertainties in fission-product decay-heat calculations

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, K.; Ohta, H.; Miyazono, T.; Tasaka, K. [Nagoya Univ. (Japan)

    1997-03-01

    The present precision of the aggregate decay heat calculations is studied quantitatively for 50 fissioning systems. In this evaluation, nuclear data and their uncertainty data are taken from ENDF/B-VI nuclear data library and those which are not available in this library are supplemented by a theoretical consideration. An approximate method is proposed to simplify the evaluation of the uncertainties in the aggregate decay heat calculations so that we can point out easily nuclei which cause large uncertainties in the calculated decay heat values. In this paper, we attempt to clarify the justification of the approximation which was not very clear at the early stage of the study. We find that the aggregate decay heat uncertainties for minor actinides such as Am and Cm isotopes are 3-5 times as large as those for {sup 235}U and {sup 239}Pu. The recommended values by Atomic Energy Society of Japan (AESJ) were given for 3 major fissioning systems, {sup 235}U(t), {sup 239}Pu(t) and {sup 238}U(f). The present results are consistent with the AESJ values for these systems although the two evaluations used different nuclear data libraries and approximations. Therefore, the present results can also be considered to supplement the uncertainty values for the remaining 17 fissioning systems in JNDC2, which were not treated in the AESJ evaluation. Furthermore, we attempt to list nuclear data which cause large uncertainties in decay heat calculations for the future revision of decay and yield data libraries. (author)

  15. ESBWR related passive decay heat removal tests in PANDA

    International Nuclear Information System (INIS)

    Huggenberger, M.; Aubert, C.; Bandurski, T.; Dreier, J.; Fischer, O.; Strassberger, H.J.; Yadigaroglu, G.

    1999-01-01

    A number of test series to investigate passive safety systems for the next generation of Light Water Reactors have been performed in the PANDA multi-purpose facility at the Paul Scherrer Institut (PSI). The large scale thermal-hydraulic test facility allows to investigate LWR containment phenomena and system behaviour. PANDA was first used to examine the Passive Containment Cooling System (PCCS) for the Simplified Boiling Water Reactor (SBWR). In 1996 new test series were initiated; all related to projects of the EC Fourth Framework Programme on Nuclear Fission Safety. One of these projects (TEPSS) is focused on the European Simplified Boiling Water Reactor (ESBWR). The ESBWR containment features and PCCS long-term post LOCA response were investigated in PANDA. The PCCS start-up was demonstrated, the effect of nitrogen hidden somewhere in the drywell and released later in the transient was simulated and the effect of light gases (helium) on the PCCS performance was investigated. Finally, the influence of low PCC pool levels on PCCS and containment performance was examined. The main findings were that the PCCS works as intended and shows generally a favorable and robust long-term post LOCA behaviour. The system starts working even under extreme conditions and trapped air released from the drywell later in the transient does only temporarily reduce the PCCS performance. The new PANDA test series provided an extensive data base which will contribute to further improve containment design of passive plants and allow for system code assessment in a wide parameter range. (author)

  16. Fission yields data generation and benchmarks of decay heat estimation of a nuclear fuel

    Science.gov (United States)

    Gil, Choong-Sup; Kim, Do Heon; Yoo, Jae Kwon; Lee, Jounghwa

    2017-09-01

    Fission yields data with the ENDF-6 format of 235U, 239Pu, and several actinides dependent on incident neutron energies have been generated using the GEF code. In addition, fission yields data libraries of ORIGEN-S, -ARP modules in the SCALE code, have been generated with the new data. The decay heats by ORIGEN-S using the new fission yields data have been calculated and compared with the measured data for validation in this study. The fission yields data ORIGEN-S libraries based on ENDF/B-VII.1, JEFF-3.1.1, and JENDL/FPY-2011 have also been generated, and decay heats were calculated using the ORIGEN-S libraries for analyses and comparisons.

  17. Sensitivity and uncertainty analysis for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Rebah, J.; Lee, Y.K.; Nimal, J.C.; Nimal, B.; Luneville, L.; Duchemin, B.

    1994-01-01

    The calculated uncertainty in decay heat due to the uncertainty in basic nuclear data given in the CEA86 Library, is presented. Uncertainties in summation calculation arise from several sources: fission product yields, half-lives and average decay energies. The correlation between basic data is taken into account. The uncertainty analysis were obtained for thermal-neutron-induced fission of U235 and Pu239 in the case of burst fission and irradiation time. The calculated decay heat in this study is compared with experimental results and with new calculation using the JEF2 Library. (from authors) 6 figs., 19 refs

  18. A proposed Regulatory Guide basis for spent fuel decay heat

    International Nuclear Information System (INIS)

    Hermann, O.W.; Parks, C.V.; Renier, J.P.

    1991-01-01

    A proposed revision to Regulatory Guide 3.54, ''Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation'' has been developed for the US Nuclear Regulatory Commission. The proposed revision includes a data base of decay heat rates calculated as a function of burnup, specific power, cooling time, initial fuel 235 U enrichment and assembly type (i.e., PWR or BWR). Validation of the calculational method was done by comparison with existing measured decay heat rates. Procedures for proper use of the data base, adjustment formulae accounting for effects due to differences in operating history and initial enrichment, and a defensible safety factor were derived. 15 refs., 6 tabs

  19. Reduction of weighing errors caused by tritium decay heating

    International Nuclear Information System (INIS)

    Shaw, J.F.

    1978-01-01

    The deuterium-tritium source gas mixture for laser targets is formulated by weight. Experiments show that the maximum weighing error caused by tritium decay heating is 0.2% for a 104-cm 3 mix vessel. Air cooling the vessel reduces the weighing error by 90%

  20. An application program for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Pham, Ngoc Son; Katakura, Jun-ichi

    2007-10-01

    The precise knowledge of decay heat is one of the most important factors in safety design and operation of nuclear power facilities. Furthermore, decay heat data also play an important role in design of fuel discharges, fuel storage and transport flasks, and in spent fuel management and processing. In this study, a new application program, called DHP (Decay Heat Power program), has been developed for exact decay heat summation calculations, uncertainty analysis, and for determination of the individual contribution of each fission product. The analytical methods were applied in the program without any simplification or approximation, in which all of linear and non-linear decay chains, and 12 decay modes, including ground state and meta-stable states, are automatically identified, and processed by using a decay data library and a fission yield data file, both in ENDF/B-VI format. The window interface of the program is designed with optional properties which is very easy for users to run the code. (author)

  1. A revised ANS standard for decay heat from fission products

    International Nuclear Information System (INIS)

    Schrock, V.E.

    1978-01-01

    The draft ANS 5.1 standard on decay heat was published in 1971 and given minor revision in 1973. Its basis was the best estimate working curve developed by K. Shure in 1961. Liberal uncertainties were assigned to the standard values because of lack of data for short cooling times and large discrepancies among experimental data. Research carried out over the past few years has greatly improved the knowledge of this phenomenon and a major revision of the standard has been completed. Very accurate determination of the decay heat is now possible, expecially within the first 10 4 seconds, where the influence of neutron capture in fission products may be treated as a small correction to the idealized zero capture case. The new standard accounts for differences among fuel nuclides. It covers cooling time to 10 9 seconds, but provides only an ''upper bound'' on the capture correction in the interval 10 4 9 seconds. (author)

  2. Current status of decay heat measurements, evaluations, and needs

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1986-01-01

    Over a decade ago serious concern over possible consequences of a loss-of-coolant accident in a commercial light-water reactor prompted support of several experiments designed specifically to measure the latent energy of beta-ray and gamma-ray emanations from fission products for thermal reactors. This latent energy was termed Decay Heat. At about the same time the American Nuclear Society convened a working group to develop a standard for use in computing decay heat in real reactor environs primarily for regulatory requirements. This working group combined the new experimental results and best evaluated data into a standard which was approved by the ANS and by the ANSI. The primary work since then has been (a) on improvements to computational efforts and (b) experimental measurements for fast reactors. In addition, the need for decay-heat data has been extended well beyond the time regime of a loss-of-coolant accident; new concerns involve, for example, away-from-reactor shipments and storage. The efficacy of the ANS standard for these longer time regimes has been a subject of study with generally positive results. However, a specific problem, namely, the consequences of fission-product neutron capture, remains contentious. Satisfactory resolution of this problem merits a high priority. 31 refs

  3. Contribution of short-lived nuclides to decay heat

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1987-01-01

    Comments are made on the calculation of decay heat, centering on evaluation of average decay energy. It is difficult to obtain sufficiently useful decay diagrams of short lived nucleides. High-energy levels are often missing in inferior decay diagrams, leading to an overestimation of the intensity of beta-rays at low-energy levels. Such an overestimation or underestimation due to the inferiority of a decay diagram is referred to as pandemonium effect. The pandemonium effect can be assessed by means of the ratio of the measured energy of the highest level of the daughter nuclide to the Q β -value of the beta-decay. When a satisfactory decay diagram cannot be obtained, the average decay energy has to be estimated by theoretical calculation. The gross theory for beta-decay proposed by Yamada and Takahashi is employed for the calculation. To carry out the calculation according to this theory, it is required to determine the value for the parameter Q 00 , the lowest energy of the daughter nuclide that meets the selection rule for beta-decay. Currently, Q 00 to be used for this purpose is estimated from data on the energy of the lowest level found in a decay diagram, even if it is inferior. Some examples of calculation of decay heat using the average beta- or gamma-ray energy are shown and compared with measurements. (author)

  4. The effect of load factor on fission product decay heat from discharged reactor fuel

    International Nuclear Information System (INIS)

    Davies, B.S.J.

    1978-07-01

    A sum-of-exponentials expression representing the decay heat power following a burst thermal irradiation of 235 U has been used to investigate the effect of load factor during irradiation on subsequent decay heat production. A sequence of random numbers was used to indicate reactor 'on' and 'off' periods for irradiations which continued for a total of 1500 days at power and were followed by 100 days cooling. It was found that for these conditions decay heat is almost proportional to load factor. Estimates of decay heat uncertainty arising from the random irradiation pattern are also given. (author)

  5. PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

    International Nuclear Information System (INIS)

    De Wet, G. J.; Dent, C.

    2008-01-01

    A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

  6. Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

    International Nuclear Information System (INIS)

    Parks, C.V.; Hermann, O.W.; Ryman, J.C.

    1986-01-01

    In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code

  7. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  8. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  9. Design and transient analyses of emergency passive residual heat removal system of CPR1000

    International Nuclear Information System (INIS)

    Zhang, Y.P.; Qiu, S.Z.; Su, G.H.; Tian, W.X.

    2012-01-01

    Highlights: ► Designing an EPRHRs for CPR1000. ► Developing a RELAP model of the EPRHRs. ► The EPRHRs could take away the decay heat effectively. - Abstract: The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the

  10. A review of U-235 decay heat measurements and calculations

    International Nuclear Information System (INIS)

    Walker, W.H.

    1979-08-01

    Recent scintillator measurements of fission product decay β and γ power, and calorimetric measurements of their sum are analyzed to obtain estimates of E sub(β) and E sub(γ), the β and γ components of the delayed energy per fission in a reactor. Calculations using the ENDF/B-4 fission product file are compared to the measured results and used to estimate the contributions to E sub(β) and E sub(γ) for decay times greater than 10 5 s. A value of E sub(ν), the anti-neutrino component, consistent with the measured component is also calculated. It is found that the decay heat measured in two calorimetric experiments (the sum of the β and γ components) is about 15 percent greater than the separately-measured energies (averages of five β and two γ measurements). Thus, depending on normalization, E sub(β) and E sub(γ) can vary widely. After all experimental uncertainties are taken into account the range of possible values has as lower limits the values calculated using ENDF/B-4, with upper limits about 40 percent greater. (author)

  11. Activation, decay heat, and waste classification studies of the European DEMO concept

    Science.gov (United States)

    Gilbert, M. R.; Eade, T.; Bachmann, C.; Fischer, U.; Taylor, N. P.

    2017-04-01

    Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

  12. Development of whole energy absorption spectrometer for decay heat measurement on fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    To measure decay heat on fusion reactor materials irradiated by D-T neutrons, a Whole Energy Absorption Spectrometer (WEAS) consisting of a pair of large BGO (bismuth-germanate) scintillators was developed. Feasibility of decay heat measurement with WEAS for various materials and for a wide range of half-lives (seconds - years) was demonstrated by experiments at FNS. Features of WEAS, such as high sensitivity, radioactivity identification, and reasonably low experimental uncertainty of {approx} 10 %, were found. (author)

  13. Material composition and nuclear data libraries' influence on nickel-chromium alloys activation evaluation: a comparison with decay heat experiments

    CERN Document Server

    Cepraga, D G

    2000-01-01

    The paper presents the activation analyses on Inconel-600 nickel-chromium alloy. Three activation data libraries, namely the EAF-4.1, the EAF-97 and the FENDL/A-2, and the FENDL/D-2 decay data library, have been used to perform the calculation with the European activation code ANITA-4/M. The neutron flux distribution into the material samples was provided by JAERI as results of 3D Monte-Carlo MCNP transport code experiment simulation. A comparison with integral decay heat measurement performed at the Fusion Neutronics Source (FNS), JAERI, Tokai, Japan, is used to validate the computational approach. The calculation results are given and discussed. The impact of the material composition, including impurities, on the decay heat of samples irradiated in fusion-like neutron spectra is assessed and discussed. The discrepancies calculations-experiments are within the experimental errors, that is between 6% and 10%, except for the short cooling times (less than 40 min after the end of irradiation). To improve calcul...

  14. Properties of Fission-Product decay heat from Minor-Actinide fissioning systems

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro; Mori, Hideki

    2000-01-01

    The aggregate Fission-Product (FP) decay heat after a pulse fission is examined for Minor Actinide (MA) fissiles 237 Np, 241 Am, 243 Am, 242 Cm and 244 Cm. We find that the MA decay heat is comparable but smaller than that of 235 U except for cooling times at about 10 8 s (approx. = 3 y). At these cooling times, either the β or γ component of the FP decay heat for these MA's is substantially larger than the one for 235 U. This difference is found to originate from the cumulative fission yield of 106 Ru (T 1/2 = 3.2x10 7 s). This nuclide is the parent of 106 Rh (T 1/2 = 29.8 s) which is the dominant source of the decay heat at 10 8 s (approx. = 3 y). The fission yield is nearly an increasing function of the fissile mass number so that the FP decay heat is the largest for 244 Cm among the MA's at the cooling time. (author)

  15. Decay heat rates calculated using ORIGEN-S and CINDER10 with common data libraries

    International Nuclear Information System (INIS)

    Brady, M.C.; Hermann, O.W.; Beard, C.A.; Bohnhoff, W.J.; England, T.R.

    1991-01-01

    A set of two benchmark problems were proposed as part of an international comparison of decay heat codes. Problem specifications included explicit fission-yield, decay and capture data libraries to be used in the calculations. This paper describes the results obtained using these common data to perform the benchmark calculations with two popular depletion codes, ORIGEN-S and CINDER10. Short descriptions of the methods used by each of these codes are also presented. Results from other contributors to the international comparison are discussed briefly. This comparison of decay heat codes using common data libraries demonstrates that discrepant results in calculated decay heat rates are the result of differences in the nuclear data input to the codes and not the method of solution. 15 refs., 2 figs., 8 tabs

  16. Detailed comparison between decay heat data calculated by the summation method and integral measurements

    International Nuclear Information System (INIS)

    Rudstam, G.

    1979-01-01

    The fission product library FPLIB has been used for a calculation of the decay heat effect in nuclear fuel. The results are compared with integral determinations and with results obtained using the ENDF/BIV data base. In the case of the beta part, and also for the total decay heat, the FPLIB-data seem to be superior to the ENDF/BIV-data. The experimental integral data are in many cases reproduced within the combined limits of error of the methods. (author)

  17. clubber: removing the bioinformatics bottleneck in big data analyses

    Science.gov (United States)

    Miller, Maximilian; Zhu, Chengsheng; Bromberg, Yana

    2018-01-01

    With the advent of modern day high-throughput technologies, the bottleneck in biological discovery has shifted from the cost of doing experiments to that of analyzing results. clubber is our automated cluster-load balancing system developed for optimizing these “big data” analyses. Its plug-and-play framework encourages re-use of existing solutions for bioinformatics problems. clubber’s goals are to reduce computation times and to facilitate use of cluster computing. The first goal is achieved by automating the balance of parallel submissions across available high performance computing (HPC) resources. Notably, the latter can be added on demand, including cloud-based resources, and/or featuring heterogeneous environments. The second goal of making HPCs user-friendly is facilitated by an interactive web interface and a RESTful API, allowing for job monitoring and result retrieval. We used clubber to speed up our pipeline for annotating molecular functionality of metagenomes. Here, we analyzed the Deepwater Horizon oil-spill study data to quantitatively show that the beach sands have not yet entirely recovered. Further, our analysis of the CAMI-challenge data revealed that microbiome taxonomic shifts do not necessarily correlate with functional shifts. These examples (21 metagenomes processed in 172 min) clearly illustrate the importance of clubber in the everyday computational biology environment. PMID:28609295

  18. clubber: removing the bioinformatics bottleneck in big data analyses.

    Science.gov (United States)

    Miller, Maximilian; Zhu, Chengsheng; Bromberg, Yana

    2017-06-13

    With the advent of modern day high-throughput technologies, the bottleneck in biological discovery has shifted from the cost of doing experiments to that of analyzing results. clubber is our automated cluster-load balancing system developed for optimizing these "big data" analyses. Its plug-and-play framework encourages re-use of existing solutions for bioinformatics problems. clubber's goals are to reduce computation times and to facilitate use of cluster computing. The first goal is achieved by automating the balance of parallel submissions across available high performance computing (HPC) resources. Notably, the latter can be added on demand, including cloud-based resources, and/or featuring heterogeneous environments. The second goal of making HPCs user-friendly is facilitated by an interactive web interface and a RESTful API, allowing for job monitoring and result retrieval. We used clubber to speed up our pipeline for annotating molecular functionality of metagenomes. Here, we analyzed the Deepwater Horizon oil-spill study data to quantitatively show that the beach sands have not yet entirely recovered. Further, our analysis of the CAMI-challenge data revealed that microbiome taxonomic shifts do not necessarily correlate with functional shifts. These examples (21 metagenomes processed in 172 min) clearly illustrate the importance of clubber in the everyday computational biology environment.

  19. clubber: removing the bioinformatics bottleneck in big data analyses

    Directory of Open Access Journals (Sweden)

    Miller Maximilian

    2017-06-01

    Full Text Available With the advent of modern day high-throughput technologies, the bottleneck in biological discovery has shifted from the cost of doing experiments to that of analyzing results. clubber is our automated cluster-load balancing system developed for optimizing these “big data” analyses. Its plug-and-play framework encourages re-use of existing solutions for bioinformatics problems. clubber’s goals are to reduce computation times and to facilitate use of cluster computing. The first goal is achieved by automating the balance of parallel submissions across available high performance computing (HPC resources. Notably, the latter can be added on demand, including cloud-based resources, and/or featuring heterogeneous environments. The second goal of making HPCs user-friendly is facilitated by an interactive web interface and a RESTful API, allowing for job monitoring and result retrieval. We used clubber to speed up our pipeline for annotating molecular functionality of metagenomes. Here, we analyzed the Deepwater Horizon oil-spill study data to quantitatively show that the beach sands have not yet entirely recovered. Further, our analysis of the CAMI-challenge data revealed that microbiome taxonomic shifts do not necessarily correlate with functional shifts. These examples (21 metagenomes processed in 172 min clearly illustrate the importance of clubber in the everyday computational biology environment.

  20. Easy-to-use application programs for decay heat and delayed neutron calculations on personal computers

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

    1998-03-01

    Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)

  1. Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

    1976-03-01

    The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

  2. Measurements of decay heat and gamma-ray intensity of spent LWR fuel assemblies

    International Nuclear Information System (INIS)

    Vogt, J.; Agrenius, L.; Jansson, P.; Baecklin, A.; Haakansson, A.; Jacobsson, S.

    1999-01-01

    Calorimetric measurements of the decay heat of a number of BWR and PWR fuel assemblies have been performed in the pools at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, CLAB. Gamma-ray measurements, using high-resolution gamma-ray spectroscopy (HRGS), have been carried out on the same fuel assemblies in order to test if it is possible to find a simple and accurate correlation between the 137 CS -intensity and the decay heat for fuel with a cooling time longer than 10-12 years. The results up to now are very promising and may ultimately lead to a qualified method for quick and accurate determination of the decay heat of old fuel by gamma-ray measurements. By means of the gamma spectrum the operator declared data on burnup, cooling time and initial enrichment can be verified as well. CLAB provides a unique opportunity in the world to follow up the decay heat of individual fuel assemblies during several decades to come. The results will be applicable for design and operation of facilities for wet and dry interim storage and subsequent encapsulation for final disposal of the fuel. (author)

  3. Total Absorption Spectroscopy of Fission Fragments Relevant for Reactor Antineutrino Spectra and Decay Heat Calculations

    Directory of Open Access Journals (Sweden)

    Porta A.

    2016-01-01

    Full Text Available Beta decay of fission products is at the origin of decay heat and antineutrino emission in nuclear reactors. Decay heat represents about 7% of the reactor power during operation and strongly impacts reactor safety. Reactor antineutrino detection is used in several fundamental neutrino physics experiments and it can also be used for reactor monitoring and non-proliferation purposes. 92,93Rb are two fission products of importance in reactor antineutrino spectra and decay heat, but their β-decay properties are not well known. New measurements of 92,93Rb β-decay properties have been performed at the IGISOL facility (Jyväskylä, Finland using Total Absorption Spectroscopy (TAS. TAS is complementary to techniques based on Germanium detectors. It implies the use of a calorimeter to measure the total gamma intensity de-exciting each level in the daughter nucleus providing a direct measurement of the beta feeding. In these proceedings we present preliminary results for 93Rb, our measured beta feedings for 92Rb and we show the impact of these results on reactor antineutrino spectra and decay heat calculations.

  4. Beta and gamma decay heat evaluation for the thermal fission of 235U

    International Nuclear Information System (INIS)

    Schenter, G.K.; Schmittroth, F.

    1979-01-01

    Beta and gamma fission product decay heat curves are evaluated for the thermal fission of 235 U. Experimental data that include beta, gamma, and total measurements are combined with summation calculations based on ENDF/B in a consistent evaluation. Least-squares methods are used that take proper account of data uncertainties and correlations. 4 figures, 2 tables

  5. Fission yield covariance generation and uncertainty propagation through fission pulse decay heat calculation

    International Nuclear Information System (INIS)

    Fiorito, L.; Diez, C.J.; Cabellos, O.; Stankovskiy, A.; Van den Eynde, G.; Labeau, P.E.

    2014-01-01

    Highlights: • Fission yield data and uncertainty comparison between major nuclear data libraries. • Fission yield covariance generation through Bayesian technique. • Study of the effect of fission yield correlations on decay heat calculations. • Covariance information contribute to reduce fission pulse decay heat uncertainty. - Abstract: Fission product yields are fundamental parameters in burnup/activation calculations and the impact of their uncertainties was widely studied in the past. Evaluations of these uncertainties were released, still without covariance data. Therefore, the nuclear community expressed the need of full fission yield covariance matrices to be able to produce inventory calculation results that take into account the complete uncertainty data. State-of-the-art fission yield data and methodologies for fission yield covariance generation were researched in this work. Covariance matrices were generated and compared to the original data stored in the library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235 U. Calculations were carried out using different libraries and codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the libraries. The uncertainty quantification was performed first with Monte Carlo sampling and then compared with linear perturbation. Indeed, correlations between fission yields strongly affect the uncertainty of decay heat. Eventually, a sensitivity analysis of fission product yields to fission pulse decay heat was performed in order to provide a full set of the most sensitive nuclides for such a calculation

  6. Experiments and analyses in support of the US ALMR thermal-hydraulic design

    International Nuclear Information System (INIS)

    Hunsbedt, A.

    1993-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) which is based on the modular PRISM concept utilizes passive safety characteristics to simplify the reactor design and enhance its safety performance. The relatively small size of each reactor facilitates the use of strong negative feedback with rising temperature for inherent reactivity control and direct, natural air cooling for decay heat removal. The tall, slender reactor geometry of the ALMR enhances uniformity and stability of internal flow distribution during steady state operation and natural circulation flow during transient conditions. The flow uniformity and low operating pressure and temperature of the reactor contributes to high structural margins. A number of experiments and associated analyses have been performed to evaluate natural convection and thermal-hydraulic phenomena experienced under decay heat removal conditions. This paper summarizes these various efforts as described separately below and presents the main results. (author)

  7. Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

    1997-03-01

    Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

  8. Deposition of aerosols formed by HCDA due to decay heat transport in inner containment atmospheres

    International Nuclear Information System (INIS)

    Vate, J.F. van de

    1976-01-01

    Coupling of decay heat transfer by aerosol-laden inner containment atmospheres with aerosol deposition from such atmospheres leads to useful and simple models for calculation of the time dependence of the aerosol mass concentration. Special attention is given to thermophoretic deposition (dry case) and condensation followed by gravitational deposition (wet case). Attractive features of the models are: 1) coagulation can be omitted and therefore complicated and doubtful calculations on coagulation are avoided, 2) material and particle size of the aerosol are not important for the aerosol decay rate, 3) the aerosol decay rate is related to the decay heat production which is known function of time, and the relevant part of it must be assessed usually for other purposes as well. (orig.) [de

  9. Filtered thermal neutron captured cross sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Pham Ngoc Son; Vuong Huu Tan

    2015-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R ed ) of 420 and neutron flux (Φ th ) of 1.6*10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross sections for nuclide of 51 V, by the activation method relative to the standard reaction 197 Au(n,γ) 198 Au. In addition to the activities of neutron capture cross sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U are introduced in this report. (author)

  10. Decay heat of 235U fission products by beta- and gamma-ray spectrometry

    International Nuclear Information System (INIS)

    Dickens, J.K.; Love, T.A.; McConnell, J.W.; Peelle, R.W.

    1976-09-01

    The fast-rabbit facilities of the ORRR were used to irradiate 1- to 10-μg samples of 235 U for 1, 10, and 100 s. Released power is observed using nuclear spectroscopy to permit separate observations of emitted β and γ spectra in successive time intervals. The spectra were integrated over energy to obtain total decay heat and the β- and γ-ray results are summed together. 10 fig, 2 tables

  11. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within ±10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the 92 Mo(n, 2n) 91g Mo reaction in FENDL, and lack of activation cross section data, e.g., the 138 Ba(n, 2n) 137m Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  12. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within {+-}10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the {sup 92}Mo(n, 2n){sup 91g}Mo reaction in FENDL, and lack of activation cross section data, e.g., the {sup 138}Ba(n, 2n){sup 137m}Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  13. Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

    International Nuclear Information System (INIS)

    Stankunas, Gediminas; Tidikas, Andrius; Pereslavstev, Pavel; Catalán, Juan; García, Raquel; Ogando, Francisco; Fischer, Ulrich

    2016-01-01

    Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

  14. Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Tidikas, Andrius [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Pereslavstev, Pavel [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Catalán, Juan; García, Raquel; Ogando, Francisco [Departamento de Ingeniería Energética, UNED, 28040 Madrid (Spain); Fischer, Ulrich [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-11-01

    Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

  15. Analysis of loss of decay heat removal sequences at Browns Ferry Unit One: Chapter 17

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1983-01-01

    This paper summarizes the Oak Ridge National Laboratory (ORNL) report ''Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis'' (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA), using Unit One of the Browns Ferry Nuclear Plant as the model design. Each unit of this three-unit plant has a maximum authorized power of 3293 MW(t) or 1067 net MW(e). The primary containments are of the Mark I pressure suppression pool type and the three units share a secondary containment of the controlled leakage, elevated release design. Each unit occupies a separate reactor building located in one structure underneath the common refueling floor

  16. Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

    International Nuclear Information System (INIS)

    Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

    2015-01-01

    Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were developed based on the simulation results from two analysis codes, MARS and MAAP. The early initiation and high mass flow rate from the passive safety injection system could guarantee the convergent containment pressure. In addition, in the loss of coolant accident with the failures of safety injection systems, the passive safety injection system and the containment filtered venting system can be used as a new accident management method. (authors)

  17. Alternate method of decay-heat removal in a C-E plant following a SBLOCA

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1986-01-01

    The use of an atmospheric steam-dump procedure to cool and depressurize a Combustion-Engineering plant, Calvert Cliffs-1, following small-break loss-of-coolant accidents with failure of the high-pressure injection system to operate has been investigated. The procedure was effective in depressuizing the primary to the low-pressure injection system operating pressure and design temperature using water supplies from the safety-grade condensate water storage tank only. The procedure was found to be effective even if additional failures occurred. Specifically, low-pressure injection conditions were attained if only a single atmospheric dump valve was available or if the safety-injection tanks (accumulators) were not available

  18. Impact of wind velocity on the performance of the RVACS decay heat removal system

    International Nuclear Information System (INIS)

    Tzanos, C.P.

    1997-01-01

    The impact of wind velocity on the performance of the reactor vessel auxiliary cooling system (RVACS) of an advanced liquid-metal reactor design is analyzed, and design modifications that mitigate adverse wind effects are investigated. In the reference design, the reactor is served by four communicating RVACS stacks, and each stack has two air inlets. In this two-inlet stack design, winds blowing in a direction 90 deg from the axis formed by the two stack inlets result in pressure distributions around the stacks that drastically change the desired airflow pattern in the RVACS. This leads to significantly elevated RVACS air temperatures and significant azimuthal guard vessel temperature variations. For example, a 27 m/s (60 mph) wind leads to an air temperature at the exit of the RVACS heated section that is ∼115 C higher than that under no-wind conditions. The addition of two more inlets per stack, one inlet per stack side, significantly improves RVACS performance. The air temperature at the exit of the heated RVACS section is significantly reduced below that of the two-inlet design, and this temperature decreases as the wind speed increases. An increase in wind speed from 3 to 27 m/s leads to an air temperature change from 186 to 165 C. The azimuthal temperature variation is also improved. At the top of the guard vessel, this variation is reduced from 62.5 to 8.5 C at the low wind speed of 3 m/s and from 85.0 to 30.5 C at the high wind speed of 27 m/s

  19. Validation of intermediate heat and decay heat exchanger model in MARS-LMR with STELLA-1 and JOYO tests

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok; Hong, Jonggan; Yeom, Sujin; Eoh, Jaehyuk [Sodium-cooled Fast Reactor Design Division, Korea Atomic Energy Research Institute (KAERI), 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Jeong, Hae-yong, E-mail: hyjeong@sejong.ac.kr [Department of Nuclear Engineering, Sejong University, 209 Neungdong-ro, Gwangjin-gu, Seoul 143-747 (Korea, Republic of)

    2016-11-15

    Highlights: • The capability of the MARS-LMR for heat transfer through IHX and DHX is evaluated. • Prediction of heat transfer through IHXs and DHXs is essential in the SFR analysis. • Data obtained from the STELLA-1 and the JOYO test are analyzed with the MARS-LMR. • MARS-LMR adopts the Aoki’s correlation for tube side and Graber-Rieger’s for shell. • The performance of the basic models and other available correlations is evaluated. • The current models in MARS-LMR show best prediction for JOYO and STELLA-1 data. - Abstract: The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The decay heat exchangers (DHXs) and the intermediate heat exchangers (IHXs) were designed as a sodium-sodium counter-flow tube bundle type for decay heat removal system (DHRS) and intermediate heat transport system (IHTS), respectively. The IHX and DHX are important components for a heat removal function under normal and accident conditions, respectively. Therefore, sodium heat transfer models for the DHX and IHX heat exchangers were added in MARS-LMR. In order to validate the newly added heat transfer model, experimental data were obtained from the JOYO and STELLA-1 facilities were analyzed. JOYO has two different types of IHXs: type-A (co-axial circular arrangement) and type-B (triangular arrangement). For the code validation, 38 and 39 data points for type A and type B were selected, respectively. A DHX performance test was conducted in STELLA-1, which is the test facility for heat exchangers and primary pump in the PGSFR. The DHX test in STELLA-1 provided eight data points for a code validation. Ten nodes are used in the heat transfer region is used, based on the verification test for the heat transfer models. RMS errors for JOYO IHX type A and type B of 19.1% and 4.3% are obtained

  20. Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

    International Nuclear Information System (INIS)

    Bilodid, Yurii; Fridman, Emil; Shwageraus, E.

    2017-01-01

    This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

  1. Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

    Energy Technology Data Exchange (ETDEWEB)

    Bilodid, Yurii; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety; Kotlyar, D. [Georgia Institute of Technology, Atlanta, GA (United States); Shwageraus, E. [Cambridge Univ. (United Kingdom)

    2017-06-01

    This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

  2. Derivation of decay heat benchmarks for U235 and Pu239 by a least squares fit to measured data

    International Nuclear Information System (INIS)

    Tobias, A.

    1989-05-01

    A least squares technique used by previous authors has been applied to an extended set of available decay heat measurements for both U235 and Pu239 to yield simultaneous fits to the corresponding beta, gamma and total decay heat. The analysis takes account of both systematic and statistical uncertainties, including correlations, via calculations which use covariance matrices constructed for the measured data. The results of the analysis are given in the form of beta, gamma and total decay heat estimates following fission pulses and a range of irradiation times in both U235 and Pu239. These decay heat estimates are considered to form a consistent set of benchmarks for use in the assessment of summation calculations. (author)

  3. Transient Performance of Air-cooled Condensing Heat Exchanger in Long-term Passive Cooling System during Decay Heat Load

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung Jun; Lee, Hee Joon [Kookmin University, Seoul (Korea, Republic of); Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In the event of a 'loss of coolant accident'(LOCA) and a non-LOCA, the secondary passive cooling system would be activated to cool the steam in a condensing heat exchanger that is immersed in an emergency cooldown tank (ECT). Currently, the capacities of these ECTs are designed to be sufficient to remove the sensible and residual heat from the reactor coolant system for 72 hours after the occurrence of an accident. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. Therefore, the tank should be refilled regularly from an auxiliary water supply system when the system is used for more than 72 hours. Otherwise, the system would fail to dissipate heat from the condensing heat exchanger due to the loss of the cooling water. Ultimately, the functionality of the passive cooling system would be seriously compromised. As a passive means of overcoming the water depletion in the tank, Kim et al. applied for a Korean patent covering the concept of a long-term passive cooling system for an ECT even after 72 hours. This study presents transient performance of ECT with installing air-cooled condensing heat exchanger under decay heat load. The cooling capacity of an air-cooled condensing heat exchanger was evaluated to determine its practicality.

  4. An evaluation of nodalization/decay heat/ volatile fission product release models in ISAAC code

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong; Kim, Dong Ha

    2003-03-01

    An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. In case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option shows mitigated conservative results.

  5. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  6. Beta-decay and decay heat. Summary report of consultants' meeting

    International Nuclear Information System (INIS)

    Nicols, A.L.

    2006-01-01

    Experts on decay data and decay heat calculations participated in a Consultants' Meeting organized at IAEA Headquarters on 12-14 December 2005. Debate focused on the validation of decay heat calculations as a function of cooling time for fuel irradiated in power reactors through comparisons with experimental benchmark data. Both the current understanding and quantification of mean beta and gamma decay energies were reviewed with respect to measurements and the Gross Theory of Beta Decay. Particular emphasis was placed on the known development of total absorption gamma-ray spectroscopy (TAGS), and detailed discussions took place to formulate the measurement requirements for mean beta and gamma data of individual radionuclides. This meeting was organized in cooperation with the OECD/NEA Working Party for Evaluation and Cooperation (WPEC). Proposals and recommendations were made to resolve particular difficulties, and an initial list of fission products was produced for TAGS studies. The discussions, conclusions and recommendations of the meeting are briefly described in this report. (author)

  7. Impact of the total absorption gamma-ray spectroscopy on FP decay heat calculations

    International Nuclear Information System (INIS)

    Yoshida, Tadashi; Tachibana, Takahiro; Katakura, Jun-ichi

    2004-01-01

    We calculated the average β- and γ-ray energies, E β and E γ , for 44 short-lived isotopes of Rb, Sr, Y, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm and Eu from the data by Greenwood et al, who measured the β-feed in the decay of these nuclides using the total absorption γ-ray spectrometer. These E β and E γ were incorporated into the decay files from JENDL, JEF2.2 and ENDF-B/VI, and the decay heats were calculated. The results were compared with the integral measurements by the University of Tokyo, ORNL and Lowell. In the case of JENDL, where the correction for the so-called Pandemonium effect is applied on the basis of the gross theory, the very good agreement is no longer maintained. The γ-ray component is overestimated in the cooling time range from 3 to 300 seconds, suggesting a kind of an over-correction as for the Pandemonium effect. We have to evaluate both the applicability of the TAGS results and the correction method itself in order to generate a more consistent data basis for decay heat summation calculations. (author)

  8. Decay heat and gamma dose-rate prediction capability in spent LWR fuel

    International Nuclear Information System (INIS)

    Neely, G.J.; Schmittroth, F.

    1982-08-01

    The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

  9. Filtered thermal neutron captured cross-sections measurements and decay heat calculations

    International Nuclear Information System (INIS)

    Son, Pham Ngoc; Tan, Vuong Huu

    2014-01-01

    Recently, a pure thermal neutron beam has been developed for neutron capture measurements based on the horizontal channel No.2 of the research reactor at the Nuclear Research Institute, Dalat. The original reactor neutron spectrum is transmitted through an optimal composition of Bi and Si single crystals for delivering a thermal neutron beam with Cadmium ratio (R cd ) of 420 and neutron flux (Φ th ) of 1.6x10 6 n/cm 2 .s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51 V, 55 Mn, 180 Hf and 186 W by the activation method relative to the standard reaction 197 Au(n,g) 198 Au. In addition to the activities of neutron capture cross-sections measurements, the study on nuclear decay heat calculations has been also considered to be developed at the Institute. Some results on calculation procedure and decay heat values calculated with update nuclear database for 235 U, 238 U, 239 Pu and 232 Th are introduced in this report. (author)

  10. FAKIR: a user-friendly standard for decay heat and activity calculation of LWR fuel

    International Nuclear Information System (INIS)

    Pretesacque, P.; Nimal, J.C.; Huynh, T.D.; Zachar, M.

    1993-01-01

    The shipping casks owned by the transporters and the unloading and storage facilities are subjected by their design safety report to decay heat and activity limits. It is the responsibility of the consignor or the consignee to check the compliance of the fuel assemblies to the shipped or stored with regard to these limiting safety parameters. Considering the diversity of the parties involved in the transport and storage cycle, a standardization has become necessary. This has been achieved by the FAKIR code. The FAKIR development started in 1984 in collaboration between COGEMA, CEA-SERMA and NTL. Its main specifications were to be a user-friendly code, to use the contractual data given in the COGEMA transport and reprocessing sheet 1 as input, and to over-estimate decay heat and activity. Originally based on computerizable standards such as ANSI or USNRC, the FAKIR equations and data libraries are now based on the fully qualified PEPIN/APOLLO calculation codes. FAKIR is applicable to all patterns of irradiation histories, with burn up from 1000 MWd/TeU to 70.000 MWd/TeU and cooling times from 1 second to 100 years. (J.P.N.)

  11. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    Directory of Open Access Journals (Sweden)

    Stankunas Gediminas

    2017-01-01

    Full Text Available This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

  12. A brief description of ENDF/B-IV format data for inventory and decay heating calculations

    International Nuclear Information System (INIS)

    Tobias, A.

    1976-07-01

    In recent years there has been considerable effort directed towards establishing an international standard format for computerised nuclear data files. At the recent conference on Fission Product Nuclear Data (Bologna, 1973) it was agreed that the ENDF/B format, with certain modifications, be adopted as the standard format for the exchange of such data. A brief description of the basic ENDF/B-IV format of nuclear data files for inventory and decay heat calculations is presented. Although data exchange and inter-comparison will be simple for all files using this format, the data is not generally in a form which can be used directly by inventory codes. One solution to this problem may be for each code to possess a 'translating' routine for rearranging the data into its own format. (author)

  13. The Collection of Event Data and its Relevance to the Optimisation of Decay Heat Rejection Systems

    International Nuclear Information System (INIS)

    Roughley, R.; Jones, N.

    1975-01-01

    The precision with which the reliability of DHR (Decay Heat Rejection) systems for nuclear reactors can be predicted depends not only upon model representation but also on the accuracy of the data used. In the preliminary design stages when models are being used to arrive at major engineering decisions in relation to plant configuration, the best the designer can do is use the data available at the time. With the present state of the art it is acknowledged that some degree of judgement will have to be exercised particularly for plant involving sodium technology where a large amount of operational experience has not yet been generated. This paper reviews the current efforts being deployed in the acquisition of field data relevant to DHR systems so that improvements in reliability predictions may be realised

  14. Beta decay heat following U-235, U-238 and Pu-239 neutron fission

    Science.gov (United States)

    Li, Shengjie

    1997-09-01

    This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

  15. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and γ ray spectrum. FPGS90

    International Nuclear Information System (INIS)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting γ ray and β ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted γ ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library 'JNDC Nuclear Data Library of Fission Products - second version -', which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author)

  16. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and {gamma} ray spectrum. FPGS90

    Energy Technology Data Exchange (ETDEWEB)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).

  17. Regulatory analyses for severe accident issues: an example

    International Nuclear Information System (INIS)

    Burke, R.P.; Strip, D.R.; Aldrich, D.C.

    1984-09-01

    This report presents the results of an effort to develop a regulatory analysis methodology and presentation format to provide information for regulatory decision-making related to severe accident issues. Insights and conclusions gained from an example analysis are presented. The example analysis draws upon information generated in several previous and current NRC research programs (the Severe Accident Risk Reduction Program (SARRP), Accident Sequence Evaluation Program (ASEP), Value-Impact Handbook, Economic Risk Analyses, and studies of Vented Containment Systems and Alternative Decay Heat Removal Systems) to perform preliminary value-impact analyses on the installation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The results presented in this report are first-cut estimates, and are presented only for illustrative purposes in the context of this document. This study should serve to focus discussion on issues relating to the type of information, the appropriate level of detail, and the presentation format which would make a regulatory analysis most useful in the decisionmaking process

  18. Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

    International Nuclear Information System (INIS)

    Sterbentz, J.W.

    1997-03-01

    Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups

  19. Integral decay-heat measurements and comparisons to ENDF/B--IV and V

    International Nuclear Information System (INIS)

    England, T.R.; Schenter, R.E.; Schmittroth, F.

    Results from recent integral decay-power experiments are presented and compared with summation calculations. The experiments include the decay power following thermal fission of 233 U, 235 U, and 239 Pu. The summation calculations use ENDF/B-IV decay data and yields from Versions IV and V. Limited comparisons of experimental β and γ spectra with summation calculations using ENDF/B-IV are included. Generalized least-squares methods are applied to the recent 235 U and 239 Pu decay-power experiments and summation calculations to arrive at evaluated values and uncertainties. Results for 235 U imply uncertainties less than 2% (1 sigma) for the ''infinite'' exposure case for all cooling times greater than 10 seconds. The uncertainties for 239 Pu are larger. Accurate analytical representations of the decay power are presented for 235 , 238 U, and 239 Pu for use in light-water reactors and as the nominal values in the new ANS 5.1 Draft Standard (1978). Comparisons of the nominal values with ENDF/B-IV and the 1973 ANS Draft Standard in current use are included. Gas content, important to decay-heat experiments, and absorption effects on decay power are reviewed. 37 figures, 8 tables

  20. Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1998-01-01

    Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

  1. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    OpenAIRE

    Stankunas Gediminas; Tidikas Andrius

    2017-01-01

    This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-c...

  2. Technical support for a proposed decay heat guide using SAS2H/ORIGEN-S data

    International Nuclear Information System (INIS)

    Hermann, O.W.; Parks, C.V.; Renier, J.P.

    1994-09-01

    Major revisions are proposed to the current US Nuclear Regulatory Commission decay heat rate guide entitled ''Regulatory Guide 3.54, Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation,'' using a new data base produced by the SAS2H analysis sequence of the SCALE-4 system. The data base for the proposed guide revision has been significantly improved by increasing the number and range of parameters that generally characterize pressurized-water-reactor (PWR) and boiling-water-reactor (BWR) spent fuel assemblies. Using generic PWR and BWR assembly models, calculations were performed with each model for six different burnups at each of three separate specific powers to produce heat rates at 20 cooling times in the range of 1 to 110 y. The proposed procedure specifies proper interpolation formulae for the tabulated heat generation rates. Adjustment formulae for the interpolated values are provided to account for differences in initial 235 U enrichment and changes in the specific power of a cycle from the average value. Finally, safety factor formulae were derived as a function of burnup, cooling time, and type of reactor. The proposed guide revision was designed to be easier to use. Also, the complete data base and guide procedure is incorporated into an interactive code called LWRARC which can be executed on a personal computer. The report shows adequate comparisons of heat rates computed by SAS2H/ORIGEN-S and measurements for 10 BWR and 10 PWR fuel assemblies. The average differences of the computed minus the measured heat rates of fuel assemblies were -07 ± 2.6% for the BWR and 1.5 ± 1.3% for the PWR. In addition, a detailed analysis of the proposed procedure indicated the method and equations to be valid

  3. Decay heat from products of 235U thermal fission by fast-response boil-off calorimetry

    International Nuclear Information System (INIS)

    Yarnell, J.L.; Bendt, P.J.

    1977-09-01

    A cryogenic boil-off calorimeter was used to measure the decay heat from the products of thermal-neutron-induced fission of 235 U. Data are presented for cooling times between 10 and 10 5 s following a 2 x 10 4 s irradiation at constant thermal-neutron flux. The experimental uncertainty (1 sigma) in these measurements was approximately 2 percent, except at the shortest cooling times where it rose to approximately 4 percent. The beta and gamma energy from an irradiated 235 U sample was absorbed in a thermally isolated 52-kg copper block that was held at 4 K by an internal liquid helium reservoir. The absorbed energy evaporated liquid helium from the reservoir and a hot-film anemometer flowmeter recorded the evolution rate of the boil-off gas. The decay heat was calculated from the gas-flow rate using the heat of vaporization of helium. The calorimeter had a thermal time constant of 0.85 s. The energy loss caused by gamma leakage from the absorber was less than or equal to 3 percent; a correction was made by Monte Carlo calculations based on experimentally determined gamma spectra. The data agree within the combined uncertainties with summation calculations using the ENDF/B-IV data base. The experimental data were combined with summation calculations to give the decay heat for infinite (10 13 s) irradiation

  4. Studies of decay heat removal by natural convection using the SONACO sodium-cooled 37-pin bundle

    International Nuclear Information System (INIS)

    Wydler, P.; Dury, T.V.; Hudina, M.; Weissenfluh, T. von; Sigg, B.; Dutton, P.

    1986-01-01

    Natural convection measurements in an electrically heated sodium-cooled rod bundle are being performed with the aim of contributing to a better understanding of natural convection effects in subassemblies with stagnant sodium and providing data for code validation. Measurements include temperature distributions in the bundle for different cooling configurations which simulate heat transfer to the intersubassembly gap and neighbouring subassemblies and possible thermosyphonic interaction between a subassembly and the reactor plenum above. Conditions for which stable natural convection patterns exist are identified, and results are compared with predictions of different computer codes of the porous-medium type. (author)

  5. A generic study of phenomena affecting two-phase mixing in BWR suppression pools during passive decay-heat removal

    International Nuclear Information System (INIS)

    Smith, B. L.; Milelli, M.; Shepel, S.; Lakehal, D.

    2003-01-01

    The paper describes some advancements made in the use of two-phase Computational Fluid Dynamics (CFD), sometimes called Computational Multi-Fluid Dynamics (CMFD), techniques in simulating the phenomena occurring in pressure suppression pools in Advanced Boiling Water Reactors which utilise passive containment cooling systems. An interface tracking procedure based on the Level-Set approach has been implemented into a commercial CFD code with the specific purpose of providing a computational environment for the development of suitable models to describe the inter-phase mass and energy transport processes which would take place when a large gas bubble is discharged into a pool. Details of the implementation and validation of the tracking algorithm are described, together with some illustrations of how the method is utilised. The paper also reports on the progress which is being made in the use of Large-Eddy Simulation (LES) to describe turbulent mixing in such plumes. The research efforts are aimed at ultimately combining the approaches to develop a mechanistic tool for fully describing the pool dynamics and steam condensation phenomena

  6. Uncertainty of decay heat calculations originating from errors in the nuclear data and the yields of individual fission products

    International Nuclear Information System (INIS)

    Rudstam, G.

    1979-01-01

    The calculation of the abundance pattern of the fission products with due account taken of feeding from the fission of 235 U, 238 U, and 239 Pu, from the decay of parent nuclei, from neutron capture, and from delayed-neutron emission is described. By means of the abundances and the average beta and gamma energies the decay heat in nuclear fuel is evaluated along with its error derived from the uncertainties of fission yields and nuclear properties of the inddividual fission products. (author)

  7. Study on concrete cask for practical use. Heat removal test under normal condition

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

    2005-01-01

    In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

  8. Use of financial and economic analyses by federal forest managers for woody biomass removal

    Science.gov (United States)

    Todd A. Morgan; Jason P. Brandt; John D. Baldridge; Dan R. Loeffler

    2011-01-01

    This study was sponsored by the Joint Fire Science Program to understand and enhance the ability of federal land managers to address financial and economic (F&E) aspects of woody biomass removal as a component of fire hazard reduction. Focus groups were conducted with nearly 100 federal land managers throughout the western United States. Several issues and...

  9. Localized dryout: An approach for managing the thermal hydrologi-cal effects of decay heat at Yucca Mountain

    International Nuclear Information System (INIS)

    Buscheck, T. A.; Nitao, J.J.; Ramspott, L.D.

    1995-11-01

    For a nuclear waste repository in the unsaturated zone at Yucca Mountain, there are two thermal loading approaches to using decay heat constructively -- that is, to substantially reduce relative humidity and liquid flow near waste packages for a considerable time, and thereby limit waste package degradation and radionuclide dissolution and release. ''Extended dryout'' achieves these effects with a thermal load high enough to generate large-scale (coalesced) rock dryout. ''Localized dryout''(which uses wide drift spacing and a thermal load too low for coalesced dryout) achieves them by maintaining a large temperature difference between the waste package and drift wall; this is done with close waste package spacing (generating a high line-heat load) and/or low-thermal-conductivity backfill in the drift. Backfill can greatly reduce relative humidity on the waste package in both the localized and extended dryout approaches. Besides using decay heat constructively, localized dryout reduces the possibility that far-field temperature rise and condensate buildup above the drifts might adversely affect waste isolation

  10. Development of a water boil-off spent-fuel calorimeter system. [To measure decay heat generation rate

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Shupe, J.W. Jr.

    1981-05-01

    A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW.

  11. Chemical and microstructural analyses for heavy metals removal from water media by ceramic membrane filtration.

    Science.gov (United States)

    Ali, Asmaa; Ahmed, Abdelkader; Gad, Ali

    2017-01-01

    This study aims to investigate the ability of low cost ceramic membrane filtration in removing three common heavy metals namely; Pb 2+ , Cu 2+ , and Cd 2+ from water media. The work includes manufacturing ceramic membranes with dimensions of 15 by 15 cm and 2 cm thickness. The membranes were made from low cost materials of local clay mixed with different sawdust percentages of 0.5%, 2.0%, and 5.0%. The used clay was characterized by X-ray diffraction (XRD) and X-ray fluorescence analysis. Aqueous solutions of heavy metals were prepared in the laboratory and filtered through the ceramic membranes. The influence of the main parameters such as pH, initial driving pressure head, and concentration of heavy metals on their removal efficiency by ceramic membranes was investigated. Water samples were collected before and after the filtration process and their heavy metal concentrations were determined by chemical analysis. Moreover, a microstructural analysis using scanning electronic microscope (SEM) was performed on ceramic membranes before and after the filtration process. The chemical analysis results showed high removal efficiency up to 99% for the concerned heavy metals. SEM images approved these results by showing adsorbed metal ions on sides of the internal pores of the ceramic membranes.

  12. Decay heat and activity of the structural materials of the fuel and blanket assemblies of the second and third core of KNK II

    International Nuclear Information System (INIS)

    Winterhagen, D.

    1986-06-01

    The decay heat and activity caused by structural materials have been calculated for the fuel assemblies of KNK II (second and third core) with a residence time of 720 equivalent full-power days (efpd) and the blanket assemblies with 1880 efpd. The values are given for the different zones of the assemblies (head, active zone, fission gas plenum, foot and stellite area) for decay times from 1 to 20 years. For decay times beyond 2 years more than 80 % of the decay heat are caused by the Co60-decay, more than 60 % of which result from the stellite in the foot area [de

  13. Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Hu, Rui; Lisowski, Darius; Kraus, Adam

    2016-04-17

    The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.

  14. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Science.gov (United States)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  15. Influence of high burnup on the decay heat power of spent fuel at long-term storage

    International Nuclear Information System (INIS)

    Bergelson, B.; Gerasimov, A.; Tikhomirov, G.

    2005-01-01

    Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)

  16. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Directory of Open Access Journals (Sweden)

    Ternovykh Mikhail

    2017-01-01

    Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  17. ORIGEN2.1 Cycle Specific Calculation of Krsko Nuclear Power Plant Decay Heat and Core Inventory

    International Nuclear Information System (INIS)

    Vukovic, J.; Grgic, D.; Konjarek, D.

    2010-01-01

    This paper presents ORIGEN2.1 computer code calculation of Krsko Nuclear Power Plant core for Cycle 24. The isotopic inventory, core activity and decay heat are calculated in one run for the entire core using explicit depletion and decay of each fuel assembly. Separate pre-ori application which was developed is utilized to prepare corresponding ORIGEN2.1 inputs. This application uses information on core loading pattern to determine fuel assembly specific depletion history using 3D burnup which is obtained from related PARCS computer code calculation. That way both detailed single assembly calculations as well as whole core inventory calculations are possible. Because of the immense output of the ORIGEN2.1, another application called post-ori is used to retrieve and plot any calculated property on the basis of nuclide, element, summary isotope or group of elements for activation products, actinides and fission products segments. As one additional possibility, with the post-ori application it is able to calculate radiotoxicity from calculated ORIGEN2.1 inventory. The results which are obtained using the calculation model of ORIGEN2.1 computer code are successfully compared against corresponding ORIGEN-S computer code results.(author).

  18. On Error Analysis of ORIGEN Decay Data Library Based on ENDF/B-VII.1 via Decay Heat Estimation after a Fission Event

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The method is strongly dependent on the available nuclear structure data, i.e., fission product yield data and decay data. Consequently, the improvements in the nuclear structure data could have guaranteed more reliable decay heat estimation for short cooling times after fission. The SCALE-6.1.3 code package includes the ENDF/B-VII.0-based fission product yield data and ENDF/B-VII.1-based decay data libraries for the ORIGEN-S code. The generation and validation of the new ORIGEN-S yield data libraries based on the recently available fission product yield data such as ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0 have been presented in the previous study. According to the study, the yield data library in the SCALE-6.1.3 could be regarded as the latest one because it resulted in almost the same outcomes as the ENDF/B-VII.1. A research project on the production of the nuclear structure data for decay heat estimation of nuclear fuel has been carried out in Korea Atomic Energy Research Institute (KAERI). The data errors contained in the ORIGEN-S decay data library of SCALE-6.1.3 have been clearly identified by their changing variables. Also, the impacts of the decay data errors have been analyzed by estimating the decay heats for the fission product nuclides and their daughters after {sup 235}U thermal-neutron fission. Although the impacts of decay data errors are quite small, it reminds us the possible importance of decay data when estimating the decay heat for short cooling times after a fission event.

  19. Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

    Since its inception in October 1980, the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory (ORNL) has completed four studies including Station Blackout, Scram Discharge Volume Break, Loss of Decay Heat Removal, and Loss of Injection accident sequences for the Browns Ferry Nuclear Plant. The accident analyses incorporated in a SASA study provide much greater detail than that practically achievable in a Probabilistic Risk Assessment (PRA). When applied to the candidate dominant accident sequences identified by a PRA, the detailed SASA results determine if factors neglected by the PRA would have a significant effect on the order of dominant sequences. Ongoing SASA work at ORNL involves the analysis of Anticipated Transients Without Scram (ATWS) sequences for Browns Ferry

  20. Feasibility of passive heat removal systems

    Energy Technology Data Exchange (ETDEWEB)

    Ashurko, Yu M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1996-12-01

    This paper presents a review of decay heat removal systems (DHRSs) used in liquid metal-cooled fast reactors (LMFRs). Advantages and the disadvantages of these DHRSs, extent of their passivity and prospects for their use in advanced fast reactor projects are analyzed. Methods of extending the limitations on the employment of individual systems, allowing enhancement in their effectiveness as safety systems and assuring their total passivity are described. (author). 10 refs, 10 figs.

  1. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  2. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  3. Total absorption gamma-ray spectroscopy (TAGS): Current status of measurement programmes for decay heat calculations and other applications. Summary report of consultants' meeting

    International Nuclear Information System (INIS)

    Nichols, A.L.; Nordborg, C.

    2009-02-01

    A Consultants' Meeting on 'Total Absorption Gamma-ray Spectroscopy (TAGS)' was held on 27-28 January 2009 at the IAEA Headquarters, Vienna, Austria. All presentations, discussions and recommendations of this meeting are contained within this report. The purpose of the meeting was to report and discuss progress and plans to measure total gamma-ray spectra in order to derive mean beta and gamma decay data for decay heat calculations and other applications. This form of review had been recommended by contributors to Subgroup 25 of the OECD-NEA Working Party on International Evaluation Cooperation of the Nuclear Science Committee, for implementation in 2008/09. Hence, relevant specialists were invited to discuss their recently performed and planned TAGS studies, along with experimentalists proposing to assemble and operate such dedicated facilities. Knowledge and quantification of antineutrino spectra is believed to be a significant asset in the non-invasive monitoring of reactor operations and possible application in safeguards, as well as fundamental in the study of neutrino oscillations - these data needs were also debated in terms of appropriate TAGS measurements. A re-assessment of the current request list for TAGS studies is merited and was undertaken in the context of decay heat calculations, and agreement was reached to extend these requirements to the derivation of antineutrino spectra. (author)

  4. Study on the impact of transition from 3-batch to 4-batch loading at Loviisa NPP on the long-term decay heat and activity inventory

    Energy Technology Data Exchange (ETDEWEB)

    Lahtinen, Tuukka [Fortum Power and Heat Ltd., Fortum (Finland)

    2017-09-15

    The fuel economy of Loviisa NPP was improved by implementing a transition from 3-batch to 4-batch loading scheme between 2009 and 2013. Equilibrium cycle length as well as all process parameters were retained unchanged while the increase of fuel enrichment enabled to reduce the annual reload batch size from 102 to 84 assemblies. The fuel cycle transition obviously had an effect on the long-term decay heat and activity inventory. However, due to simultaneous change in several quantities the net effect over the relevant cooling time region is not self-evident. In this study the effect is analyzed properly, i. e. applying consistent calculation models and detailed description of assembly-wise irradiation histories. The study concludes that for the cooling time, foreseen typical prior to encapsulation of assemblies, the decay heat of discharge batch increases 2 - 3%. It is also concluded that, in order to maintain 100% filling degree of final disposal canisters, the cooling time prior to encapsulation needs to be prolonged by 10 - 15 years.

  5. Preliminary analyses of AP600 using RELAP5

    International Nuclear Information System (INIS)

    Modro, S.M.; Beelman, R.J.; Fisher, J.E.

    1991-01-01

    This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

  6. Dynamic informational system for control and monitoring the tritium removal pilot plant with data transfer and process analyses

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

    2005-01-01

    The dynamic informational system with datalogging and supervisory control module includes a motion control module and is a new conception used in tritium removal installation with isotopic exchange and cryogenic distillation. The control system includes an event-driven engine that maintains a real-time database, logs historical data, processes alarm information, and communicates with I/O devices. Also, it displays the operator interfaces and performs tasks that are defined for advanced control algorithms, supervisory control, analysis, and display with data transfer from data acquisition room to the control room. By using the parameters, we compute the deuterium and tritium concentration, respectively, of the liquid at the inlet of the isotopic exchange column and, consequently, we can compute at the outlet of the column, the tritium concentration in the water vapors. (authors)

  7. Probabilistic reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

    International Nuclear Information System (INIS)

    Schilling, R.

    1984-01-01

    Requirements made by Federal German licensing authorities called for the analysis of the second-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems are identified, remedial measures of a system-specific and test-strategic nature are presented and their contribution to improving system availability is quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

  8. Reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

    International Nuclear Information System (INIS)

    Schilling, R.

    1983-01-01

    Requirements made by Federal German licensing authorities called for the analysis of the secondary-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems identified, remedial measures of a system-specific and test-strategic nature presented and their contribution to improving system availability quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

  9. A breakthrough biosorbent in removing heavy metals: Equilibrium, kinetic, thermodynamic and mechanism analyses in a lab-scale study

    Energy Technology Data Exchange (ETDEWEB)

    Abdolali, Atefeh [Centre for Technology in Water and Wastewater, School of Civil and Environmental Engineering, University of Technology Sydney, Broadway, NSW 2007 (Australia); Ngo, Huu Hao, E-mail: h.ngo@uts.edu.au [Centre for Technology in Water and Wastewater, School of Civil and Environmental Engineering, University of Technology Sydney, Broadway, NSW 2007 (Australia); Guo, Wenshan [Centre for Technology in Water and Wastewater, School of Civil and Environmental Engineering, University of Technology Sydney, Broadway, NSW 2007 (Australia); Lu, Shaoyong [Chinese Research Academy of Environmental Science, Beijing 100012 (China); Chen, Shiao-Shing; Nguyen, Nguyen Cong [Institute of Environmental Engineering and Management, National Taipei University of Technology, No. 1, Sec. 3, Chung-Hsiao E. Rd, Taipei 106, Taiwan (China); Zhang, Xinbo [Department of Environmental and Municipal Engineering, Tianjin Key Laboratory of Aquatic Science and Technology, Tianjin Chengjian University, Jinjing Road 26, Tianjin 300384 (China); Wang, Jie; Wu, Yun [School of Environmental and Chemical Engineering, Tianjin Polytechnic University, Tianjin 300387 (China)

    2016-01-15

    A breakthrough biosorbent namely multi-metal binding biosorbent (MMBB) made from a combination of tea wastes, maple leaves and mandarin peels, was prepared to evaluate their biosorptive potential for removal of Cd(II), Cu(II), Pb(II) and Zn(II) from multi-metal aqueous solutions. FTIR and SEM were conducted, before and after biosorption, to explore the intensity and position of the available functional groups and changes in adsorbent surface morphology. Carboxylic, hydroxyl and amine groups were found to be the principal functional groups for the sorption of metals. MMBB exhibited best performance at pH 5.5 with maximum sorption capacities of 31.73, 41.06, 76.25 and 26.63 mg/g for Cd(II), Cu(II), Pb(II) and Zn(II), respectively. Pseudo-first and pseudo-second-order models represented the kinetic experimental data in different initial metal concentrations very well. Among two-parameter adsorption isotherm models, the Langmuir equation gave a better fit of the equilibrium data. For Cu(II) and Zn(II), the Khan isotherm describes better biosorption conditions while for Cd(II) and Pb(II), the Sips model was found to provide the best correlation of the biosorption equilibrium data. The calculated thermodynamic parameters indicated feasible, spontaneous and exothermic biosorption process. Overall, this novel MMBB can effectively be utilized as an adsorbent to remove heavy metal ions from aqueous solutions. - Highlights: • A novel multi-metal binding biosorbent (MMBB) was studied. • The biosorption of Cd{sup 2+}, Cu{sup 2+}, Pb{sup 2+} and Zn{sup 2+} on MMBB was evaluated. • Hydroxyl, carbonyl and amine groups are involved in metal binding of MMBB. • Equilibrium data were presented and the best fitting models were identified. • The obtained results recommend this MMBB as potentially low-cost biosorbent.

  10. On-site phytoremediation applicability assessment in Alur Ilmu, Universiti Kebangsaan Malaysia based on spatial and pollution removal analyses.

    Science.gov (United States)

    Mahmud, Mohd Hafiyyan; Lee, Khai Ern; Goh, Thian Lai

    2017-10-01

    The present paper aims to assess the phytoremediation performance based on pollution removal efficiency of the highly polluted region of Alur Ilmu urban river for its applicability of on-site treatment. Thirteen stations along Alur Ilmu were selected to produce thematic maps through spatial distribution analysis based on six water quality parameters of Malaysia's Water Quality Index (WQI) for dry and raining seasons. The maps generated were used to identify the highly polluted region for phytoremediation applicability assessment. Four free-floating plants were tested in treating water samples from the highly polluted region under three different conditions, namely controlled, aerated and normal treatments. The selected free-floating plants were water hyacinth (Eichhornia crassipes), water lettuce (Pistia stratiotes), rose water lettuce (Pistia sp.) and pennywort (Centella asiatica). The results showed that Alur Ilmu was more polluted during dry season compared to raining season based on the water quality analysis. During dry season, four parameters were marked as polluted along Alur Ilmu, namely dissolve oxygen (DO), 4.72 mg/L (class III); ammoniacal nitrogen (NH 3 -N), 0.85 mg/L (class IV); total suspended solid (TSS), 402 mg/L (class V) and biological oxygen demand (BOD), 3.89 mg/L (class III), whereas, two parameters were classed as polluted during raining season, namely total suspended solid (TSS), 571 mg/L (class V) and biological oxygen demand (BOD), 4.01 mg/L (class III). The thematic maps generated from spatial distribution analysis using Kriging gridding method showed that the highly polluted region was recorded at station AL 5. Hence, water samples were taken from this station for pollution removal analysis. All the free-floating plants were able to reduce TSS and COD in less than 14 days. However, water hyacinth showed the least detrimental effect from the phytoremediation process compared to other free-floating plants, thus made it a suitable

  11. Safety design analyses of Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  12. Design and analysis of a new passive residual heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Xing [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Peng, Minjun, E-mail: heupmj@163.com [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Yuan, Xiao [Guangxi Fangchenggang Nuclear Power Co., Ltd (China); Xia, Genglei [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China)

    2016-07-15

    Highlights: • An air cooling passive residual heat removal System (PRHRs) is designed. • Using RELAP5/MOD3.4 code to analyze the operation characteristics of the PRHRs. • Noncondensable gas is used to simulate the hydrodynamic behavior in the air cooling tower. • The natural circulations could respectively establish in the primary circuit and the PRHRs circuit. • The PRHRs could remove the residual heat effectively. - Abstract: The inherent safety functions will mitigate the consequences of the accidents, and it can be accomplished through the passive safety systems which employed in the typical pressurized water reactor (PWR). In this paper, a new passive residual heat removal system (PRHRS) is designed for a typical nuclear power plant. PRHRS consists of a steam generator (SG), a cooling tank with two groups of cooling pipes, an air-cooling heat exchanger (AHX), an air-cooling tower, corresponding pipes and valves. The cooling tank which works as an intermediate buffer device is used to transfer the core decay heat to the AHX, and then the core decay heat will be removed to the atmosphere finally. The RELAP5/MOD3.4 code is used to analyze the operation characteristics of PRHRS and the primary loop system. It shows PRHRS could remove the decay heat from the primary loop effectively, and the natural circulations can be established in the primary circuit and the PRHRS circuit respectively. Furthermore, the sensitivity study has also been done to research the effect of various factors on the heat removal capacity.

  13. Nuclear energy waste: space transportation and removal

    International Nuclear Information System (INIS)

    Burns, R.E.

    1975-12-01

    A method for utilizing the decay heat of actinide wastes to power an electric thrust vehicle is proposed. The vehicle, launched by shuttle to earth orbit and to earth escape by a tug, obtains electrical power from the actinide waste heat by thermionic converters. The heavy gamma ray and neutron shielding which is necessary as a safety feature is removed in orbit and returned to earth for reuse. The problems associated with safety are dealt with in depth. A method for eliminating fission wastes via chemical propulsion is briefly discussed

  14. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  15. Beta and gamma decay heat measurements between 0.1s--50,000s for neutron fission of 235U, 238U and 239Pu. Final report, June 1, 1992--December 31, 1996

    International Nuclear Information System (INIS)

    Schier, W.A.; Couchell, G.P.

    1996-01-01

    This is a final reporting on the composition of separate beta and gamma decay heat measurements following neutron fission of 235 U and 238 U and 239 Pu and on cumulative and independent yield measurements of fission products of 235 U and 238 U. What made these studies unique was the very short time of 0.1 s after fission that could be achieved by incorporating the helium jet and tape transport system as the technique for transporting fission fragments from the neutron environment of the fission chamber to the low-background environment of the counting area. This capability allowed for the first time decay heat measurements to extend nearly two decades lower on the logarithmic delay time scale, a region where no comprehensive aggregate decay heat measurements had extended to. This short delay time capability also allowed the measurement of individual fission products with half lives as short as 0.2s. The purpose of such studies was to provide tests both at the aggregate level and at the individual nuclide level of the nation's evaluated nuclear data file associated with fission, ENDF/B-VI. The results of these tests are in general quite encouraging indicating this data base generally predicts correctly the aggregate beta and aggregate gamma decay heat as a function of delay time for 235 U, 238 U and 239 Pu. Agreement with the measured individual nuclide cumulative and independent yields for fission products of 235 U and 238 U was also quite good although the present measurements suggest needed improvements in several individual cases

  16. Sodium removal of fuel elements by vacuum distillation

    International Nuclear Information System (INIS)

    Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

    1978-01-01

    Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

  17. Analyses of fecal and hair glucocorticoids to evaluate short- and long-term stress and recovery of Asiatic black bears (Ursus thibetanus) removed from bile farms in China.

    Science.gov (United States)

    Malcolm, K D; McShea, W J; Van Deelen, T R; Bacon, H J; Liu, F; Putman, S; Zhu, X; Brown, J L

    2013-05-01

    Demand for traditional Chinese medicines has given rise to the practice of maintaining Asiatic black bears (Ursus thibetanus) in captivity to harvest bile. We evaluated hypothalamic-pituitary-adrenal (HPA) activity in Asiatic black bears on a bile farm in China by measuring cortisol in hair. We also monitored hair and fecal glucocorticoid metabolites as bears acclimated to improved husbandry at the Animals Asia Foundation China Bear Rescue Center (CBRC) after removal from other bile farms. Fecal samples were collected twice weekly for ~1 year, and hair was obtained from bears upon arrival at the CBRC and again ≥163 days later. Paired hair samples showed declines in cortisol concentrations of 12-88% in 38 of 45 (84%, pbears after arrival and acclimation at the rehabilitation facility. Concentrations of cortisol in hair from bears on the bile farm were similar to initial concentrations upon arrival at the CBRC but were higher than those collected after bears had been at the CBRC for ≥163 days. Fecal glucocorticoid concentrations varied across months and were highest in April and declined through December, possibly reflecting seasonal patterns, responses to the arrival and socialization of new bears at the CBRC, and/or annual metabolic change. Data from segmental analysis of hair supports the first of these explanations. Our findings indicate that bears produced elevated concentrations of glucocorticoids on bile farms, and that activity of the HPA axis declined following relocation. Thus, hair cortisol analyses are particularly well suited to long-term, retrospective assessments of glucocorticoids in ursids. By contrast, fecal measures were not clearly associated with rehabilitation, but rather reflected more subtle endocrine changes, possibly related to seasonality. Copyright © 2013 Elsevier Inc. All rights reserved.

  18. CAREM-25: Residual heat removal system

    International Nuclear Information System (INIS)

    Arvia, Roberto P.; Coppari, Norberto R.; Gomez de Soler, Susana M.; Ramilo, Lucia B.

    2000-01-01

    The objective of this work was the definition and consolidation of the residual heat removal system for the CAREM 25 reactor. The function of this system is cool down the primary circuit, removing the core decay heat from hot stand-by to cold shutdown and during refueling. In addition, this system heats the primary water from the cold shutdown condition to hot stand-by condition during the reactor start up previous to criticality. The system has been designed according to the requirements of the standards: ANSI/ANS 51.1 'Nuclear safety criteria for the design of stationary PWR plants'; ANSI/ANS 58.11 'Design criteria for safe shutdown following selected design basis events in light water reactors' and ANSI/ANS 58.9 'Single failure criteria for light water reactor safety-related fluid systems'. The suggested design fulfills the required functions and design criteria standards. (author)

  19. Nuclear energy waste-space transportation and removal

    Science.gov (United States)

    Burns, R. E.

    1975-01-01

    A method for utilizing the decay heat of actinide wastes to power an electric thrust vehicle is proposed. The vehicle, launched by shuttle to earth orbit and to earth escape by a tug, obtains electrical power from the actinide waste heat by thermionic converters. The heavy gamma ray and neutron shielding which is necessary as a safety feature is removed in orbit and returned to earth for reuse. The problems associated with safety are dealt with in depth. A method for eliminating fission wastes via chemical propulsion is briefly discussed.

  20. Passive heat removal system with injector-condenser

    Energy Technology Data Exchange (ETDEWEB)

    Soplenkov, K I [All-Russian Inst. of Nuclear Power Plant Operation, Electrogorsk Research and Engineering Centre of Nuclear Power Safety (Russian Federation)

    1996-12-01

    The system described in this paper is a passive system for decay heat removal from WWERs. It operates off the secondary side of the steam generators (SG). Steam is taken from the SG to operate a passive injector pump which causes secondary fluid to be pumped through a heat exchanger. Variants pass either water or steam from the SG through the heat exchanger. There is a passive initiation scheme. The programme for experimental and theoretical validation of the system is described. (author). 8 figs.

  1. Analyses of the response of a complex weighted network to nodes removal strategies considering links weight: The case of the Beijing urban road system

    Science.gov (United States)

    Bellingeri, Michele; Lu, Zhe-Ming; Cassi, Davide; Scotognella, Francesco

    2018-02-01

    Complex network response to node loss is a central question in different fields of science ranging from physics, sociology, biology to ecology. Previous studies considered binary networks where the weight of the links is not accounted for. However, in real-world networks the weights of connections can be widely different. Here, we analyzed the response of real-world road traffic complex network of Beijing, the most prosperous city in China. We produced nodes removal attack simulations using classic binary node features and we introduced weighted ranks for node importance. We measured the network functioning during nodes removal with three different parameters: the size of the largest connected cluster (LCC), the binary network efficiency (Bin EFF) and the weighted network efficiency (Weg EFF). We find that removing nodes according to weighted rank, i.e. considering the weight of the links as a number of taxi flows along the roads, produced in general the highest damage in the system. Our results show that: (i) in order to model Beijing road complex networks response to nodes (intersections) failure, it is necessary to consider the weight of the links; (ii) to discover the best attack strategy, it is important to use nodes rank accounting links weight.

  2. Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

    Science.gov (United States)

    Xiaokang, ZHANG; Songlin, LIU; Xia, LI; Qingjun, ZHU; Jia, LI

    2017-11-01

    The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR). Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage, and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW. The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3. The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.

  3. Simplified analysis of passive residual heat removal systems for small size PWR's

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1992-02-01

    The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

  4. Safety studies on heat transport and afterheat removal for GCR accident conditions

    International Nuclear Information System (INIS)

    Hishida, Makoto

    1996-01-01

    The IAEA coordinated an international research program on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions (CRP-3)'. America, China, France, Germany, Japan, Netherlands and Russia participate the program. Final goal of the program is to show clearly to the world one of the most important salient features of the HTGR, that is the HTGR reactor can be cooled down by passive measures without causing any damage to the nuclear reactor system even in accidental conditions, and to make clear the boundaries (or restrictions) for the passive cooling regime. The first 5 year term of the coordinate program started in 1993 and established a goal to improve common knowledge for decay heat removal and to improve our tools, like computer codes and analytical models for the prediction of the performance of decay heat removal system. We are now performing benchmark problems for these purposes. The present efforts are concentrated on the benchmark for the passive heat removal performance outside the reactor vessel, partly because we have two different type of the HTGR in the world, the pebble bed type and the block type reactor. They have quite different heat dissipation behavior inside the reactor vessel. However, they have quite similar residual heat removal process outside the reactor vessel. For the first step of the international cooperation, we selected the common problem. After finishing the present benchmark we are planning to proceed to tackle the inside heat removal problem. (J.P.N.)

  5. Monitoring system in Labview data logging and supervisory control module and Labview 8 with process analyses of the Cryogenic Pilot Plant for Tritium Removal

    International Nuclear Information System (INIS)

    Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Bucur, Ciprian; Stefan, Liviu; Bornea, Anisia; Stefanescu, Ioan

    2008-01-01

    The system responds to the monitoring requirements of the technological processes specific to the nuclear installation that processes radioactive substances, with severe consequences in case of technological failure, as is the case with a tritium processing nuclear plant. The big amount of data that needs to be processed, stored and accessed for real time simulation and optimization demands the achievement of the virtual technologic platform where the data acquiring, control and analysis systems of the technological process can be integrated with an advanced technological monitoring system. Thus, integrated computing and monitoring systems needed for the supervising of the technological process will be effected, to be then continued with optimization of the system, by implementing new and performing methods corresponding to the technological processes within the tritium removal processing nuclear plants. (authors)

  6. Genomic and in Situ Analyses Reveal the Micropruina spp. as Abundant Fermentative Glycogen Accumulating Organisms in Enhanced Biological Phosphorus Removal Systems

    Directory of Open Access Journals (Sweden)

    Simon J. McIlroy

    2018-05-01

    Full Text Available Enhanced biological phosphorus removal (EBPR involves the cycling of biomass through carbon-rich (feast and carbon-deficient (famine conditions, promoting the activity of polyphosphate accumulating organisms (PAOs. However, several alternate metabolic strategies, without polyphosphate storage, are possessed by other organisms, which can compete with the PAO for carbon at the potential expense of EBPR efficiency. The most studied are the glycogen accumulating organisms (GAOs, which utilize aerobically stored glycogen to energize anaerobic substrate uptake and storage. In full-scale systems the Micropruina spp. are among the most abundant of the proposed GAO, yet little is known about their ecophysiology. In the current study, genomic and metabolomic studies were performed on Micropruina glycogenica str. Lg2T and compared to the in situ physiology of members of the genus in EBPR plants using state-of-the-art single cell techniques. The Micropruina spp. were observed to take up carbon, including sugars and amino acids, under anaerobic conditions, which were partly fermented to lactic acid, acetate, propionate, and ethanol, and partly stored as glycogen for potential aerobic use. Fermentation was not directly demonstrated for the abundant members of the genus in situ, but was strongly supported by the confirmation of anaerobic uptake of carbon and glycogen storage in the absence of detectable polyhydroxyalkanoates or polyphosphate reserves. This physiology is markedly different from the classical GAO model. The amount of carbon stored by fermentative organisms has potentially important implications for phosphorus removal – as they compete for substrates with the Tetrasphaera PAO and stored carbon is not made available to the “Candidatus Accumulibacter” PAO under anaerobic conditions. This study shows that the current models of the competition between PAO and GAO are too simplistic and may need to be revised to take into account the impact of

  7. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  8. Kinetic and isotherm analyses for thorium (IV) adsorptive removal from aqueous solutions by modified magnetite nanoparticle using response surface methodology (RSM)

    Energy Technology Data Exchange (ETDEWEB)

    Karimi, Mohammad, E-mail: m.karimi407@alumni.ut.ac.ir [School of Chemical Engineering, College of Engineering, University of Tehran, P.O. Box: 11365-4563, Tehran (Iran, Islamic Republic of); Milani, Saeid Alamdar [Nuclear Fuel Cycle Research School, Nuclear Science and Technology Research Institute, AEOI, P.O. Box: 14893-836, Tehran (Iran, Islamic Republic of); Abolgashemi, Hossein [School of Chemical Engineering, College of Engineering, University of Tehran, P.O. Box: 11365-4563, Tehran (Iran, Islamic Republic of)

    2016-10-15

    In this study, the ability and the adsorption capacity of magnetite/aminopropyltriethoxysilane/glutaraldehyde (Fe{sub 3}O{sub 4}/APTES/GA) adsorbent were evaluated for the adsorption of thorium (IV) ions from aqueous solutions. The influence of the several variables such as pH (1–5), Th (IV) initial concentration (50–300 mg L{sup −1}) and adsorbent concentration (1–5 g L{sup −1}) on the Th (IV) adsorption were investigated by response surface methodology (RSM). The results showed that the highest absorption capacity (q) was 107.23 mg g{sup −1} with respect to pH = 4.5, initial concentration of 250 mg L{sup −1} and adsorbent concentration of 1 g L{sup −1} for 90 min. Modeling equilibrium sorption data with the Langmuir, Freundlich and Dubinin–Radushkevich models pointed out that the results were in good agreement with Langmuir model. The experimental kinetic data were well fitted to pseudo-second-order equation with R{sup 2} = 0.9739. Also thermodynamic parameters (ΔG{sup o}, ΔH{sup o}, ΔS{sup o}) declared that the Th (IV) adsorption was endothermic and spontaneous. - Highlights: • Thorium ions were removed from aqueous solutions by modified magnetite nanoparticle. • The effects of process variables on adsorption capacity were investigated by RSM. • Thermodynamic parameters showed that the adsorption was endothermic and spontaneous. • The equilibrium data for the adsorption of Thorium followed the Langmuir isotherm. • The experimental kinetic data were described by the pseudo-second-order equation.

  9. The status of work in the USSR on using inherent self-protection features of fast reactors, of passive and active means of shutdown and decay heat removal system

    International Nuclear Information System (INIS)

    Buksha, Yu.K.

    1991-01-01

    Extensive studies on fast reactor safety, aimed to increase intrinsic safety features and introduce passive safety means, are under way in the USSR. Design of the BN-800 reactor core with a close-to-zero sodium void effect of reactivity has been developed, complementary reactivity control means, based on passive principles are being implemented. As a whole, after the Chernobyl accident, the preference is given to the 'passive' full proof methods of safety. This approach may possibly seem excessive and may result in some losses concerning reactor economic characteristics

  10. Residual heat removal pump retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; McKenna, J.M.

    1990-01-01

    Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

  11. Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity when heat removing from a power transformer of combined heat and power station

    Directory of Open Access Journals (Sweden)

    Nurpeiis Atlant

    2017-01-01

    Full Text Available Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity was conducted when heat removing from an oil tank of a power transformer of combined heat and power station (CHP. The power transformer cooling system by the closed biphasic thermosyphon was proposed. The mathematical modeling of heat transfer and phase transitions of coolant in the thermosyphon was performed. The problem of heat transfer is formulated in dimensionless variables “velocity vorticity vector – current function – temperature” and solved by finite difference method. As a result of numerical simulation it is found that an increase in the vapor channel length from 0.15m to 1m leads to increasing the temperature difference by 3.5 K.

  12. Hair Removal

    Science.gov (United States)

    ... Staying Safe Videos for Educators Search English Español Hair Removal KidsHealth / For Teens / Hair Removal What's in ... you need any of them? Different Types of Hair Before removing hair, it helps to know about ...

  13. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  14. Decay heat measurement of U-235

    International Nuclear Information System (INIS)

    Baumung, K.

    1976-01-01

    The calorimeter and the transport mechanism for the fuel samples was designed and is under construction now. Calculations of the heat-source distributions for different 235U-contents led to an optimal enrichment of the UO 2 -samples which minimizes the effects of the bad heat conductivity of the oxide on temperature measurement. Monte-Carlo-calculations of the γ-leakage-spectra yielded data which allow, from the γ-energy-flow measurements, to calculate the total γ-energy loss as well as the portions of the β- and γ-heating. (orig.) [de

  15. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  16. Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

    International Nuclear Information System (INIS)

    Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

    2015-01-01

    The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

  17. Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

    International Nuclear Information System (INIS)

    Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

    2006-01-01

    The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

  18. Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

    Science.gov (United States)

    Bersano, Andrea

    With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

  19. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  20. Concepts for passive heat removal and filtration systems under core meltdown conditions

    International Nuclear Information System (INIS)

    Wilhelm, J.G.; Neitzel, H.-J.

    1993-01-01

    The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

  1. Update and evaluation of decay data for spent nuclear fuel analyses

    Directory of Open Access Journals (Sweden)

    Simeonov Teodosi

    2017-01-01

    Full Text Available Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL and processed (ESTAR sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources. Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

  2. Update and evaluation of decay data for spent nuclear fuel analyses

    Science.gov (United States)

    Simeonov, Teodosi; Wemple, Charles

    2017-09-01

    Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

  3. Spleen removal

    Science.gov (United States)

    ... spleen. Sickle cell anemia . Splenic artery aneurysm (rare). Trauma to the spleen. Risks Risks for anesthesia and surgery in general ... removal - series References Brandow AM, Camitta BM. Hyposplenism, splenic trauma, and splenectomy. In: Kliegman RM, Stanton BF, St. ...

  4. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  5. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  6. Hair Removal

    DEFF Research Database (Denmark)

    Hædersdal, Merete

    2011-01-01

    Hair removal with optical devices has become a popular mainstream treatment that today is considered the most efficient method for the reduction of unwanted hair. Photothermal destruction of hair follicles constitutes the fundamental concept of hair removal with red and near-infrared wavelengths...... suitable for targeting follicular and hair shaft melanin: normal mode ruby laser (694 nm), normal mode alexandrite laser (755 nm), pulsed diode lasers (800, 810 nm), long-pulse Nd:YAG laser (1,064 nm), and intense pulsed light (IPL) sources (590-1,200 nm). The ideal patient has thick dark terminal hair......, white skin, and a normal hormonal status. Currently, no method of lifelong permanent hair eradication is available, and it is important that patients have realistic expectations. Substantial evidence has been found for short-term hair removal efficacy of up to 6 months after treatment with the available...

  7. Hair removal

    DEFF Research Database (Denmark)

    Haedersdal, Merete; Haak, Christina S

    2011-01-01

    Hair removal with optical devices has become a popular mainstream treatment that today is considered the most efficient method for the reduction of unwanted hair. Photothermal destruction of hair follicles constitutes the fundamental concept of hair removal with red and near-infrared wavelengths...... suitable for targeting follicular and hair shaft melanin: normal mode ruby laser (694 nm), normal mode alexandrite laser (755 nm), pulsed diode lasers (800, 810 nm), long-pulse Nd:YAG laser (1,064 nm), and intense pulsed light (IPL) sources (590-1,200 nm). The ideal patient has thick dark terminal hair......, white skin, and a normal hormonal status. Currently, no method of lifelong permanent hair eradication is available, and it is important that patients have realistic expectations. Substantial evidence has been found for short-term hair removal efficacy of up to 6 months after treatment with the available...

  8. Plate removal following orthognathic surgery.

    Science.gov (United States)

    Little, Mhairi; Langford, Richard Julian; Bhanji, Adam; Farr, David

    2015-11-01

    The objectives of this study are to determine the removal rates of orthognathic plates used during orthognathic surgery at James Cook University Hospital and describe the reasons for plate removal. 202 consecutive orthognathic cases were identified between July 2004 and July 2012. Demographics and procedure details were collected for these patients. Patients from this group who returned to theatre for plate removal between July 2004 and November 2012 were identified and their notes were analysed for data including reason for plate removal, age, smoking status, sex and time to plate removal. 3.2% of plates were removed with proportionally more plates removed from the mandible than the maxilla. 10.4% of patients required removal of one or more plate. Most plates were removed within the first post-operative year. The commonest reasons for plate removal were plate exposure and infection. The plate removal rates in our study are comparable to those seen in the literature. Copyright © 2015 European Association for Cranio-Maxillo-Facial Surgery. Published by Elsevier Ltd. All rights reserved.

  9. Removing Bureaucracy

    Science.gov (United States)

    2015-08-01

    11 Defense AT&L: July–August 2015 Removing Bureaucracy Katharina G. McFarland McFarland is Assistant Secretary of Defense for Acquisition. I once...involvement from all of the Service warfighting areas came together to scrub the program requirements due to concern over the “ bureaucracy ” and... Bureaucracy ” that focuses on reducing cycle time, staffing time and all forms of inefficiencies. This includes review of those burdens that Congress

  10. Effect analysis of core barrel openings under CEFR normal condition

    International Nuclear Information System (INIS)

    Zhang Yabo; Yang Hongyi

    2008-01-01

    Openings on the bottom of core barrel are important part of the decay heat removal system of China Experimental Fast Reactor (CEFR), which are designed to discharge the decay heat from reactor under accident condition. This paper analyses the effect of the openings design on the normal operation condition using the famouse CFD code CFX. The result indicates that the decay heat can be discharged safely and at the same time the effect of core barrel openings on the normal operation condition is acceptable. (authors)

  11. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    International Nuclear Information System (INIS)

    Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

    1977-01-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

  12. Milk removal

    OpenAIRE

    Ferneborg, Sabine

    2016-01-01

    Milk from dairy cows is a staple dietary component for humans all over the world. Regardless of whether milk is consumed in its purest, unaltered form or as high-end products such as fine cheese or ice cream, it needs to be of high quality when taken from the cow, produced at a low price and produced in a system that consider aspects such as animal health, animal welfare and sustainability. This thesis investigated the role of milk removal and the importance of residual milk on milk yield...

  13. Organic micropollutant removal during river bank filtration

    NARCIS (Netherlands)

    Bertelkamp, C.

    2015-01-01

    This study investigated the factors influencing the main removal mechanisms (adsorption and biodegradation) for organic micropollutant (OMP) removal during river bank filtration (RBF) and the possibility of developing a predictive model of this process for OMP removal during RBF. Chapter 2 analysed

  14. PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

    International Nuclear Information System (INIS)

    1991-03-01

    An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

  15. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  16. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

    2015-01-01

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  17. Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

    Directory of Open Access Journals (Sweden)

    Afshin Hedayat

    2017-08-01

    Full Text Available In this paper, a complete station blackout (SBO or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR. The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank, safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

  18. Simulation and transient analyses of a complete passive heat removal system in a downward cooling pool-type material testing reactor against a complete station blackout and long-term natural convection mode using the RELAP5/3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin [Reactor and Nuclear Safety School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2017-08-15

    In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal–hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

  19. Study of passive residual heat removal system of a modular small PWR reactor; Estudo do sistema passivo de remoção de calor residual de um reator PWR pequeno modular

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Nathália N., E-mail: nathalianunes@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Faccini, José L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Su, Jian, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS.

  20. Post-accident heat removal research: A state of the art review

    International Nuclear Information System (INIS)

    Mueller, U.; Schulenberg, T.

    1983-11-01

    For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

  1. The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

    2007-01-01

    CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

  2. Neutronic and thermal hydraulic analyses of LEU targets irradiated in a research reactor for Molybdenum-99 production

    International Nuclear Information System (INIS)

    Jo, Daeseong; Lee, Kyung-Hoon; Kim, Hong-Chul; Chae, Heetaek

    2014-01-01

    Highlights: • Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99. • Heat production during and after irradiation was evaluated using MCNP and ORIGEN-APR. • Cooling capacities under various cooling conditions were evaluated using TMAP. • Natural convective cooling was adequate for the decay power after 0.03 h from withdrawal. • Maximum temperature of the target decayed for 24 h does not exceed the blistering threshold. - Abstract: Neutronic and thermal hydraulic analyses of irradiated fuel plates for Molybdenum-99 production in a research reactor were performed to investigate (1) the heat production during irradiation, (2) decay heat after irradiation, and (3) cooling capacities under various cooling conditions. The heat production on the target plates irradiated in the core was evaluated using the MCNP code. The decay heat after irradiation was evaluated using the ORIGEN-APR code, and compared against ANSI/ANS-5.1-1979. The cooling capacities of forced convective cooling during irradiation and natural convective cooling after irradiation were estimated using the TMAP code. An equilibrium core with different core statuses i.e., BOC, MOC, and EOC was used to evaluate power released from the targets and the axial power distribution. Based on the neutronic calculations, thermal margins i.e., the maximum wall temperature, minimum ONB temperature margin, and minimum CHF ratio were estimated, and the cooling strategy of the fission Mo targets was discussed. The targets were cooled by forced convective cooling during irradiation, and cooled by natural convective cooling after irradiation. For a further production process, the targets transported to a hot cell were exposed to the air, and cooled by natural convection cooling in air. As a result, the maximum wall temperature remained below the ONB temperature while the targets were under water, and the maximum wall temperature remained under the blistering limit while the targets

  3. Removing Hair Safely

    Science.gov (United States)

    ... For Consumers Home For Consumers Consumer Updates Removing Hair Safely Share Tweet Linkedin Pin it More sharing ... related to common methods of hair removal. Laser Hair Removal In this method, a laser destroys hair ...

  4. Investigation of Characteristics of Passive Heat Removal System Based on the Assembled Heat Transfer Tube

    Directory of Open Access Journals (Sweden)

    Xiangcheng Wu

    2016-12-01

    Full Text Available To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450°C to 700°C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

  5. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    International Nuclear Information System (INIS)

    Gautier, G.M.

    1995-01-01

    In the framework of research on diversified means for removing the residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system), which includes integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: it is independent of the state of the loops, even if the volume of water in the primary circuit is small, it is compatible with either a passive or an active operation mode, and compatible with any other decay heat removal systems. An evaluation is presented here of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of another system. The results of this evaluation show the interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system, no need for the use of a high pressure safety injection system. (author). 4 refs., 7 figs., 1 tab

  6. Investigation of characteristics of passive heat removal system based on the assembled heat transfer tube

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Xiang Cheng; Yan, Changqi; Meng, Zhao Ming; Chen, Kailun; Song, Shao Chuang; Yang, Zong Hao; Yu, Jie [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2016-12-15

    To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450 .deg. C to 700 .deg. C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

  7. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [Commissariat a l`Energie Atomique, Saint-Paul-Lez-Durance (France)

    1995-09-01

    In the framework of research on diversified means for removing residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system). This system consists of integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: such as independent from the state of the loops, low volume of water in the primary circuit, compatibility with either a passive or an active operation mode, and compatibility with any other decay heat removal systems. This paper presents an evaluation of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of any another system. The results of this evaluation show the potential interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system and no need for the use of a high pressure safety injection system.

  8. Physical limits on steam generation by radioactive decay heat

    International Nuclear Information System (INIS)

    Chesnut, D.A.

    1991-12-01

    This report briefly discusses the possibilities that flood water contacting the hot radioactive waste and rock at Yucca Mountain could produce enough steam to lift the top of the mountain off the repository

  9. Method for utilizing decay heat from radioactive nuclear wastes

    International Nuclear Information System (INIS)

    Busey, H.M.

    1974-01-01

    Management of radioactive heat-producing waste material while safely utilizing the heat thereof is accomplished by encapsulating the wastes after a cooling period, transporting the capsules to a facility including a plurality of vertically disposed storage tubes, lowering the capsules as they arrive at the facility into the storage tubes, cooling the storage tubes by circulating a gas thereover, employing the so heated gas to obtain an economically beneficial result, and continually adding waste capsules to the facility as they arrive thereat over a substantial period of time

  10. ARSENIC REMOVAL BY IRON REMOVAL PROCESSES

    Science.gov (United States)

    Presentation will discuss the removal of arsenic from drinking water using iron removal processes that include oxidation/filtration and the manganese greensand processes. Presentation includes results of U.S. EPA field studies conducted in Michigan and Ohio on existing iron remo...

  11. Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

    International Nuclear Information System (INIS)

    David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

  12. Assessment of ASME code examinations on regenerative, letdown and residual heat removal heat exchangers

    International Nuclear Information System (INIS)

    Gosselin, Stephen R.; Cumblidge, Stephen E.; Anderson, Michael T.; Simonen, Fredric A.; Tinsley, G A.; Lydell, B.; Doctor, Steven R.

    2005-01-01

    Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principle reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers regenerative heat exchangers and residual (decay) heat removal heat exchangers Design, operating, common preventative maintenance practices, and potential degradation mechanisms are reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp- database and covers PWR plants worldwide for the period 1970-2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers is performed. The results are then reviewed to discuss the examinations relative to plant safety and

  13. Spider Vein Removal

    Science.gov (United States)

    Spider veins: How are they removed? I have spider veins on my legs. What options are available ... M.D. Several options are available to remove spider veins — thin red lines or weblike networks of ...

  14. Three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

    International Nuclear Information System (INIS)

    Yu, S.-O.; Kim, M.; Kim, H.-J.

    2002-01-01

    A CANDU reactor has the unique features and the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors. If there is the loss of coolant accident (LOCA) and a coincident failure of the emergency coolant injection (ECI) system, the heavy water moderator is continuously cooled, providing a heat sink for decay heat produced in the fuel. Therefore, it is one of major concerns to estimate the local subcooling of moderator inside the calandria vessel under postulated accident in CANDU safety analyses. The Canadian Nuclear Safety Commission (CNSC), a regulatory body in Canada, categorized the integrity of moderator as a generic safety issue and recommended that a series of experimental works be performed to verify the safety evaluation codes for individual simulated condition of nuclear power plant, comparing with the results of three-dimensional experimental data. In this study, three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL (Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. Three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. (author)

  15. PLANNING YOUR REMOVALS

    CERN Multimedia

    Service déménagement; ST Division

    1999-01-01

    To give you better service and avoid lengthy delays, the Removals Service advises you to refrain from programming moves between 26 July and 3 September, as large-scale removals are already planned during this summer period.Thanking you in advance for your co-operation and understanding.Removals Service STTel. 74185 / Mobile 164017

  16. PROGRAMMING OFFICE REMOVALS

    CERN Multimedia

    Groupe ST-HM

    2000-01-01

    The Removals Service recommends you to plan your removals well in advance, taking into account the fact that the Transport and Handling Group’s main priority remains the dismantling of LEP and the installation of the LHC. The requests can be made by: http://st.web.cern.ch/st/hm/removal/DEMEE.HTM Thank you for your cooperation.

  17. Dam removal: Listening in

    Science.gov (United States)

    Foley, M. M.; Bellmore, J. R.; O'Connor, J. E.; Duda, J. J.; East, A. E.; Grant, G. E.; Anderson, C. W.; Bountry, J. A.; Collins, M. J.; Connolly, P. J.; Craig, L. S.; Evans, J. E.; Greene, S. L.; Magilligan, F. J.; Magirl, C. S.; Major, J. J.; Pess, G. R.; Randle, T. J.; Shafroth, P. B.; Torgersen, C. E.; Tullos, D.; Wilcox, A. C.

    2017-07-01

    Dam removal is widely used as an approach for river restoration in the United States. The increase in dam removals—particularly large dams—and associated dam-removal studies over the last few decades motivated a working group at the USGS John Wesley Powell Center for Analysis and Synthesis to review and synthesize available studies of dam removals and their findings. Based on dam removals thus far, some general conclusions have emerged: (1) physical responses are typically fast, with the rate of sediment erosion largely dependent on sediment characteristics and dam-removal strategy; (2) ecological responses to dam removal differ among the affected upstream, downstream, and reservoir reaches; (3) dam removal tends to quickly reestablish connectivity, restoring the movement of material and organisms between upstream and downstream river reaches; (4) geographic context, river history, and land use significantly influence river restoration trajectories and recovery potential because they control broader physical and ecological processes and conditions; and (5) quantitative modeling capability is improving, particularly for physical and broad-scale ecological effects, and gives managers information needed to understand and predict long-term effects of dam removal on riverine ecosystems. Although these studies collectively enhance our understanding of how riverine ecosystems respond to dam removal, knowledge gaps remain because most studies have been short (< 5 years) and do not adequately represent the diversity of dam types, watershed conditions, and dam-removal methods in the U.S.

  18. Hair removal in adolescence

    Directory of Open Access Journals (Sweden)

    Sandra Pereira

    2015-06-01

    Full Text Available Introduction: Due to hormonal stimulation during puberty, changes occur in hair type and distribution. In both sexes, body and facial unwanted hair may have a negative psychological impact on the teenager. There are several available methods of hair removal, but the choice of the most suitable one for each individual can raise doubts. Objective: To review the main methods of hair removal and clarify their indications, advantages and disadvantages. Development: There are several removal methods currently available. Shaving and depilation with chemicals products are temporary methods, that need frequent repetition, because hair removal is next to the cutaneous surface. The epilating methods in which there is full hair extraction include: epilation with wax, thread, tweezers, epilating machines, laser, intense pulsed light, and electrolysis. Conclusions: The age of beginning hair removal and the method choice must be individualized and take into consideration the skin and hair type, location, dermatological and endocrine problems, removal frequency, cost and personal preferences.

  19. Particle adhesion and removal

    CERN Document Server

    Mittal, K L

    2015-01-01

    The book provides a comprehensive and easily accessible reference source covering all important aspects of particle adhesion and removal.  The core objective is to cover both fundamental and applied aspects of particle adhesion and removal with emphasis on recent developments.  Among the topics to be covered include: 1. Fundamentals of surface forces in particle adhesion and removal.2. Mechanisms of particle adhesion and removal.3. Experimental methods (e.g. AFM, SFA,SFM,IFM, etc.) to understand  particle-particle and particle-substrate interactions.4. Mechanics of adhesion of micro- and  n

  20. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely

  1. Skin lesion removal

    Science.gov (United States)

    ... likely to be done when there is a concern about a skin cancer. Most often, an area the shape of an ellipse is removed, as this makes it easier to close with stitches. The entire lesion is removed, going as deep as the fat, if needed, to ...

  2. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.; Boss, C. R.

    2006-01-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  3. An estimation of core damage frequency of a pressurized water reactor during midloop operation due to loss of residual heat removal

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    1995-01-01

    The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 x 10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation

  4. Device for removing fur

    International Nuclear Information System (INIS)

    Hanawa, Minoru; Nakagawa, Takao; Sakuma, Toyoo; Yonemura, Eizo.

    1976-01-01

    Purpose: To effectively remove fur adhered to fuel rods and to increase working efficiency without use of a lengthy hose. Constitution: In the fur removing device of the present invention, brushes rotated by gears are provided within a casing so that fur adhered to the fuel rods are removed by the brushes and water is rotatably moved by blades housed therein to outwardly blow fur floating in water by means of a centrifugal force. Then, the fur is filtered by a filter outwardly provided. In this way, the fur may be collected within the device to avoid contamination to others. (Kamimura, M.)

  5. Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

  6. Mitigation Measures Following a Loss-of-Residual-Heat-Removal Event During Shutdown

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

    2000-01-01

    The transient following a loss-of-residual-heat-removal event during shutdown was analyzed to determine the containment closure time (CCT) to prevent uncontrolled release of fission products and the gravity-injection path and rate (GIPR) for effective core cooling using the RELAP5/MOD3.2 code. The plant conditions of Yonggwang Units 3 and 4, a pressurized water reactor (PWR) of 2815-MW(thermal) power in Korea, were reviewed, and possible event sequences were identified. From the CCT analysis for the five cases of typical plant configurations, it was estimated for the earliest CCT to be 40 min after the event in a case with a large cold-leg opening and emptied steam generators (SGs). However, the case with water-filled SGs significantly delayed the CCT through the heat removal to the secondary side. From the GIPR analysis for the six possible gravity-injection paths from the refueling water storage tank (RWST), the case with the injection point and opening on the other leg side was estimated to be the most suitable path to avoid core boiling. In addition, from the sensitivity study, it was evaluated for the plant to be capable of providing the core cooling for the long-term transient if nominal RWST water is available. As a result, these analysis methods and results will provide useful information in understanding the plant behavior and preparing the mitigation measures after the event, especially for Combustion Engineering-type PWR plants. However, to directly apply the analysis results to the emergency procedure for such an event, additional case studies are needed for a wide range of operating conditions such as reactor coolant inventory, RWST water temperature, and core decay heat rate

  7. (suspended solids and metals) removal efficiencies

    African Journals Online (AJOL)

    ABSTRACT. Presented in this paper are the results of correlational analyses and logistic regression between metal substances (Cd, Cu,. Pb, Zn), as well as suspended solids removal, and physical pond parameters of 19 stormwater retention pond case studies obtained from the International Stormwater BMP database.

  8. Bridge removal plan requirements.

    Science.gov (United States)

    2011-04-15

    This report provides resources that detail specifications and guidelines related to bridge removal plans across the : United States. We have organized the information into three sections: : ! National Guidance : Includes language from AASHTO specific...

  9. Reactor for removing ammonia

    Science.gov (United States)

    Luo, Weifang [Livermore, CA; Stewart, Kenneth D [Valley Springs, CA

    2009-11-17

    Disclosed is a device for removing trace amounts of ammonia from a stream of gas, particularly hydrogen gas, prepared by a reformation apparatus. The apparatus is used to prevent PEM "poisoning" in a fuel cell receiving the incoming hydrogen stream.

  10. Optical hair removal.

    Science.gov (United States)

    Ort, R J; Anderson, R R

    1999-06-01

    Traditional methods of hair removal have proven unsatisfactory for many individuals with excessive or unwanted hair. In the last few years, several lasers and xenon flashlamps have been developed that promise to fulfill the need for a practical, safe, and long-lasting method of hair removal. Aggressive marketing of these has contributed to their popularity among patients and physicians. However, significant controversy and confusion surrounds this field. This article provides a detailed explanation of the scientific underpinnings for optical hair removal and explores the advantages and disadvantages of the various devices currently available (Nd:YAG, ruby, alexandrite, diode lasers, and xenon flashlamp). Treatment and safety guidelines are provided to assist the practitioner in the use of these devices. Although the field of optical hair removal is still in its infancy, initial reports of long-term efficacy are encouraging.

  11. Laparoscopic Spleen Removal (Splenectomy)

    Science.gov (United States)

    ... Affairs and Humanitarian Efforts Login Laparoscopic Spleen Removal (Splenectomy) Patient Information from SAGES Download PDF Find a ... are suspected. What are the Advantages of Laparoscopic Splenectomy? Individual results may vary depending on your overall ...

  12. Space Debris Removal: A Game Theoretic Analysis

    Directory of Open Access Journals (Sweden)

    Richard Klima

    2016-08-01

    Full Text Available We analyse active space debris removal efforts from a strategic, game-theoretical perspective. Space debris is non-manoeuvrable, human-made objects orbiting Earth, which pose a significant threat to operational spacecraft. Active debris removal missions have been considered and investigated by different space agencies with the goal to protect valuable assets present in strategic orbital environments. An active debris removal mission is costly, but has a positive effect for all satellites in the same orbital band. This leads to a dilemma: each agency is faced with the choice between the individually costly action of debris removal, which has a positive impact on all players; or wait and hope that others jump in and do the ‘dirty’ work. The risk of the latter action is that, if everyone waits, the joint outcome will be catastrophic, leading to what in game theory is referred to as the ‘tragedy of the commons’. We introduce and thoroughly analyse this dilemma using empirical game theory and a space debris simulator. We consider two- and three-player settings, investigate the strategic properties and equilibria of the game and find that the cost/benefit ratio of debris removal strongly affects the game dynamics.

  13. Tube plug removal machine

    International Nuclear Information System (INIS)

    Hawkins, P.J.

    1987-01-01

    In a nuclear steam generator wherein some faulty tubes have been isolated by mechanical plugging, to remove a selected plug without damaging the associated tube, a plug removal machine is used. The machine drills into a plug portion with a tap drill bit having a drill portion a tap portion and a threaded portion, engaging that plug portion with the threaded portion after the drilled hole has been threaded by the tap portion thereof, and removing a portion of the plug in the tube with a counterbore drill bit mounted concentrically about the tap drill bit. A trip pin and trip spline disengage the tap drill bit from the motor. The counterbore drill bit is thereafter self-centered with respect to the tube and plug about the now stationary tap drill bit. After a portion of the plug has been removed by the counterbore drill bit, pulling on the top drill bit by grippers on slots will remove the remaining plug portion from the tube. (author)

  14. Laser Beam Focus Analyser

    DEFF Research Database (Denmark)

    Nielsen, Peter Carøe; Hansen, Hans Nørgaard; Olsen, Flemming Ove

    2007-01-01

    the obtainable features in direct laser machining as well as heat affected zones in welding processes. This paper describes the development of a measuring unit capable of analysing beam shape and diameter of lasers to be used in manufacturing processes. The analyser is based on the principle of a rotating......The quantitative and qualitative description of laser beam characteristics is important for process implementation and optimisation. In particular, a need for quantitative characterisation of beam diameter was identified when using fibre lasers for micro manufacturing. Here the beam diameter limits...... mechanical wire being swept through the laser beam at varying Z-heights. The reflected signal is analysed and the resulting beam profile determined. The development comprised the design of a flexible fixture capable of providing both rotation and Z-axis movement, control software including data capture...

  15. Optimising laser tattoo removal

    Directory of Open Access Journals (Sweden)

    Kabir Sardana

    2015-01-01

    Full Text Available Lasers are the standard modality for tattoo removal. Though there are various factors that determine the results, we have divided them into three logical headings, laser dependant factors such as type of laser and beam modifications, tattoo dependent factors like size and depth, colour of pigment and lastly host dependent factors, which includes primarily the presence of a robust immune response. Modifications in the existing techniques may help in better clinical outcome with minimal risk of complications. This article provides an insight into some of these techniques along with a detailed account of the factors involved in tattoo removal.

  16. Successful removable partial dentures.

    Science.gov (United States)

    Lynch, Christopher D

    2012-03-01

    Removable partial dentures (RPDs) remain a mainstay of prosthodontic care for partially dentate patients. Appropriately designed, they can restore masticatory efficiency, improve aesthetics and speech, and help secure overall oral health. However, challenges remain in providing such treatments, including maintaining adequate plaque control, achieving adequate retention, and facilitating patient tolerance. The aim of this paper is to review the successful provision of RPDs. Removable partial dentures are a successful form of treatment for replacing missing teeth, and can be successfully provided with appropriate design and fabrication concepts in mind.

  17. Optimising Laser Tattoo Removal

    Science.gov (United States)

    Sardana, Kabir; Ranjan, Rashmi; Ghunawat, Sneha

    2015-01-01

    Lasers are the standard modality for tattoo removal. Though there are various factors that determine the results, we have divided them into three logical headings, laser dependant factors such as type of laser and beam modifications, tattoo dependent factors like size and depth, colour of pigment and lastly host dependent factors, which includes primarily the presence of a robust immune response. Modifications in the existing techniques may help in better clinical outcome with minimal risk of complications. This article provides an insight into some of these techniques along with a detailed account of the factors involved in tattoo removal. PMID:25949018

  18. Laparoscopic Removal of Gossypiboma

    Directory of Open Access Journals (Sweden)

    Zeki Özsoy

    2015-01-01

    Full Text Available Gossypiboma is defined as a mass caused by foreign body reaction developed around the retained surgical item in the operative area. When diagnosed, it should be removed in symptomatic patients. Minimal invasive surgery should be planned for the removal of the retained item. The number of cases treated by laparoscopic approach is rare in the literature. We present a case of forty-year-old woman referred to emergency room with acute abdomen diagnosed as gossypiboma and treated successfully with laparoscopic surgery.

  19. Contesting Citizenship: Comparative Analyses

    DEFF Research Database (Denmark)

    Siim, Birte; Squires, Judith

    2007-01-01

    importance of particularized experiences and multiple ineequality agendas). These developments shape the way citizenship is both practiced and analysed. Mapping neat citizenship modles onto distinct nation-states and evaluating these in relation to formal equality is no longer an adequate approach....... Comparative citizenship analyses need to be considered in relation to multipleinequalities and their intersections and to multiple governance and trans-national organisinf. This, in turn, suggests that comparative citizenship analysis needs to consider new spaces in which struggles for equal citizenship occur...

  20. Active Space Debris Removal System

    Directory of Open Access Journals (Sweden)

    Gabriele GUERRA

    2017-06-01

    Full Text Available Since the start of the space era, more than 5000 launches have been carried out, each carrying satellites for many disparate uses, such as Earth observation or communication. Thus, the space environment has become congested and the problem of space debris is now generating some concerns in the space community due to our long-lived belief that “space is big”. In the last few years, solutions to this problem have been proposed, one of those is Active Space Debris Removal: this method will reduce the increasing debris growth and permit future sustainable space activities. The main idea of the method proposed below is a drag augmentation system: use a system capable of putting an expanded foam on a debris which will increase the area-to-mass ratio to increase the natural atmospheric drag and solar pressure. The drag augmentation system proposed here requires a docking system; the debris will be pushed to its release height and then, after un-docking, an uncontrolled re-entry takes place ending with a burn up of the object and the foam in the atmosphere within a given time frame. The method requires an efficient way to change the orbit between two debris. The present paper analyses such a system in combination with an Electric Propulsion system, and emphasizes the choice of using two satellites to remove five effective rockets bodies debris within a year.

  1. Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

    2004-01-01

    Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

  2. Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Kilsdonk, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bremer, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Aeschlimann, R. W. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-06-01

    The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m2 to accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.

  3. Risico-analyse brandstofpontons

    NARCIS (Netherlands)

    Uijt de Haag P; Post J; LSO

    2001-01-01

    Voor het bepalen van de risico's van brandstofpontons in een jachthaven is een generieke risico-analyse uitgevoerd. Er is een referentiesysteem gedefinieerd, bestaande uit een betonnen brandstofponton met een relatief grote inhoud en doorzet. Aangenomen is dat de ponton gelegen is in een

  4. Fast multichannel analyser

    Energy Technology Data Exchange (ETDEWEB)

    Berry, A; Przybylski, M M; Sumner, I [Science Research Council, Daresbury (UK). Daresbury Lab.

    1982-10-01

    A fast multichannel analyser (MCA) capable of sampling at a rate of 10/sup 7/ s/sup -1/ has been developed. The instrument is based on an 8 bit parallel encoding analogue to digital converter (ADC) reading into a fast histogramming random access memory (RAM) system, giving 256 channels of 64 k count capacity. The prototype unit is in CAMAC format.

  5. A fast multichannel analyser

    International Nuclear Information System (INIS)

    Berry, A.; Przybylski, M.M.; Sumner, I.

    1982-01-01

    A fast multichannel analyser (MCA) capable of sampling at a rate of 10 7 s -1 has been developed. The instrument is based on an 8 bit parallel encoding analogue to digital converter (ADC) reading into a fast histogramming random access memory (RAM) system, giving 256 channels of 64 k count capacity. The prototype unit is in CAMAC format. (orig.)

  6. Removable pipeline plug

    International Nuclear Information System (INIS)

    Vassalotti, M.; Anastasi, F.

    1984-01-01

    A removable plugging device for a pipeline, and particularly for pressure testing a steam pipeline in a boiling water reactor, wherein an inflatable annular sealing member seals off the pipeline and characterized by radially movable shoes for holding the plug in place, each shoe being pivotally mounted for self-adjusting engagement with even an out-of-round pipeline interior

  7. Kidney removal - slideshow

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/presentations/100069.htm Kidney removal (nephrectomy) - series—Normal anatomy To use the sharing features on this page, please enable JavaScript. Go to slide 1 out of 5 Go to slide 2 out of ... to slide 5 out of 5 Overview The kidneys are paired organs that lie posterior to the ...

  8. Possible future HERA analyses

    International Nuclear Information System (INIS)

    Geiser, Achim

    2015-12-01

    A variety of possible future analyses of HERA data in the context of the HERA data preservation programme is collected, motivated, and commented. The focus is placed on possible future analyses of the existing ep collider data and their physics scope. Comparisons to the original scope of the HERA pro- gramme are made, and cross references to topics also covered by other participants of the workshop are given. This includes topics on QCD, proton structure, diffraction, jets, hadronic final states, heavy flavours, electroweak physics, and the application of related theory and phenomenology topics like NNLO QCD calculations, low-x related models, nonperturbative QCD aspects, and electroweak radiative corrections. Synergies with other collider programmes are also addressed. In summary, the range of physics topics which can still be uniquely covered using the existing data is very broad and of considerable physics interest, often matching the interest of results from colliders currently in operation. Due to well-established data and MC sets, calibrations, and analysis procedures the manpower and expertise needed for a particular analysis is often very much smaller than that needed for an ongoing experiment. Since centrally funded manpower to carry out such analyses is not available any longer, this contribution not only targets experienced self-funded experimentalists, but also theorists and master-level students who might wish to carry out such an analysis.

  9. Biomass feedstock analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wilen, C.; Moilanen, A.; Kurkela, E. [VTT Energy, Espoo (Finland). Energy Production Technologies

    1996-12-31

    The overall objectives of the project `Feasibility of electricity production from biomass by pressurized gasification systems` within the EC Research Programme JOULE II were to evaluate the potential of advanced power production systems based on biomass gasification and to study the technical and economic feasibility of these new processes with different type of biomass feed stocks. This report was prepared as part of this R and D project. The objectives of this task were to perform fuel analyses of potential woody and herbaceous biomasses with specific regard to the gasification properties of the selected feed stocks. The analyses of 15 Scandinavian and European biomass feed stock included density, proximate and ultimate analyses, trace compounds, ash composition and fusion behaviour in oxidizing and reducing atmospheres. The wood-derived fuels, such as whole-tree chips, forest residues, bark and to some extent willow, can be expected to have good gasification properties. Difficulties caused by ash fusion and sintering in straw combustion and gasification are generally known. The ash and alkali metal contents of the European biomasses harvested in Italy resembled those of the Nordic straws, and it is expected that they behave to a great extent as straw in gasification. Any direct relation between the ash fusion behavior (determined according to the standard method) and, for instance, the alkali metal content was not found in the laboratory determinations. A more profound characterisation of the fuels would require gasification experiments in a thermobalance and a PDU (Process development Unit) rig. (orig.) (10 refs.)

  10. Preliminary Design of Molecular Sieve for Removing Organic Iodide in Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Tong Kyu; Shin, So Eun; Lee, Byung Chul [Heungdeok IT Valley Bldg., Yongin (Korea, Republic of); Kim, Hong Hyun; Lee, Kyung Jun [Gemvax and KAEL Inc., Daejeon (Korea, Republic of)

    2014-05-15

    In this paper, to increase the DF for gaseous iodine species, especially organic iodide, molecular sieve filled by silver exchanged zeolites is proposed and designed preliminarily. Its aerodynamic analysis is also performed and presented. In order to increase the DF for gaseous organic iodide, deep-bed type molecular sieve was proposed and designed preliminarily. Total 1,620kg of silver exchanged zeolites were filled evenly in 10 beds of the molecular sieve. The safety factor in the case of 20m{sup 3}/s will be smaller than the counterpart of the standard case (6m{sup 3}/s). However, if the adsorption capacity of the zeolites is larger than 3.09mg/g when the residence time is 0.09 second, the designed molecular sieve can be used at 20m3/s of volumetric flow rate. The removal efficiency for organic iodide should be considered as well as economical aspects in the design of molecular sieve. In the event of nuclear power plant (NPP) severe accident, the nuclear reactor containment might suffer damage resulting from overpressure caused by decay heat. In order to prevent this containment damage, containment venting has been considered as one of effective methods. However, since vented gases contain radioactive fission products, they should be filtered to be released to environment. Generally, containment filtered venting system (CFVS) is installed on NPP to achieve this aim. Even though great amount of efforts have been devoted to developing the CFVS using various filtering methods, the decontaminant factor (DF) for radioactive gaseous iodide is still unsatisfactory while DFs for radioactive aerosols and elemental iodine are very high.

  11. Water Distribution and Removal Model

    International Nuclear Information System (INIS)

    Y. Deng; N. Chipman; E.L. Hardin

    2005-01-01

    The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD and R) Model; (2) EBS Physical and Chemical Environment (P and CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD and R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment

  12. Water Distribution and Removal Model

    Energy Technology Data Exchange (ETDEWEB)

    Y. Deng; N. Chipman; E.L. Hardin

    2005-08-26

    The design of the Yucca Mountain high level radioactive waste repository depends on the performance of the engineered barrier system (EBS). To support the total system performance assessment (TSPA), the Engineered Barrier System Degradation, Flow, and Transport Process Model Report (EBS PMR) is developed to describe the thermal, mechanical, chemical, hydrological, biological, and radionuclide transport processes within the emplacement drifts, which includes the following major analysis/model reports (AMRs): (1) EBS Water Distribution and Removal (WD&R) Model; (2) EBS Physical and Chemical Environment (P&CE) Model; (3) EBS Radionuclide Transport (EBS RNT) Model; and (4) EBS Multiscale Thermohydrologic (TH) Model. Technical information, including data, analyses, models, software, and supporting documents will be provided to defend the applicability of these models for their intended purpose of evaluating the postclosure performance of the Yucca Mountain repository system. The WD&R model ARM is important to the site recommendation. Water distribution and removal represents one component of the overall EBS. Under some conditions, liquid water will seep into emplacement drifts through fractures in the host rock and move generally downward, potentially contacting waste packages. After waste packages are breached by corrosion, some of this seepage water will contact the waste, dissolve or suspend radionuclides, and ultimately carry radionuclides through the EBS to the near-field host rock. Lateral diversion of liquid water within the drift will occur at the inner drift surface, and more significantly from the operation of engineered structures such as drip shields and the outer surface of waste packages. If most of the seepage flux can be diverted laterally and removed from the drifts before contacting the wastes, the release of radionuclides from the EBS can be controlled, resulting in a proportional reduction in dose release at the accessible environment. The purposes

  13. Removal of ciprofloxacin from water by birnessite

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Wei-Teh, E-mail: atwtj@mail.ncku.edu.tw [Department of Earth Sciences, National Cheng Kung University, Tainan 70101, Taiwan (China); Chang, Po-Hsiang; Wang, Ya-Siang; Tsai, Yolin; Jean, Jiin-Shuh [Department of Earth Sciences, National Cheng Kung University, Tainan 70101, Taiwan (China); Li, Zhaohui, E-mail: li@uwp.edu [Department of Earth Sciences, National Cheng Kung University, Tainan 70101, Taiwan (China); Department of Geosciences, University of Wisconsin – Parkside, Kenosha, WI 53144 (United States); Krukowski, Keith [Department of Geosciences, University of Wisconsin – Parkside, Kenosha, WI 53144 (United States)

    2013-04-15

    Highlights: ► Ciprofloxacin removal by birnessite was accompanied by interlayer cation exchange. ► Layer expansion and FTIR data suggested ciprofloxacin intercalation into birnessite. ► Adsorption capacity of ciprofloxacin into birnessite was limited by surface area. ► Birnessite in soil systems may provide host for ciprofloxacin accumulation. -- Abstract: With more pharmaceuticals and personal care products detected in the surface and waste waters, studies on interactions between these contaminants and soils or sediments have attracted great attention. In this study, the removal of ciprofloxacin (CIP), a fluoroquinolone antibiotic, by birnessite, a layered manganese oxide, in aqueous solution was investigated by batch studies supplemented by X-ray diffraction (XRD) and Fourier transform infrared analyses. Stoichiometric release of exchangeable cations accompanying CIP removal from water confirmed cation exchange as the major mechanism for CIP uptake by birnessite. Interlayer expansion after CIP adsorption on birnessite as revealed by XRD analyses indicated that intercalation contributed significantly to CIP uptake in addition to external surface adsorption. Correlation of CIP adsorption to specific surface area and cation exchange capacity suggested that the former was the limiting factor for CIP uptake. At the adsorption maximum, CIP molecules formed a monolayer on the birnessite surfaces. The adsorbed CIP could be partially removed using a cationic surfactant at a low initial concentration and mostly removed by AlCl{sub 3} at a higher initial concentration, which further supported the cation exchange mechanism for CIP removal by birnessite. The results indicated that the presence of layered Mn-oxide in the soil and waste water treatment systems may provide host for CIP accumulation.

  14. Removal of ciprofloxacin from water by birnessite

    International Nuclear Information System (INIS)

    Jiang, Wei-Teh; Chang, Po-Hsiang; Wang, Ya-Siang; Tsai, Yolin; Jean, Jiin-Shuh; Li, Zhaohui; Krukowski, Keith

    2013-01-01

    Highlights: ► Ciprofloxacin removal by birnessite was accompanied by interlayer cation exchange. ► Layer expansion and FTIR data suggested ciprofloxacin intercalation into birnessite. ► Adsorption capacity of ciprofloxacin into birnessite was limited by surface area. ► Birnessite in soil systems may provide host for ciprofloxacin accumulation. -- Abstract: With more pharmaceuticals and personal care products detected in the surface and waste waters, studies on interactions between these contaminants and soils or sediments have attracted great attention. In this study, the removal of ciprofloxacin (CIP), a fluoroquinolone antibiotic, by birnessite, a layered manganese oxide, in aqueous solution was investigated by batch studies supplemented by X-ray diffraction (XRD) and Fourier transform infrared analyses. Stoichiometric release of exchangeable cations accompanying CIP removal from water confirmed cation exchange as the major mechanism for CIP uptake by birnessite. Interlayer expansion after CIP adsorption on birnessite as revealed by XRD analyses indicated that intercalation contributed significantly to CIP uptake in addition to external surface adsorption. Correlation of CIP adsorption to specific surface area and cation exchange capacity suggested that the former was the limiting factor for CIP uptake. At the adsorption maximum, CIP molecules formed a monolayer on the birnessite surfaces. The adsorbed CIP could be partially removed using a cationic surfactant at a low initial concentration and mostly removed by AlCl 3 at a higher initial concentration, which further supported the cation exchange mechanism for CIP removal by birnessite. The results indicated that the presence of layered Mn-oxide in the soil and waste water treatment systems may provide host for CIP accumulation

  15. Pneumatic soil removal tool

    International Nuclear Information System (INIS)

    Neuhaus, J.E.

    1992-01-01

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw. 3 figs

  16. Investigations in gallium removal

    Energy Technology Data Exchange (ETDEWEB)

    Philip, C.V.; Pitt, W.W. [Texas A and M Univ., College Station, TX (United States); Beard, C.A. [Amarillo National Resource Center for Plutonium, TX (United States)

    1997-11-01

    Gallium present in weapons plutonium must be removed before it can be used for the production of mixed-oxide (MOX) nuclear reactor fuel. The main goal of the preliminary studies conducted at Texas A and M University was to assist in the development of a thermal process to remove gallium from a gallium oxide/plutonium oxide matrix. This effort is being conducted in close consultation with the Los Alamos National Laboratory (LANL) personnel involved in the development of this process for the US Department of Energy (DOE). Simple experiments were performed on gallium oxide, and cerium-oxide/gallium-oxide mixtures, heated to temperatures ranging from 700--900 C in a reducing environment, and a method for collecting the gallium vapors under these conditions was demonstrated.

  17. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  18. Measures for removing hydrogen

    International Nuclear Information System (INIS)

    Baukal, W.; Koehling, A.; Langer, G.; Poeschel, E.

    1984-01-01

    Basis for the investigation is a 1300-MW-PWR. The evolution of hydrogen was studied in design-basis and three hypothetical accident scenarios, the loss-of-coolant accident, the failure of emergency cooling system and core meltdown. It was shown that in the case of release rates of 4m 3 H 2 /h, the known post-accident hydrogen removal systems can be used and at medium rates up to 80 m 3 H 2 /h recombines of nuclear and non-nuclear industries are suitable under certain conditions. In the case of larger release rates it appears useful to apply a small recombiner of the type of the post-accident hydrogen removal system combined with an other hydrogen countermeasures. Recommendations are being made for the installation of an accident-proof hydrogen measuring system. (DG) [de

  19. Pneumatic soil removal tool

    Science.gov (United States)

    Neuhaus, John E.

    1992-01-01

    A soil removal tool is provided for removing radioactive soil, rock and other debris from the bottom of an excavation, while permitting the operator to be located outside of a containment for that excavation. The tool includes a fixed jaw, secured to one end of an elongate pipe, which cooperates with a movable jaw pivotably mounted on the pipe. Movement of the movable jaw is controlled by a pneumatic cylinder mounted on the pipe. The actuator rod of the pneumatic cylinder is connected to a collar which is slidably mounted on the pipe and forms part of the pivotable mounting assembly for the movable jaw. Air is supplied to the pneumatic cylinder through a handle connected to the pipe, under the control of an actuator valve mounted on the handle, to provide movement of the movable jaw.

  20. Investigations in gallium removal

    International Nuclear Information System (INIS)

    Philip, C.V.; Pitt, W.W.; Beard, C.A.

    1997-11-01

    Gallium present in weapons plutonium must be removed before it can be used for the production of mixed-oxide (MOX) nuclear reactor fuel. The main goal of the preliminary studies conducted at Texas A and M University was to assist in the development of a thermal process to remove gallium from a gallium oxide/plutonium oxide matrix. This effort is being conducted in close consultation with the Los Alamos National Laboratory (LANL) personnel involved in the development of this process for the US Department of Energy (DOE). Simple experiments were performed on gallium oxide, and cerium-oxide/gallium-oxide mixtures, heated to temperatures ranging from 700--900 C in a reducing environment, and a method for collecting the gallium vapors under these conditions was demonstrated

  1. Removing water from gels

    International Nuclear Information System (INIS)

    Lane, E.S.; Winter, J.A.

    1982-01-01

    Water is removed from a gel material by contacting the gel material with an organic liquid and contacting the organic liquid with a gas such that water is taken up by the gas. The invention, in one embodiment, may be used to dry gel materials whilst maintaining an open porous network therein. In one example, the invention is applied to gel precipitated spheres containing uranium and plutonium. (author)

  2. Power plant removal costs

    International Nuclear Information System (INIS)

    Ferguson, J.S.

    1998-01-01

    The financial, regulatory and political significance of the estimated high removal costs of nuclear power plants has generated considerable interest in recent years, and the political significance has resulted in the Nuclear Regulatory Commission (NRC) eliminating the use of conventional depreciation accounting for the decontamination portion of the removal (decommissioning). While nuclear plant licensees are not precluded from utilizing conventional depreciation accounting for the demolition of non-radioactive structures and site restoration, state and federal utility regulators have not been favorably inclined to requests for this distinction. The realization that steam-generating units will be more expensive to remove, relative to their original cost, predates the realization that nuclear units will be expensive. However, the nuclear issues have overshadowed this realization, but are unlikely to continue to do so. Numerous utilities have prepared cost estimates for steam generating units, and this presentation discusses the implications of a number of such estimates that are a matter of public record. The estimates cover nearly 400 gas, oil, coal and lignite generating units. The earliest estimate was made in 1978, and for analysis purposes the author has segregated them between gas and oil units, and coal and lignite units

  3. AMS analyses at ANSTO

    Energy Technology Data Exchange (ETDEWEB)

    Lawson, E.M. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia). Physics Division

    1998-03-01

    The major use of ANTARES is Accelerator Mass Spectrometry (AMS) with {sup 14}C being the most commonly analysed radioisotope - presently about 35 % of the available beam time on ANTARES is used for {sup 14}C measurements. The accelerator measurements are supported by, and dependent on, a strong sample preparation section. The ANTARES AMS facility supports a wide range of investigations into fields such as global climate change, ice cores, oceanography, dendrochronology, anthropology, and classical and Australian archaeology. Described here are some examples of the ways in which AMS has been applied to support research into the archaeology, prehistory and culture of this continent`s indigenous Aboriginal peoples. (author)

  4. AMS analyses at ANSTO

    International Nuclear Information System (INIS)

    Lawson, E.M.

    1998-01-01

    The major use of ANTARES is Accelerator Mass Spectrometry (AMS) with 14 C being the most commonly analysed radioisotope - presently about 35 % of the available beam time on ANTARES is used for 14 C measurements. The accelerator measurements are supported by, and dependent on, a strong sample preparation section. The ANTARES AMS facility supports a wide range of investigations into fields such as global climate change, ice cores, oceanography, dendrochronology, anthropology, and classical and Australian archaeology. Described here are some examples of the ways in which AMS has been applied to support research into the archaeology, prehistory and culture of this continent's indigenous Aboriginal peoples. (author)

  5. Removal of chloride from MSWI fly ash.

    Science.gov (United States)

    Chen, Wei-Sheng; Chang, Fang-Chih; Shen, Yun-Hwei; Tsai, Min-Shing; Ko, Chun-Han

    2012-10-30

    The high levels of alkali chloride and soluble metal salts present in MSWI fly ash is worth noting for their impact on the environment. In addition, the recycling or reuse of fly ash has become an issue because of limited landfill space. The chloride content in fly ash limits its application as basis for construction materials. Water-soluble chlorides such as potassium chloride (KCl), sodium chloride (NaCl), and calcium chloride hydrate (CaCl(2) · 2H(2)O) in fly ash are easily washed away. However, calcium chloride hydroxide (Ca(OH)Cl) might not be easy to leach away at room temperature. The roasting and washing-flushing processes were applied to remove chloride content in this study. Additionally, air and CO(2) were introduced into the washing process to neutralize the hazardous nature of chlorides. In comparison with the water flushing process, the roasting process is more efficient in reducing the process of solid-liquid separation and drying for the reuse of Cl-removed fly ash particles. In several roasting experiments, the removal of chloride content from fly ash at 1050°C for 3h showed the best results (83% chloride removal efficiency). At a solid to liquid ratio of 1:10 the water-flushing process can almost totally remove water-soluble chloride (97% chloride removal efficiency). Analyses of mineralogical change also prove the efficiency of the fly ash roasting and washing mechanisms for chloride removal. Copyright © 2012 Elsevier B.V. All rights reserved.

  6. Analyses of MHD instabilities

    International Nuclear Information System (INIS)

    Takeda, Tatsuoki

    1985-01-01

    In this article analyses of the MHD stabilities which govern the global behavior of a fusion plasma are described from the viewpoint of the numerical computation. First, we describe the high accuracy calculation of the MHD equilibrium and then the analysis of the linear MHD instability. The former is the basis of the stability analysis and the latter is closely related to the limiting beta value which is a very important theoretical issue of the tokamak research. To attain a stable tokamak plasma with good confinement property it is necessary to control or suppress disruptive instabilities. We, next, describe the nonlinear MHD instabilities which relate with the disruption phenomena. Lastly, we describe vectorization of the MHD codes. The above MHD codes for fusion plasma analyses are relatively simple though very time-consuming and parts of the codes which need a lot of CPU time concentrate on a small portion of the codes, moreover, the codes are usually used by the developers of the codes themselves, which make it comparatively easy to attain a high performance ratio on the vector processor. (author)

  7. Uncertainty Analyses and Strategy

    International Nuclear Information System (INIS)

    Kevin Coppersmith

    2001-01-01

    The DOE identified a variety of uncertainties, arising from different sources, during its assessment of the performance of a potential geologic repository at the Yucca Mountain site. In general, the number and detail of process models developed for the Yucca Mountain site, and the complex coupling among those models, make the direct incorporation of all uncertainties difficult. The DOE has addressed these issues in a number of ways using an approach to uncertainties that is focused on producing a defensible evaluation of the performance of a potential repository. The treatment of uncertainties oriented toward defensible assessments has led to analyses and models with so-called ''conservative'' assumptions and parameter bounds, where conservative implies lower performance than might be demonstrated with a more realistic representation. The varying maturity of the analyses and models, and uneven level of data availability, result in total system level analyses with a mix of realistic and conservative estimates (for both probabilistic representations and single values). That is, some inputs have realistically represented uncertainties, and others are conservatively estimated or bounded. However, this approach is consistent with the ''reasonable assurance'' approach to compliance demonstration, which was called for in the U.S. Nuclear Regulatory Commission's (NRC) proposed 10 CFR Part 63 regulation (64 FR 8640 [DIRS 101680]). A risk analysis that includes conservatism in the inputs will result in conservative risk estimates. Therefore, the approach taken for the Total System Performance Assessment for the Site Recommendation (TSPA-SR) provides a reasonable representation of processes and conservatism for purposes of site recommendation. However, mixing unknown degrees of conservatism in models and parameter representations reduces the transparency of the analysis and makes the development of coherent and consistent probability statements about projected repository

  8. A simple beam analyser

    International Nuclear Information System (INIS)

    Lemarchand, G.

    1977-01-01

    (ee'p) experiments allow to measure the missing energy distribution as well as the momentum distribution of the extracted proton in the nucleus versus the missing energy. Such experiments are presently conducted on SACLAY's A.L.S. 300 Linac. Electrons and protons are respectively analysed by two spectrometers and detected in their focal planes. Counting rates are usually low and include time coincidences and accidentals. Signal-to-noise ratio is dependent on the physics of the experiment and the resolution of the coincidence, therefore it is mandatory to get a beam current distribution as flat as possible. Using new technologies has allowed to monitor in real time the behavior of the beam pulse and determine when the duty cycle can be considered as being good with respect to a numerical basis

  9. EEG analyses with SOBI.

    Energy Technology Data Exchange (ETDEWEB)

    Glickman, Matthew R.; Tang, Akaysha (University of New Mexico, Albuquerque, NM)

    2009-02-01

    The motivating vision behind Sandia's MENTOR/PAL LDRD project has been that of systems which use real-time psychophysiological data to support and enhance human performance, both individually and of groups. Relevant and significant psychophysiological data being a necessary prerequisite to such systems, this LDRD has focused on identifying and refining such signals. The project has focused in particular on EEG (electroencephalogram) data as a promising candidate signal because it (potentially) provides a broad window on brain activity with relatively low cost and logistical constraints. We report here on two analyses performed on EEG data collected in this project using the SOBI (Second Order Blind Identification) algorithm to identify two independent sources of brain activity: one in the frontal lobe and one in the occipital. The first study looks at directional influences between the two components, while the second study looks at inferring gender based upon the frontal component.

  10. Pathway-based analyses.

    Science.gov (United States)

    Kent, Jack W

    2016-02-03

    New technologies for acquisition of genomic data, while offering unprecedented opportunities for genetic discovery, also impose severe burdens of interpretation and penalties for multiple testing. The Pathway-based Analyses Group of the Genetic Analysis Workshop 19 (GAW19) sought reduction of multiple-testing burden through various approaches to aggregation of highdimensional data in pathways informed by prior biological knowledge. Experimental methods testedincluded the use of "synthetic pathways" (random sets of genes) to estimate power and false-positive error rate of methods applied to simulated data; data reduction via independent components analysis, single-nucleotide polymorphism (SNP)-SNP interaction, and use of gene sets to estimate genetic similarity; and general assessment of the efficacy of prior biological knowledge to reduce the dimensionality of complex genomic data. The work of this group explored several promising approaches to managing high-dimensional data, with the caveat that these methods are necessarily constrained by the quality of external bioinformatic annotation.

  11. Analysing Access Control Specifications

    DEFF Research Database (Denmark)

    Probst, Christian W.; Hansen, René Rydhof

    2009-01-01

    When prosecuting crimes, the main question to answer is often who had a motive and the possibility to commit the crime. When investigating cyber crimes, the question of possibility is often hard to answer, as in a networked system almost any location can be accessed from almost anywhere. The most...... common tool to answer this question, analysis of log files, faces the problem that the amount of logged data may be overwhelming. This problems gets even worse in the case of insider attacks, where the attacker’s actions usually will be logged as permissible, standard actions—if they are logged at all....... Recent events have revealed intimate knowledge of surveillance and control systems on the side of the attacker, making it often impossible to deduce the identity of an inside attacker from logged data. In this work we present an approach that analyses the access control configuration to identify the set...

  12. Network class superposition analyses.

    Directory of Open Access Journals (Sweden)

    Carl A B Pearson

    Full Text Available Networks are often used to understand a whole system by modeling the interactions among its pieces. Examples include biomolecules in a cell interacting to provide some primary function, or species in an environment forming a stable community. However, these interactions are often unknown; instead, the pieces' dynamic states are known, and network structure must be inferred. Because observed function may be explained by many different networks (e.g., ≈ 10(30 for the yeast cell cycle process, considering dynamics beyond this primary function means picking a single network or suitable sample: measuring over all networks exhibiting the primary function is computationally infeasible. We circumvent that obstacle by calculating the network class ensemble. We represent the ensemble by a stochastic matrix T, which is a transition-by-transition superposition of the system dynamics for each member of the class. We present concrete results for T derived from boolean time series dynamics on networks obeying the Strong Inhibition rule, by applying T to several traditional questions about network dynamics. We show that the distribution of the number of point attractors can be accurately estimated with T. We show how to generate Derrida plots based on T. We show that T-based Shannon entropy outperforms other methods at selecting experiments to further narrow the network structure. We also outline an experimental test of predictions based on T. We motivate all of these results in terms of a popular molecular biology boolean network model for the yeast cell cycle, but the methods and analyses we introduce are general. We conclude with open questions for T, for example, application to other models, computational considerations when scaling up to larger systems, and other potential analyses.

  13. Removable molar power arm

    Directory of Open Access Journals (Sweden)

    Raj Kumar Verma

    2013-01-01

    Full Text Available Attachment of force elements from the gingival hook of maxillary molar tubes during the retraction of the anterior teeth is very common in orthodontic practice. As the line of force passes below the center of resistance (CR of molar, it results its mesial tipping and also anchorage loss. To overcome this problem, the line of force should pass along the CR of molar. This article highlights a method to overcome this problem by attaching a removable power arm to the headgear tube of molar tube during the retraction of the anterior teeth.

  14. Study of a micro-sublimation apparatus with removal of the vapours by pumping; application to the analysis of fluorinated products (1963); Etude d'un appareillage de microsublimation avec entrainement des vapeurs par pompage et application a l'analyse des produits fluores (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Delvalle, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    Micro-sublimation analysis presents definite advantages both from the qualitative and quantitative points of view. An automatic micro-sublimation analysis apparatus has been developed for the analysis of fluorinated products (ClF{sub 3}, HF, UF{sub 6}, etc.) but this is only one particular application of a method which has a far wider field of possible applications. We give first the most favorable conditions for the operation of such an apparatus. These conditions are the use of a detector which is linear and independent of the nature of the gas, the flow of the sublimed vapours in the conditions of molecular flow, and finally a reproducible and linear re-heating of the separating trap. The apparatus thus built has the advantage of yielding any analysis without prior calibration. It also makes possible the easy identification of an unknown product by the determination of its vapour pressure curve and its molecular weight. The analysis of fluorinated products with this apparatus has shown that the experimental results agree well with what is expected. (author) [French] L'analyse par microsublimation presente un interet certain tant au point de vue qualitatif que quantitatif. Nous avons mis au point un appareillage automatique d'analyse par microsublimation pour l'analyse des produits fluores (CIF{sub 3}, HF, UF{sub 6}, etc.) mais il ne s'agit que de l'application a un cas particulier d'une methode ayant un champ d'applications bien plus vaste. Nous exposons tout d'abord les conditions les plus favorables au bon fonctionnement d'un tel appareil. Ces conditions sont l'emploi d'un detecteur lineaire et independant de la nature du gaz, l'ecoulement des vapeurs sublimees en regime moleculaire et enfin un rechauffage reproductible et lineaire du piege separateur. L'appareil ainsi realise presente l'avantage d'effectuer une analyse quelconque sans etalonnage prealable. Il permet en outre d'identifier aisement un corps inconnu par la determination de sa courbe de tension

  15. Mercury removal sorbents

    Science.gov (United States)

    Alptekin, Gokhan

    2016-03-29

    Sorbents and methods of using them for removing mercury from flue gases over a wide range of temperatures are disclosed. Sorbent materials of this invention comprise oxy- or hydroxyl-halogen (chlorides and bromides) of manganese, copper and calcium as the active phase for Hg.sup.0 oxidation, and are dispersed on a high surface porous supports. In addition to the powder activated carbons (PACs), this support material can be comprised of commercial ceramic supports such as silica (SiO.sub.2), alumina (Al.sub.2O.sub.3), zeolites and clays. The support material may also comprise of oxides of various metals such as iron, manganese, and calcium. The non-carbon sorbents of the invention can be easily injected into the flue gas and recovered in the Particulate Control Device (PCD) along with the fly ash without altering the properties of the by-product fly ash enabling its use as a cement additive. Sorbent materials of this invention effectively remove both elemental and oxidized forms of mercury from flue gases and can be used at elevated temperatures. The sorbent combines an oxidation catalyst and a sorbent in the same particle to both oxidize the mercury and then immobilize it.

  16. Seismic fragility analyses

    International Nuclear Information System (INIS)

    Kostov, Marin

    2000-01-01

    In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, β E determined by the relation: β E ∫[d[β(x)]/dx]P(flx)dx. β(x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(fI x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard β(x) and the fragility P(fl x). The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined as probabilistic and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency. The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinary structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper. After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The

  17. Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, C. A. [Argonne National Lab., IL (United States); Ariman, T. [Univ. of Notre Dame, IN (United States); Pierce, R. D.; Pedersen, D. R. [Argonne National Lab., IL (United States)

    1977-07-01

    The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. The cup principal components are: 1. A 38 mm diameter tungsten spike which provides initial fuel quenching and prevents fuel boiling, 2. A 73 mm inside diameter tungsten liner to isolate the support vessel from the molten material high initial temperature, 3. An insulator which is an expedient for extending the experiment time, and 4. An Inconel 625 vessel which provides the structural support to withstand the thermal and pressure stresses. The spike, liner, and insulator are supported by a hemispherical tungsten end cap which fits inside the hemispherical bottom of the support vessel. This vessel is attached to the 316 stainless steel test train with an Inconel 750 wire-formed retaining ring. Thermal analyses were performed with the Argonne-modified version of the general heat transfer code THTB, based on the instantaneous addition of 3200/sup 0/K molten fuel with a decay heat of 9 W/gm and 1920/sup 0/K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The most severe heating occurs when there is no sodium flow outside the cup. For this case the sodium boils (approximately 1200/sup 0/K) and the Inconel vessel and tungsten liner temperatures are approximately 1250/sup 0/K and 2420/sup 0/K, respectively.

  18. Website-analyse

    DEFF Research Database (Denmark)

    Thorlacius, Lisbeth

    2009-01-01

    eller blindgyder, når han/hun besøger sitet. Studier i design og analyse af de visuelle og æstetiske aspekter i planlægning og brug af websites har imidlertid kun i et begrænset omfang været under reflektorisk behandling. Det er baggrunden for dette kapitel, som indleder med en gennemgang af æstetikkens......Websitet er i stigende grad det foretrukne medie inden for informationssøgning,virksomhedspræsentation, e-handel, underholdning, undervisning og social kontakt. I takt med denne voksende mangfoldighed af kommunikationsaktiviteter på nettet, er der kommet mere fokus på at optimere design og...... planlægning af de funktionelle og indholdsmæssige aspekter ved websites. Der findes en stor mængde teori- og metodebøger, som har specialiseret sig i de tekniske problemstillinger i forbindelse med interaktion og navigation, samt det sproglige indhold på websites. Den danske HCI (Human Computer Interaction...

  19. A channel profile analyser

    International Nuclear Information System (INIS)

    Gobbur, S.G.

    1983-01-01

    It is well understood that due to the wide band noise present in a nuclear analog-to-digital converter, events at the boundaries of adjacent channels are shared. It is a difficult and laborious process to exactly find out the shape of the channels at the boundaries. A simple scheme has been developed for the direct display of channel shape of any type of ADC on a cathode ray oscilliscope display. This has been accomplished by sequentially incrementing the reference voltage of a precision pulse generator by a fraction of a channel and storing ADC data in alternative memory locations of a multichannel pulse height analyser. Alternative channels are needed due to the sharing at the boundaries of channels. In the flat region of the profile alternate memory locations are channels with zero counts and channels with the full scale counts. At the boundaries all memory locations will have counts. The shape of this is a direct display of the channel boundaries. (orig.)

  20. 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

    Energy Technology Data Exchange (ETDEWEB)

    Venker, Jeanne

    2015-03-31

    heat removal system. A first analysis of the system revealed the ability to remove the decay heat over more than 72 hours, even for combined station blackout and loss of ultimate heat sink scenarios. In addition, the simulations exposed an interaction between the retrofitted and already existing systems. Parameters, which influence the operation of the self-propelling heat removal system, have been identified and summarized in set of prerequisites. The simulations indicate the potential of the system to serve as a diverse heat removal system for existing boiling water reactors.

  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

    . A first analysis of the system revealed the ability to remove the decay heat over more than 72 hours, even for combined station blackout and loss of ultimate heat sink scenarios. In addition, the simulations exposed an interaction between the retrofitted and already existing systems. Parameters, which influence the operation of the self-propelling heat removal system, have been identified and summarized in set of prerequisites. The simulations indicate the potential of the system to serve as a diverse heat removal system for existing boiling water reactors.

  2. Black hole hair removal

    International Nuclear Information System (INIS)

    Banerjee, Nabamita; Mandal, Ipsita; Sen, Ashoke

    2009-01-01

    Macroscopic entropy of an extremal black hole is expected to be determined completely by its near horizon geometry. Thus two black holes with identical near horizon geometries should have identical macroscopic entropy, and the expected equality between macroscopic and microscopic entropies will then imply that they have identical degeneracies of microstates. An apparent counterexample is provided by the 4D-5D lift relating BMPV black hole to a four dimensional black hole. The two black holes have identical near horizon geometries but different microscopic spectrum. We suggest that this discrepancy can be accounted for by black hole hair - degrees of freedom living outside the horizon and contributing to the degeneracies. We identify these degrees of freedom for both the four and the five dimensional black holes and show that after their contributions are removed from the microscopic degeneracies of the respective systems, the result for the four and five dimensional black holes match exactly.

  3. Radioactive waste removing device

    International Nuclear Information System (INIS)

    Sakai, Takuhiko.

    1982-01-01

    Purpose: To cleanup primary coolants for LMFBR type reactors by magnetically generating a high speed rotational flow in the flow of liquid metal, and adsorbing radioactive corrosion products and fission products onto capturing material of a complicated shape. Constitution: Three-phase AC coils for generating a rotational magnetic field are provided to the outside of a container through which liquid sodium is passed to thereby generate a high speed rotational stream in the liquid sodium flowing into the container. A radioactive substance capturing material made of a metal plate such as of nickel and stainless steel in the corrugated shape with shape edges is secured within a flow channel. Magnetic field at a great slope is generated in the flow channel by the capturing material to adsorb radioactive corrosion products and fission products present in the liquid sodium onto the capturing material and removing therefrom. This enables to capture the ferri-magnetic impurities by adsorption. (Moriyama, K.)

  4. Tritium effluent removal system

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Gibbs, G.E.

    1978-01-01

    An air detritiation system has been developed and is in routine use for removing tritium and tritiated compounds from glovebox effluent streams before they are released to the atmosphere. The system is also used, in combination with temporary enclosures, to contain and decontaminate airborne releases resulting from the opening of tritium containment systems during maintenance and repair operations. This detritiation system, which services all the tritium handling areas at Mound Facility, has played an important role in reducing effluents and maintaining them at 2 percent of the level of 8 y ago. The system has a capacity of 1.7 m 3 /min and has operated around the clock for several years. A refrigerated in-line filtration system removes water, mercury, or pump oil and other organics from gaseous waste streams. The filtered waste stream is then heated and passed through two different types of oxidizing beds; the resulting tritiated water is collected on molecular sieve dryer beds. Liquids obtained from regenerating the dryers and from the refrigerated filtration system are collected and transferred to a waste solidification and packaging station. Component redundancy and by-pass capabilities ensure uninterrupted system operation during maintenance. When processing capacity is exceeded, an evacuated storage tank of 45 m 3 is automatically opened to the inlet side of the system. The gaseous effluent from the system is monitored for tritium content and recycled or released directly to the stack. The average release is less than 1 Ci/day. The tritium effluent can be reduced by isotopically swamping the tritium; this is accomplished by adding hydrogen prior to the oxidizer beds, or by adding water to the stream between the two final dryer beds

  5. Design and transient analyses of passive emergency feedwater system of CPR1000. Part 1. Air cooling condition

    International Nuclear Information System (INIS)

    Zhang Yapei; Qiu Suizheng; Su Guanghui; Tian Wenxi; Cao Jianhua; Lu Donghua; Fu Xiangang

    2011-01-01

    The steam generator secondary passive emergency feedwater system is a new design for traditional generation Ⅱ + reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively. (author)

  6. NOAA's National Snow Analyses

    Science.gov (United States)

    Carroll, T. R.; Cline, D. W.; Olheiser, C. M.; Rost, A. A.; Nilsson, A. O.; Fall, G. M.; Li, L.; Bovitz, C. T.

    2005-12-01

    NOAA's National Operational Hydrologic Remote Sensing Center (NOHRSC) routinely ingests all of the electronically available, real-time, ground-based, snow data; airborne snow water equivalent data; satellite areal extent of snow cover information; and numerical weather prediction (NWP) model forcings for the coterminous U.S. The NWP model forcings are physically downscaled from their native 13 km2 spatial resolution to a 1 km2 resolution for the CONUS. The downscaled NWP forcings drive an energy-and-mass-balance snow accumulation and ablation model at a 1 km2 spatial resolution and at a 1 hour temporal resolution for the country. The ground-based, airborne, and satellite snow observations are assimilated into the snow model's simulated state variables using a Newtonian nudging technique. The principle advantages of the assimilation technique are: (1) approximate balance is maintained in the snow model, (2) physical processes are easily accommodated in the model, and (3) asynoptic data are incorporated at the appropriate times. The snow model is reinitialized with the assimilated snow observations to generate a variety of snow products that combine to form NOAA's NOHRSC National Snow Analyses (NSA). The NOHRSC NSA incorporate all of the available information necessary and available to produce a "best estimate" of real-time snow cover conditions at 1 km2 spatial resolution and 1 hour temporal resolution for the country. The NOHRSC NSA consist of a variety of daily, operational, products that characterize real-time snowpack conditions including: snow water equivalent, snow depth, surface and internal snowpack temperatures, surface and blowing snow sublimation, and snowmelt for the CONUS. The products are generated and distributed in a variety of formats including: interactive maps, time-series, alphanumeric products (e.g., mean areal snow water equivalent on a hydrologic basin-by-basin basis), text and map discussions, map animations, and quantitative gridded products

  7. Analysis of microplastics and their removal from water

    OpenAIRE

    Oladejo, Abiola

    2017-01-01

    Removal of microplastics from water was studied using extraction with oil. The aim of the thesis was to remove microplastics from water using an organic medium, and to analyse the amount of microplastics in the media involved. The separation of different microplastic types was done by conducting experiments in the laboratory. The microplastics were made by grinding and sieving plastics with a grinding machine before adding them to water and oil, which serves as the organic medium. Tw...

  8. Phosphate removal from digested sludge supernatant using modified fly ash.

    Science.gov (United States)

    Xu, Ke; Deng, Tong; Liu, Juntan; Peng, Weigong

    2012-05-01

    The removal of phosphate in digested sludge supernatant by modified coal fly ash was investigated in this study. Modification of the fly ash by the addition of sulfuric acid could significantly enhance its immobilization ability. The experimental results also showed that adsorption of phosphate by the modified fly ash was rapid with the removal percentage of phosphate reaching an equilibrium of 98.62% in less than 5 minutes. The optimum pH for phosphate removal was 9 and the removal percentage increased with increasing adsorbent dosage. The effect of temperature on phosphate removal efficiency was not significant from 20 to 40 degrees C. X-ray diffraction and scanning electron microscope analyses showed that phosphate formed an amorphous precipitate with water-soluble calcium, aluminum, and iron ions in the modified fly ash.

  9. Unilateral removable partial dentures.

    Science.gov (United States)

    Goodall, W A; Greer, A C; Martin, N

    2017-01-27

    Removable partial dentures (RPDs) are widely used to replace missing teeth in order to restore both function and aesthetics for the partially dentate patient. Conventional RPD design is frequently bilateral and consists of a major connector that bridges both sides of the arch. Some patients cannot and will not tolerate such an extensive appliance. For these patients, bridgework may not be a predictable option and it is not always possible to provide implant-retained restorations. This article presents unilateral RPDs as a potential treatment modality for such patients and explores indications and contraindications for their use, including factors relating to patient history, clinical presentation and patient wishes. Through case examples, design, material and fabrication considerations will be discussed. While their use is not widespread, there are a number of patients who benefit from the provision of unilateral RPDs. They are a useful treatment to have in the clinician's armamentarium, but a highly-skilled dental team and a specific patient presentation is required in order for them to be a reasonable and predictable prosthetic option.

  10. Possibilities of hydrogen removal

    International Nuclear Information System (INIS)

    Langer, G.; Koehling, A.; Nikodem, H.

    1982-12-01

    In the event of hypothetical severe accidents in light-water reactors, considerable amounts of hydrogen may be produced and released into the containment. Combustion of the hydrogen may jeopardize the integrity of the containment. The study reported here aimed to identify methods to mitigate the hydrogen problem. These methods should either prevent hydrogen combustion, or limit its effects. The following methods have been investigated: pre-inerting; chemical oxygen absorption; removal of oxygen by combustion; post-inerting with N 2 , CO 2 , or halon; aqueous foam; water fog; deliberate ignition; containment purging; and containment venting. The present state of the art in both nuclear and non-nuclear facilities, has been identified. The assessment of the methods was based on accident scenarios assuming significant release of hydrogen and the spectrum of requirements derived from these scenarios was used to determine the advantages and drawbacks of the various methods, assuming their application in a pressurized-water reactor of German design. (orig.) [de

  11. Removal of unwanted fluid

    Science.gov (United States)

    Subudhi, Sudhakar; Sreenivas, K. R.; Arakeri, Jaywant H.

    2013-01-01

    This work is concerned with the removal of unwanted fluid through the source-sink pair. The source consists of fluid issuing out of a nozzle in the form of a jet and the sink is a pipe that is kept some distance from the source pipe. Of concern is the percentage of source fluid sucked through the sink. The experiments have been carried in a large glass water tank. The source nozzle diameter is 6 mm and the sink pipe diameter is either 10 or 20 mm. The horizontal and vertical separations and angles between these source and sink pipes are adjustable. The flow was visualized using KMnO4 dye, planer laser induced fluorescence and particle streak photographs. To obtain the effectiveness (that is percentage of source fluid entering the sink pipe), titration method is used. The velocity profiles with and without the sink were obtained using particle image velocimetry. The sink flow rate to obtain a certain effectiveness increase dramatically with lateral separation. The sink diameter and the angle between source and the sink axes don't influence effectiveness as much as the lateral separation.

  12. Iodine removing means

    International Nuclear Information System (INIS)

    Takeshima, Masaki.

    1975-01-01

    Object: To employ exhaust gas from an incinerator to effect regeneration of an adsorbent such as active carbon which has adsorbed a radioactive gas such as iodine contained in the ventilating system exhaust gas of a boiling water reactor power plant. Structure: Radioactive exhaust gas such as iodine, xenon and krypton is led to an active carbon adsorbing means for removal through adsorption. When the adsorbing function of the active carbon adsorption means is reduced, the exhaust gas discharged from the incinerator is cooled down to 300 0 C and then caused to flow into the active carbon layer, and after depriving it of sulfur dioxide gas, oxides of nitrogen, daughter nuclides resulting from attenuation of radioactive gas and so forth, these being adsorbed by the carbon active layer, it is led again to the incinerator, whereby the radioactivity accompanying the regenerated gas is sealed as ash within the incinerator. Further, similarly accompanying fine active carbon particles and the like are utilized as a heat source for the incinerator. (Kamimura, M.)

  13. ARSENIC REMOVAL FROM DRINKING WATER BY IRON REMOVAL PLANTS

    Science.gov (United States)

    This report documents a long term performance study of two iron removal water treatment plants to remove arsenic from drinking water sources. Performance information was collected from one system located in midwest for one full year and at the second system located in the farwest...

  14. Wholesale debris removal from LEO

    Science.gov (United States)

    Levin, Eugene; Pearson, Jerome; Carroll, Joseph

    2012-04-01

    Recent advances in electrodynamic propulsion make it possible to seriously consider wholesale removal of large debris from LEO for the first time since the beginning of the space era. Cumulative ranking of large groups of the LEO debris population and general limitations of passive drag devices and rocket-based removal systems are analyzed. A candidate electrodynamic debris removal system is discussed that can affordably remove all debris objects over 2 kg from LEO in 7 years. That means removing more than 99% of the collision-generated debris potential in LEO. Removal is performed by a dozen 100-kg propellantless vehicles that react against the Earth's magnetic field. The debris objects are dragged down and released into short-lived orbits below ISS. As an alternative to deorbit, some of them can be collected for storage and possible in-orbit recycling. The estimated cost per kilogram of debris removed is a small fraction of typical launch costs per kilogram. These rates are low enough to open commercial opportunities and create a governing framework for wholesale removal of large debris objects from LEO.

  15. [Acrylic resin removable partial dentures

    NARCIS (Netherlands)

    Baat, C. de; Witter, D.J.; Creugers, N.H.J.

    2011-01-01

    An acrylic resin removable partial denture is distinguished from other types of removable partial dentures by an all-acrylic resin base which is, in principle, solely supported by the edentulous regions of the tooth arch and in the maxilla also by the hard palate. When compared to the other types of

  16. Complications of syndesmotic screw removal

    NARCIS (Netherlands)

    T. Schepers (Tim); E.M.M. van Lieshout (Esther); M.R. de Vries (Mark); M. van der Elst (Maarten)

    2011-01-01

    textabstractBackground: Currently, the metallic syndesmotic screw is the gold standard in the treatment of syndesmotic disruption. Whether or not this screw needs to be removed remains debatable. The aim of the current study was to determine the complications which occur following routine removal of

  17. Complications of syndesmotic screw removal

    NARCIS (Netherlands)

    Schepers, Tim; van Lieshout, Esther M. M.; de Vries, Mark R.; van der Elst, Maarten

    2011-01-01

    Currently, the metallic syndesmotic screw is the gold standard in the treatment of syndesmotic disruption. Whether or not this screw needs to be removed remains debatable. The aim of the current study was to determine the complications which occur following routine removal of the syndesmotic screw

  18. Krypton-85 removal and storage

    International Nuclear Information System (INIS)

    Gutierrez Fernandez, J.

    1978-01-01

    A literature survey was made in order to predict the atmospheric Kr-85 concentration in the future and it s effect on the population. As a consequence the need for its treatment and removal as a previous step to gaseous waste disposal is justified. A literature review of possible methods of Kr-85 removal and storage is also included. (Author) 43 refs

  19. Specifically requesting surgical tattoo removal: are deep personal motivations involved?

    Science.gov (United States)

    Koljonen, V; Kluger, N

    2012-06-01

      Motivations for tattoo removal include employment reasons, stigmata, changes in lifestyles or partners, incompatibility with present attitudes and values and clothing problems. Most studies on the motivations for tattoo removal have focused on patients seeking laser therapy. We hypothesized that patients seeking surgical tattoo removal would present with different motivations.   We analysed the characteristics and motivations of patients specifically requesting surgical tattoo removal.   We retrospectively reviewed the medical records of 16 patients in Helsinki, Finland, from 2005 to 2011. Demographic, clinical data, number of tattoos, location and size, time elapsed since tattooing, reason(s) for wanting surgical tattoo removal and surgical operations were analysed and compared with the other literature on tattoo removal.   Patients were mainly Caucasian females (ratio 3 : 1, median age of 26 years). Tattoos were all done by studio artists, most measured less than 30 cm², and were quite recent (median 5.3 years). Personal reasons accounted for 42.8% of all reasons, professional/social reasons for 37.5% and miscellaneous for 18.8%. Personal concerns were usually marital status changes, with few expressing dissatisfaction with the actual design of the tattoo. Tattoos were excised during a single procedure in 70% of the cases with only one case producing a hypertrophic scar.   Patients seeking surgical removal were aware of the limits and risks of the technique. They expressed intense personal reasons for wanting radical surgical removal. The possibility of surgical tattoo removal should be accessible to patients if the tattoo is small and discussion reveals strong personal motivation. © 2011 The Authors. Journal of the European Academy of Dermatology and Venereology © 2011 European Academy of Dermatology and Venereology.

  20. Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States); Mankamo, T.

    1995-04-01

    When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require {open_quotes}immediate{close_quotes} shutdown of the plant. In this paper, we present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion.

  1. Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

    International Nuclear Information System (INIS)

    Kim, I.S.; Samanta, P.K.

    1994-01-01

    When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require ''immediate'' shutdown of the plant. In this paper, the authors present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion

  2. Technetium removal: preliminary flowsheet options

    International Nuclear Information System (INIS)

    Eager, K.M.

    1995-01-01

    This document presents the results of a preliminary investigation into options for preliminary flowsheets for 99Tc removal from Hanford Site tank waste. A model is created to show the path of 99Tc through pretreatment to disposal. The Tank Waste Remediation (TWRS) flowsheet (Orme 1995) is used as a baseline. Ranges of important inputs to the model are developed, such as 99Tc inventory in the tanks and important splits through the TWRS flowsheet. Several technetium removal options are discussed along with sensitivities of the removal schemes to important model parameters

  3. Removing EU milk quotas, soft landing versus hard landing

    NARCIS (Netherlands)

    Bouamra-Mechemache, Z.; Jongeneel, R.; Réquillart, V.

    2008-01-01

    This paper analyses EU dairy policy reforms and mainly focus on EU milk quota removal scenarios. The model used to evaluate the scenario is a spatial equilibrium model of the dairy sector. It integrates the main competitor of the EU on world markets, Oceania, as well as the main importing regions in

  4. Removal of natural radionuclides from drinking water from private wells in Finland

    International Nuclear Information System (INIS)

    Huikuri, Pia; Salonen, Laina; Turtiainen, Tuukka

    1999-01-01

    Removal of natural radionuclides is often necessary in Finland when household water is taken from a drilled well. Removal of radionuclides by various methods from Finnish groundwaters were studied in a EU-research project, TENAWA. The results indicated that radon can be removed very efficiently (up to 99%) by applying aeration or granular activated carbon (GAC) filtration. Uranium and radium were also removed (over 94%) by using strong base anion (SBA) and strong acid cation (SAC) resins. The capability of reverse osmosis (RO) equipment to remove radionuclides was over 90% for uranium, radium and polonium. The water quality analyses indicated that water quality remained mostly good during the water treatment. (au)

  5. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    Energy Technology Data Exchange (ETDEWEB)

    Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

    2017-03-15

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  6. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    International Nuclear Information System (INIS)

    Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

    2017-01-01

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  7. Tank 5 Model for Sludge Removal Analysis

    International Nuclear Information System (INIS)

    LEE, SI

    2004-01-01

    Computational fluid dynamics methods have been used to develop and provide slurry pump operational guidance for sludge heel removal in Tank 5. Flow patterns calculated by the model were used to evaluate the performance of various combinations of operating pumps and their orientation under steady-state indexed and transient oscillation modes. A model used for previous analyses has been updated to add the valve housing distribution piping and pipe clusters of the cooling coil supply system near pump no. 8 to the previous tank Type-I model. In addition, the updated model included twelve concrete support columns. This model would provide a more accurate assessment of sludge removal capabilities. The model focused on removal of the sludge heel located near the wall of Tank 5 using the two new slurry pumps. The models and calculations were based on prototypic tank geometry and expected normal operating conditions as defined by Tank Closure Project Engineering. Computational fluid dynamics models of Tank 5 with different operating conditions were developed using the FLUENT (trademark) code. The modeling results were used to assess the efficiency of sludge suspension and removal operations in the 75-ft tank. The models employed a three-dimensional approach, a two-equation turbulence model, and an approximate representation of flow obstructions. The calculated local velocity was used as a measure of sludge removal and mixing capability. For the simulations, modeling calculations were performed with indexed pump orientations until an optimum flow pattern near the potential location of the sludge heel was established for sludge removal. The calculated results demonstrated that the existing slurry pumps running at 3801 gpm flowrate per nozzle could remove the sludge from the tank with a 101 in liquid level, based on a historical minimum sludge suspension velocity of 2.27 ft/sec. The only exception is the region within maximum 4.5 ft distance from the tank wall boundary at the

  8. Plasma polymer-functionalized silica particles for heavy metals removal.

    Science.gov (United States)

    Akhavan, Behnam; Jarvis, Karyn; Majewski, Peter

    2015-02-25

    Highly negatively charged particles were fabricated via an innovative plasma-assisted approach for the removal of heavy metal ions. Thiophene plasma polymerization was used to deposit sulfur-rich films onto silica particles followed by the introduction of oxidized sulfur functionalities, such as sulfonate and sulfonic acid, via water-plasma treatments. Surface chemistry analyses were conducted by X-ray photoelectron spectroscopy and time-of-flight secondary ion mass spectroscopy. Electrokinetic measurements quantified the zeta potentials and isoelectric points (IEPs) of modified particles and indicated significant decreases of zeta potentials and IEPs upon plasma modification of particles. Plasma polymerized thiophene-coated particles treated with water plasma for 10 min exhibited an IEP of less than 3.5. The effectiveness of developed surfaces in the adsorption of heavy metal ions was demonstrated through copper (Cu) and zinc (Zn) removal experiments. The removal of metal ions was examined through changing initial pH of solution, removal time, and mass of particles. Increasing the water plasma treatment time to 20 min significantly increased the metal removal efficiency (MRE) of modified particles, whereas further increasing the plasma treatment time reduced the MRE due to the influence of an ablation mechanism. The developed particulate surfaces were capable of removing more than 96.7% of both Cu and Zn ions in 1 h. The combination of plasma polymerization and oxidative plasma treatment is an effective method for the fabrication of new adsorbents for the removal of heavy metals.

  9. Tattoo removal with ingenol mebutate.

    Science.gov (United States)

    Cozzi, Sarah-Jane; Le, Thuy T; Ogbourne, Steven M; James, Cini; Suhrbier, Andreas

    2017-01-01

    An increasing number of people are getting tattoos; however, many regret the decision and seek their removal. Lasers are currently the most commonly used method for tattoo removal; however, treatment can be lengthy, costly, and sometimes ineffective, especially for certain colors. Ingenol mebutate is a licensed topical treatment for actinic keratoses. Here, we demonstrate that two applications of 0.1% ingenol mebutate can efficiently and consistently remove 2-week-old tattoos from SKH/hr hairless mice. Treatment was associated with relocation of tattoo microspheres from the dermis into the posttreatment eschar. The skin lesion resolved about 20 days after treatment initiation, with some cicatrix formation evident. The implications for using ingenol mebutate for tattoo removal in humans are discussed.

  10. Membrane adsorber for endotoxin removal

    Directory of Open Access Journals (Sweden)

    Karina Moita de Almeida

    Full Text Available ABSTRACT The surface of flat-sheet nylon membranes was modified using bisoxirane as the spacer and polyvinyl alcohol as the coating polymer. The amino acid histidine was explored as a ligand for endotoxins, aiming at its application for endotoxin removal from aqueous solutions. Characterization of the membrane adsorber, analysis of the depyrogenation procedures and the evaluation of endotoxin removal efficiency in static mode are discussed. Ligand density of the membranes was around 7 mg/g dry membrane, allowing removal of up to 65% of the endotoxins. The performance of the membrane adsorber prepared using nylon coated with polyvinyl alcohol and containing histidine as the ligand proved superior to other membrane adsorbers reported in the literature. The lack of endotoxin adsorption on nylon membranes without histidine confirmed that endotoxin removal was due to the presence of the ligand at the membrane surface. Modified membranes were highly stable, exhibiting a lifespan of approximately thirty months.

  11. Tattoo removal with ingenol mebutate

    Science.gov (United States)

    Cozzi, Sarah-Jane; Le, Thuy T; Ogbourne, Steven M; James, Cini; Suhrbier, Andreas

    2017-01-01

    An increasing number of people are getting tattoos; however, many regret the decision and seek their removal. Lasers are currently the most commonly used method for tattoo removal; however, treatment can be lengthy, costly, and sometimes ineffective, especially for certain colors. Ingenol mebutate is a licensed topical treatment for actinic keratoses. Here, we demonstrate that two applications of 0.1% ingenol mebutate can efficiently and consistently remove 2-week-old tattoos from SKH/hr hairless mice. Treatment was associated with relocation of tattoo microspheres from the dermis into the posttreatment eschar. The skin lesion resolved about 20 days after treatment initiation, with some cicatrix formation evident. The implications for using ingenol mebutate for tattoo removal in humans are discussed. PMID:28579816

  12. Removal of root filling materials.

    LENUS (Irish Health Repository)

    Duncan, H.F. Chong, B.S.

    2011-05-01

    Safe, successful and effective removal of root filling materials is an integral component of non-surgical root canal re-treatment. Access to the root canal system must be achieved in order to negotiate to the canal terminus so that deficiencies in the original treatment can be rectified. Since a range of materials have been advocated for filling root canals, different techniques are required for their removal. The management of commonly encountered root filling materials during non-surgical re-treatment, including the clinical procedures necessary for removal and the associated risks, are reviewed. As gutta-percha is the most widely used and accepted root filling material, there is a greater emphasis on its removal in this review.

  13. Tritium removal and retention device

    International Nuclear Information System (INIS)

    Boyle, R.F.; Durigon, D.D.

    1980-01-01

    A device is provided for removing and retaining tritium from a gaseous medium, and also a method of manufacturing the device. The device, consists of an inner core of zirconium alloy, preferably Zircaloy-4, and an outer adherent layer of nickel which acts as a selective and protective window for passage of tritium. The tritium then reacts with or is absorbed by the zirconium alloy, and is retained until such time as it is desirable to remove it during reprocessing. (auth)

  14. Removal of inclusions from silicon

    Science.gov (United States)

    Ciftja, Arjan; Engh, Thorvald Abel; Tangstad, Merete; Kvithyld, Anne; Øvrelid, Eivind Johannes

    2009-11-01

    The removal of inclusions from molten silicon is necessary to satisfy the purity requirements for solar grade silicon. This paper summarizes two methods that are investigated: (i) settling of the inclusions followed by subsequent directional solidification and (infiltration by ceramic foam filters. Settling of inclusions followed by directional solidification is of industrial importance for production of low-cost solar grade silicon. Filtration is reported as the most efficient method for removal of inclusions from the top-cut silicon scrap.

  15. Method removing radioactivity from kaolin

    International Nuclear Information System (INIS)

    Conley, R.F.

    1978-01-01

    A method of reducing the radioactivity found in naturally occurring kaolins to about 40% below its native value, and the leachable radiogenic components to less than 20% is described. This reduction is achieved by removing from the kaolin particles of a size less than 0.5 microns. This removal may be carried out by gravitational settling, flocculation of non-colloidal particles, or acid leaching

  16. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  17. [Acrylic resin removable partial dentures].

    Science.gov (United States)

    de Baat, C; Witter, D J; Creugers, N H J

    2011-01-01

    An acrylic resin removable partial denture is distinguished from other types of removable partial dentures by an all-acrylic resin base which is, in principle, solely supported by the edentulous regions of the tooth arch and in the maxilla also by the hard palate. When compared to the other types of removable partial dentures, the acrylic resin removable partial denture has 3 favourable aspects: the economic aspect, its aesthetic quality and the ease with which it can be extended and adjusted. Disadvantages are an increased risk of caries developing, gingivitis, periodontal disease, denture stomatitis, alveolar bone reduction, tooth migration, triggering of the gag reflex and damage to the acrylic resin base. Present-day indications are ofa temporary or palliative nature or are motivated by economic factors. Special varieties of the acrylic resin removable partial denture are the spoon denture, the flexible denture fabricated of non-rigid acrylic resin, and the two-piece sectional denture. Furthermore, acrylic resin removable partial dentures can be supplied with clasps or reinforced by fibers or metal wires.

  18. Guidelines for removing permanent makeup

    Directory of Open Access Journals (Sweden)

    C.Bettina Rümmelein

    2016-09-01

    Full Text Available Permanent makeup (PMU is a frequently implemented cosmetic procedure performed by beauticians. From a technical point, PMU is considered a facial tattoo. Failed procedures or a change of mind can lead to the desire for removal. The purpose of this retrospective evaluation of patients who came to the clinic with the desire to remove PMU between 2011 and 2015 was to explore the problems, side effects, and results in order to define treatment guidelines for other doctors. We evaluated 87 individual cases in total. In treatable cases, i.e. 52 out of the 87 cases, laser treatments were performed using a nanosecond Q-switched neodymium-doped yttrium aluminium garnet (Nd:YAG laser. It takes between 1-12 treatments to remove the PMU. In three cases, the colour of the PMU could not be removed by laser and remained after the treatment. In two cases, laser treatment had to be terminated due to colour changes towards the green-blue spectrum. Before PMU removal, laser test shots are urgently recommended as unforeseeable colour changes can cause severe aesthetically unpleasant results. Covered up PMU (skin colour is particularly susceptible to changes in colour. Heat-induced shrinking of the eye area can cause an ectropium. Surgical solutions also have to be taken into consideration. The use of proper eye protection with intraocular eye shields is mandatory. This article is an attempt to set up some guidelines for the treatment of PMU removal.

  19. After-heat removing device

    International Nuclear Information System (INIS)

    Iwashige, Kengo; Otsuka, Masaya; Yokoyama, Iwao; Yamakawa, Masanori.

    1990-01-01

    The present invention concerns an after-heat removing device for first reactors. A heat accumulation portion provided in a cooling channel of an after-heat removing device is disposed before a coil-like heat conduction pipe for cooling of the after-heat removing device. During normal reactor operation, the temperature in the heat accumulation portion is near the temperature of the high temperature plenum due to heat conduction and heat transfer from the high temperature plenum. When the reactor is shutdown and the after-heat removing device is started, coolants cooled in the air cooler start circulation. The coolants arriving at the heat accumulation portion deprive heat from the heat accumulation portion and, ion turn, increase their temperature and then reach the cooling coil. Subsequently, the heat calorie possessed in the heat accumulation portion is reduced and the after-heat removing device is started for the operation at a full power. This can reduce the thermal shocks applied to the cooling coil or structures in a reactor vessel upon starting the after-heat removing device. (I.N.)

  20. Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan); Muranaka, Kohmei; Asai, Takayuki [Graduate School of Engineering, University of Fukui (Japan); Rooijen, W.F.G. van, E-mail: rooijen@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan)

    2014-08-15

    Highlights: • The IAEA EBR-II benchmarks SHRT-17 and SHRT-45R are analyzed with a 1D system code. • The calculated result of SHRT-17 corresponds well to the measured results. • For SHRT-45R ERANOS is used for various core parameters and reactivity coefficients. • SHRT-45R peak temperature is overestimated with the ERANOS feedback coefficients. • The peak temperature is well predicted when the feedback coefficient is reduced. - Abstract: Benchmark problems of several experiments in EBR-II, proposed by ANL and coordinated by the IAEA, are analyzed using the plant system code NETFLOW++ and the neutronics code ERANOS. The SHRT-17 test conducted as a loss-of-flow test is calculated using only the NETFLOW++ code because it is a purely thermal–hydraulic problem. The measured data were made available to the benchmark participants after the results of the blind benchmark calculations were submitted. Our work shows that major parameters of the plant are predicted with good accuracy. The SHRT-45R test, an unprotected loss of flow test is calculated using the NETFLOW++ code with the aid of delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to perform a semi-coupled analysis of the neutronics – thermal–hydraulic problem. The measured data are compared with our calculated results. In our work, the peak temperature is underestimated, indicating that the reactivity feedback coefficients are too strong. When the reactivity feedback coefficient for thermal expansion is adjusted, good agreement is obtained in general for the calculated plant parameters, with a few exceptions.

  1. Parametric Analyses of Heat Removal from High-Level Waste Tanks

    International Nuclear Information System (INIS)

    TRUITT, J.B.

    2000-01-01

    The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined

  2. Removal of gasoline vapors from air streams by biofiltration

    Energy Technology Data Exchange (ETDEWEB)

    Apel, W.A.; Kant, W.D.; Colwell, F.S.; Singleton, B.; Lee, B.D.; Andrews, G.F.; Espinosa, A.M.; Johnson, E.G.

    1993-03-01

    Research was performed to develop a biofilter for the biodegradation of gasoline vapors. The overall goal of this effort was to provide information necessary for the design, construction, and operation of a commercial gasoline vapor biofilter. Experimental results indicated that relatively high amounts of gasoline vapor adsorption occur during initial exposure of the biofilter bed medium to gasoline vapors. Biological removal occurs over a 22 to 40{degrees}C temperature range with removal being completely inhibited at 54{degrees}C. The addition of fertilizer to the relatively fresh bed medium used did not increase the rates of gasoline removal in short term experiments. Microbiological analyses indicated that high levels of gasoline degrading microbes are naturally present in the bed medium and that additional inoculation with hydrocarbon degrading cultures does not appreciably increase gasoline removal rates. At lower gasoline concentrations, the vapor removal rates were considerably lower than those at higher gasoline concentrations. This implies that system designs facilitating gasoline transport to the micro-organisms could substantially increase gasoline removal rates at lower gasoline vapor concentrations. Test results from a field scale prototype biofiltration system showed volumetric productivity (i.e., average rate of gasoline degradation per unit bed volume) values that were consistent with those obtained with laboratory column biofilters at similar inlet gasoline concentrations. In addition, total benzene, toluene, ethyl-benzene, and xylene (BTEX) removal over the operating conditions employed was 50 to 55%. Removal of benzene was approximately 10 to 15% and removal of the other members of the BTEX group was much higher, typically >80%.

  3. Removal of gasoline vapors from air streams by biofiltration

    Energy Technology Data Exchange (ETDEWEB)

    Apel, W.A.; Kant, W.D.; Colwell, F.S.; Singleton, B.; Lee, B.D.; Andrews, G.F.; Espinosa, A.M.; Johnson, E.G.

    1993-03-01

    Research was performed to develop a biofilter for the biodegradation of gasoline vapors. The overall goal of this effort was to provide information necessary for the design, construction, and operation of a commercial gasoline vapor biofilter. Experimental results indicated that relatively high amounts of gasoline vapor adsorption occur during initial exposure of the biofilter bed medium to gasoline vapors. Biological removal occurs over a 22 to 40[degrees]C temperature range with removal being completely inhibited at 54[degrees]C. The addition of fertilizer to the relatively fresh bed medium used did not increase the rates of gasoline removal in short term experiments. Microbiological analyses indicated that high levels of gasoline degrading microbes are naturally present in the bed medium and that additional inoculation with hydrocarbon degrading cultures does not appreciably increase gasoline removal rates. At lower gasoline concentrations, the vapor removal rates were considerably lower than those at higher gasoline concentrations. This implies that system designs facilitating gasoline transport to the micro-organisms could substantially increase gasoline removal rates at lower gasoline vapor concentrations. Test results from a field scale prototype biofiltration system showed volumetric productivity (i.e., average rate of gasoline degradation per unit bed volume) values that were consistent with those obtained with laboratory column biofilters at similar inlet gasoline concentrations. In addition, total benzene, toluene, ethyl-benzene, and xylene (BTEX) removal over the operating conditions employed was 50 to 55%. Removal of benzene was approximately 10 to 15% and removal of the other members of the BTEX group was much higher, typically >80%.

  4. How Effective are Existing Arsenic Removal Techniques

    Science.gov (United States)

    This presentation will summarize the system performance results of the technologies demonstrated in the arsenic demonstration program. The technologies include adsorptive media, iron removal, iron removal with iron additions, iron removal followed by adsorptive media, coagulatio...

  5. Technetium removal from aqueous wastes

    International Nuclear Information System (INIS)

    Fletcher, P.A.; Jones, C.P.; Junkison, A.R.; Turner, A.D.; Kavanagh, P.R.

    1992-03-01

    The research discussed in this report has compared several ''state of the art'' techniques for the removal of traces of the radionuclide, technetium, from aqueous wastes. The techniques investigated were: electrochemical reduction to an insoluble oxide, electrochemical ion exchange, seeded ultrafiltration and chemical reduction followed by filtration. Each technique was examined using a simulant based upon the waste generated by the Enhanced Actinide Removal Plant (EARP) at Sellafield. The technique selected for further investigation was direct electrochemical reduction which offers an ideal route for the removal of technetium from the stream (DFs 10-100) and can be operated continuously with a low power consumption 25 kW for the waste generated by EARP. Cell designs for scale up have been suggested to treat the 1000m 3 of waste produced every day. Future work is proposed to investigate the simultaneous removal of other key radionuclides, such as ruthenium, plutonium and cobalt as well as scale up of the resulting process and to investigate the effect of these other radionuclides on the efficiency of the electrochemical reduction technique for the removal of technetium. Total development and full scale plant costs are estimated to be of the order of 5 pounds - 10M, with a time scale of 5 -8 years to realisation. (author)

  6. ANALYSIS OF AMMONIA REMOVAL FROM WASTEWATER MARKET: FEASIBILITY OF SALTWORKS INTRODUCING NEW TECHNOLOGY

    OpenAIRE

    Nicholas Roch

    2015-01-01

    This paper presents an analysis of the market for removing ammonia from wastewater to assess its attractiveness and confirm the feasibility of Saltworks developing and launching its promising new ammonia removal technology. After an introduction, the paper qualitatively analyses the opportunity for Saltworks to enter the ammonia removal market using a SWOT analysis. The author’s personal experiences, Saltworks documentation, and interviews with Saltworks staff provide insight into the company...

  7. Removal of bromates from water

    Science.gov (United States)

    Barlokova, D.; Ilavsky, J.; Marko, I.; Tkacova, J.

    2017-10-01

    Bromates are substances that are usually not present in drinking water. They are obtained by ozone disinfection in the presence of bromine ions in water, as an impurity of sodium hypochlorite, respectively. Because of their specific properties, bromates are classified as vary dangers substances, that can cause serious illnesses in humans. There are several technological processes that have been used to the removal of bromates from water at present. In this article, the removal of the bromates from water by the adsorption using various sorbent materials (activated carbon, zeolite, Klinopur-Mn, Bayoxide E33, GEH, Read-As and Activated alumina) are presented. The effectiveness of selected sorbent materials in the removal of bromates from drinking water moves in the interval from 10 to 40%. Based on laboratory results, the zeolite can be used to reduce the concentration of bromates in water.

  8. Arsenic removal by electrocoagulation process: Recent trends and removal mechanism.

    Science.gov (United States)

    Nidheesh, P V; Singh, T S Anantha

    2017-08-01

    Arsenic contamination in drinking water is a major issue in the present world. Arsenicosis is the disease caused by the regular consumption of arsenic contaminated water, even at a lesser contaminated level. The number of arsenicosis patients is increasing day-by-day. Decontamination of arsenic from the water medium is the only one way to regulate this and the arsenic removal can be fulfilled by water treatment methods based on separation techniques. Electrocoagulation (EC) process is a promising technology for the effective removal of arsenic from aqueous solution. The present review article analyzes the performance of the EC process for arsenic removal. Electrocoagulation using various sacrificial metal anodes such as aluminium, iron, magnesium, etc. is found to be very effective for arsenic decontamination. The performances of each anode are described in detail. A special focus has been made on the mechanism behind the arsenite and arsenate removal by EC process. Main trends in the disposal methods of sludge containing arsenic are also included. Comparison of arsenic decontamination efficiencies of chemical coagulation and EC is also reported. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Removal (and attempted removal) of material from a Hooded Vulture ...

    African Journals Online (AJOL)

    Relatively little is documented about nest material theft in vultures. We used camera traps to monitor Hooded Vulture Necrosyrtes monachus nests for a year. We report camera trap photographs of a starling Lamprotornis sp. removing what appeared to be dung from an inactive Hooded Vulture nest on Cleveland Game ...

  10. Treatment of radioactive laboratory waste for mercury removal

    International Nuclear Information System (INIS)

    Osteen, A.B.; Bibler, J.P.

    1990-01-01

    Routine analyses of Savannah River Laboratory wastes at the Savannah River Site occasionally reveal mercury concentrations in the waste in excess of the 0.200 μg/L RCRA limit. An ion exchange resin has been demonstrated to be effective for the removal of dissolved mercury from laboratory waste in a special permitted project. The ion exchange material is Duolite trademark GT-73, a polystyrene/divinylbenzene resin with thiol functional groups. As a result of the decontamination demonstration, the resin is in use or under consideration for use with several other SRS radwaste streams as a reliable medium for mercury removal

  11. FFTF operating experience with sodium natural circulation: slides included

    Energy Technology Data Exchange (ETDEWEB)

    Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

    1981-01-01

    The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures.

  12. FFTF operating experience with sodium natural circulation: slides included

    International Nuclear Information System (INIS)

    Burke, T.M.; Additon, S.L.; Beaver, T.R.; Midgett, J.C.

    1981-01-01

    The Fast Flux Test Facility (FFTF) has been designed for passive, back-up, safety grade decay heat removal utilizing natural circulation of the sodium coolant. This paper discusses the process by which operator preparation for this emergency operating mode has been assured, in paralled with the design verification during the FFTF startup and acceptance testing program. Over the course of the test program, additional insights were gained through the testing program, through on-going plant analyses and through general safety evaluations performed throughout the nuclear industry. These insights led to development of improved operator training material for control of decay heat removal during both forced and natural circulation as well as improvements in the related plant operating procedures

  13. Laser-based coatings removal

    International Nuclear Information System (INIS)

    Freiwald, J.G.; Freiwald, D.A.

    1995-01-01

    Over the years as building and equipment surfaces became contaminated with low levels of uranium or plutonium dust, coats of paint were applied to stabilize the contaminants in place. Most of the earlier paint used was lead-based paint. More recently, various non-lead-based paints, such as two-part epoxy, are used. For D ampersand D (decontamination and decommissioning), it is desirable to remove the paints or other coatings rather than having to tear down and dispose of the entire building. This report describes the use of pulse-repetetion laser systems for the removal of paints and coatings

  14. Laser-based coatings removal

    Energy Technology Data Exchange (ETDEWEB)

    Freiwald, J.G.; Freiwald, D.A. [F2 Associates, Inc., Albuquerque, NM (United States)

    1995-10-01

    Over the years as building and equipment surfaces became contaminated with low levels of uranium or plutonium dust, coats of paint were applied to stabilize the contaminants in place. Most of the earlier paint used was lead-based paint. More recently, various non-lead-based paints, such as two-part epoxy, are used. For D&D (decontamination and decommissioning), it is desirable to remove the paints or other coatings rather than having to tear down and dispose of the entire building. This report describes the use of pulse-repetetion laser systems for the removal of paints and coatings.

  15. Removal of radionuclides at a waterworks

    International Nuclear Information System (INIS)

    Gaefvert, T.; Ellmark, C.; Holm, E.

    2002-01-01

    A waterworks, providing several large cities in the province of Scania with drinking-water, with an average production rate of 1.3 m 3. s -1 has been studied regarding its removal capacity for several natural and anthropogenic radionuclides. The raw water is surface water from lake Bolmen which is transported through an 80 km long tunnel in the bedrock before it enters the waterworks. The method used for purification is a combination of precipitation and filtration in sand filters. Two different purification lines are at the moment in use, one using A1 2 (SO 4 ) 3 as a coagulant and one using FeC1 3 . After coagulation and flocculation the precipitation is removed and the water is passed through two different sand filters (rapid-filtration and slow-filtration). Water samples have been collected at the lake, the inlet at the waterworks, after each of the flocculation basins (A1 2 (SO 4 ) 3 and FeC1 3 ), after rapid-filtration and from the municipal distribution net. The samples have been analysed with respect to its content of uranium, thorium, polonium, radium, plutonium and caesium. (au)

  16. Removal of radium from drinking water

    International Nuclear Information System (INIS)

    Lauch, R.P.

    1992-08-01

    The report summarizes processes for removal of radium from drinking water. Ion exchange, including strong acid and weak acid resin, is discussed. Both processes remove better than 95 percent of the radium from the water. Weak acid ion exchange does not add sodium to the water. Calcium cation exchange removes radium and can be used when hardness removal is not necessary. Iron removal processes are discussed in relation to radium removal. Iron oxides remove much less than 20 percent of the radium from water under typical conditions. Manganese dioxide removes radium from water when competition for sorption sites and clogging of sites is reduced. Filter sand that is rinsed daily with dilute acid will remove radium from water. Manganese dioxide coated filter sorption removes radium but more capacity would be desirable. The radium selective complexer selectively removes radium with significant capacity if iron fouling is eliminated

  17. Sample preparation in foodomic analyses.

    Science.gov (United States)

    Martinović, Tamara; Šrajer Gajdošik, Martina; Josić, Djuro

    2018-04-16

    Representative sampling and adequate sample preparation are key factors for successful performance of further steps in foodomic analyses, as well as for correct data interpretation. Incorrect sampling and improper sample preparation can be sources of severe bias in foodomic analyses. It is well known that both wrong sampling and sample treatment cannot be corrected anymore. These, in the past frequently neglected facts, are now taken into consideration, and the progress in sampling and sample preparation in foodomics is reviewed here. We report the use of highly sophisticated instruments for both high-performance and high-throughput analyses, as well as miniaturization and the use of laboratory robotics in metabolomics, proteomics, peptidomics and genomics. This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

  18. Thermal analyses of spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Ikonen, K.

    2003-06-01

    This report contains the temperature dimensioning of the KBS-3V type 1- or 2-panel repository based on the rock properties measured from the Olkiluoto investigations. The report describes first the development of a calculation methodology for the thermal analysis of a repository for nuclear fuel. The disposed canisters produce residual heat due to decay (or disintegration) of radioactive products. The decay heat is conducted to surrounding rock mass. The methods were applied to determine the effect of different parameters on the highest canister temperature and to support the planning, dimensioning and operation of the repository. The thermal diffusivity of the rock is low and the heat released from the canisters is spread into the surrounding rock volume quite slowly causing thermal gradient in the rock close to canisters and the canister temperature is increased remarkably. The maximum temperature on the canister surface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by the spacing between adjacent canisters, adjacent tunnels and the distance between separate panels of the repository and the pre-cooling time affecting power of the canisters. Because of the fact that the disposal operation takes several decades, the moment of disposal of an individual canister in addition to the location has an influence on the maximum temperature in the canister. Also, a second disposal panel in the repository has a thermal interaction with the other panel. This interaction is expressed after a few decades at the strongest. It became apparent that the temperature of canister surfaces can be determined by analytic line heat source model much more efficiently than by numerical analysis, if the analytic model is first verified and

  19. Comparison of decay heat exchangers placing in the primary circuit of pool type fast reactor

    International Nuclear Information System (INIS)

    Birbraer, P.N.; Gorbunov, V.S.; Zotov, V.G.; Kuzavkov, N.G.; Pykhonin, V.A.; Sobolev, V.A.; Ryzhov, V.A.

    1993-01-01

    Description of two alternative arrangements of decay beat exchangers (DHXs) in the fast reactor tank is presented: in 'hot' cavity and in 'cold' cavity. The results of calculation for the two alternative arrangements as regards static and dynamic parameters in the primary circuit on 1-D program are given. (author)

  20. Treatment of the decay heat production in the reactor dynamics program TINTE

    International Nuclear Information System (INIS)

    Gerwin, H.; Scherer, W.

    1993-07-01

    The TINTE code system deals with the nuclear and the thermal transient behaviour of the primary circuit of an HTGR taking into consideration the mutual feedback effects in two-dimensional r-z-geometry. An update of the treatment of delayed heat production is presented. It is based on the German norm DIN 25485, the rules of which had to be adjusted for use in a dynamics code. For the description of the fuel element power history a substitute-histogram has been constructed from local burnup and optionally from information about shuffling of the fuel balls. As an example the depressurisation accident of a MODUL-HTR is calculated. The results obtained are very similiar to others previously reported. (orig./HP) [de

  1. Decay heat measurement of 235U in the time period from 10 to 1000 seconds

    International Nuclear Information System (INIS)

    Baumung, K.

    1977-01-01

    Nearly all components of the experimental facility were completed. The automatic control of the pneumatic transfer system and the active fuel sample handling mechanism was constructed and succesfully tested. The computer programs for the calorimeter control and data recording by a 8K-minicomputer were provided and the measuring electronics completed. The shielding from γ- and delayed neutron radiation was designed and constructed and the security report for the operation of the facility at the reactor written. (orig./RW) [de

  2. The ratio between the decay heat output and activity content of discharged magnox fuel

    International Nuclear Information System (INIS)

    Davies, B.S.J.

    1977-01-01

    Values of the ratio between activity and heat production rate have been calculated for magnox fuel irradiated to 3500 and 8000 MWd.Te -1 and for cooling times of 100, 200 and 500 days. Results are expressed in terms of both MeV.decay -1 and MCi.KW -1 . The results indicate that: for these irradiation and cooling conditions 21 nuclides account for over 99% of the total activity; the calculated values show only small variations with burn-up and cooling time, although the mean energy per decay does fall slightly at 500 days cooling: so for many purposes a median value of 0.63 MeV.decay -1 (0.27 MCi.MW -1 ) may be used; the calculated values have standard deviations ranging from 2.6% at 100 days cooling to 9% at 500 days cooling. (author)

  3. Radioactivity and decay heat generation in precambrian magmatic rocks (with the South Pamirs as an example)

    International Nuclear Information System (INIS)

    Batyrmurzaev, A.S.; Alibekov, G.I.; Bekieva, A.A.

    2003-01-01

    The evaluation of the heat generation share in the results of the long-living radioactive elements (RAE) decay in the Earth surface layers is accomplished on the basis of the data on the uranium and thorium concentration in the precambrian magmatic rocks of the South Pamirs. It was supposed by the calculations, that the value of the heat flux, generated by the rocks, is determined mainly by the RAE content in the Earth upper layer crust itself of 10-15 km. It is shown that the radioheat generation share is within the range of 5-10% from the measured values of the geothermal flows [ru

  4. Beta decay far from stability and the decay heat of nuclear reactors

    International Nuclear Information System (INIS)

    Klapdor, H.V.; Metzinger, J.; Grotz, K.

    1986-01-01

    First results of a second generation microscopic calculation which we denote by KGM, in contrast to the older KMO calculations (which were actually done in 1981), are shown. It is seen that the new calculations which use an RPA approach treating the odd particle by perturbation theory, can give an improved description of the half-lives, particularly also in the chromium to cobalt region, where some systematic deviation seemed to be present. (orig./BBOE)

  5. Decay heat measurement of 235U in the time period from 10 to 1000 seconds

    International Nuclear Information System (INIS)

    Baumung, K.

    1978-01-01

    During the report period, the experimental facility was installed at the FR2 reactor. A delay was caused by the fact, that the supplying with water for the cooling loop of the pneumatic transfer system as well as discharge of exhaust gas into the halogeneventilation system of the reactor could not be performed as planned before. Therefore a cooling system was provided and tested and the efficiency of special halogene filters belonging to the experimental facility was verified. Now the apparatus is assembled and first cold tests have been performed. (orig.) [de

  6. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

  7. Arsenic removal by lime softening

    DEFF Research Database (Denmark)

    Kaosol, T.; Suksaroj, C.; Bregnhøj, Henrik

    2002-01-01

    This paper focuses on the study of arsenic removal for drinking water by lime softening. The initial arsenic (V) concentration was 500 and 1,000 ug/L in synthetic groundwater. The experiments were performed as batch tests with varying lime dosages and mixing time. For the synthetic groundwater......, arsenic (V) removal increased with increasing lime dosage and mixing time, as well as with the resulting pH. The residual arsenic (V) in all cases was lower than the WHO guideline of 10 ug/L at pH higher than 11.5. Kinetic of arsenic (V) removal can be described by a first-order equation as C1 = C0*e......^-k*t. The relation between the constant (k value) and increasing lime dosage was found to be linear, described by k = 0.0034 (Dlime). The results support a theory from the literature that the arsenic (V) was removed by precipitation af Ca3(AsO4)2. The results obtained in the present study suggest that lime...

  8. Can Acne Scars Be Removed?

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Can Acne Scars Be Removed? KidsHealth / For Teens / Can Acne ... eliminar las cicatrices del acné? Different Types of Acne Scars from acne can seem like double punishment — ...

  9. GLYPHOSATE REMOVAL FROM DRINKING WATER

    Science.gov (United States)

    Activated-carbon, oxidation, conventional-treatment, filtration, and membrane studies are conducted to determine which process is best suited to remove the herbicide glyphosate from potable water. Both bench-scale and pilot-scale studies are completed. Computer models are used ...

  10. Arsenic Removal by Liquid Membranes

    Directory of Open Access Journals (Sweden)

    Tiziana Marino

    2015-03-01

    Full Text Available Water contamination with harmful arsenic compounds represents one of the most serious calamities of the last two centuries. Natural occurrence of the toxic metal has been revealed recently for 21 countries worldwide; the risk of arsenic intoxication is particularly high in Bangladesh and India but recently also Europe is facing similar problem. Liquid membranes (LMs look like a promising alternative to the existing removal processes, showing numerous advantages in terms of energy consumption, efficiency, selectivity, and operational costs. The development of different LM configurations has been a matter of investigation by several researching groups, especially for the removal of As(III and As(V from aqueous solutions. Most of these LM systems are based on the use of phosphine oxides as carriers, when the metal removal is from sulfuric acid media. Particularly promising for water treatment is the hollow fiber supported liquid membrane (HFSLM configuration, which offers high selectivity, easy transport of the targeted metal ions, large surface area, and non-stop flow process. The choice of organic extractant(s plays an essential role in the efficiency of the arsenic removal. Emulsion liquid membrane (ELM systems have not been extensively investigated so far, although encouraging results have started to appear in the literature. For such LM configuration, the most relevant step toward efficiency is the choice of the surfactant type and its concentration.

  11. Tattoo removal with ingenol mebutate

    Directory of Open Access Journals (Sweden)

    Cozzi SJ

    2017-05-01

    Full Text Available Sarah-Jane Cozzi,1 Thuy T Le,1 Steven M Ogbourne,2 Cini James,1 Andreas Suhrbier1 1Inflammation Biology Laboratory, QIMR Berghofer Medical Research Institute, Brisbane, 2Genecology Research Center, Faculty of Science, Health, Engineering and Education, University of the Sunshine Coast, Maroochydore DC, QLD, Australia Abstract: An increasing number of people are getting tattoos; however, many regret the decision and seek their removal. Lasers are currently the most commonly used method for tattoo removal; however, treatment can be lengthy, costly, and sometimes ineffective, especially for certain colors. Ingenol mebutate is a licensed topical treatment for actinic keratoses. Here, we demonstrate that two applications of 0.1% ingenol mebutate can efficiently and consistently remove 2-week-old tattoos from SKH/hr hairless mice. Treatment was associated with relocation of tattoo microspheres from the dermis into the posttreatment eschar. The skin lesion resolved about 20 days after treatment initiation, with some cicatrix formation evident. The implications for using ingenol mebutate for tattoo removal in humans are discussed. Keywords: tattoo, ingenol mebutate, mouse 

  12. Biotechnological sulphide removal with oxygen

    NARCIS (Netherlands)

    Buisman, C.

    1989-01-01

    This thesis deals with the development of a new process for biotechnological sulphide removal from wastewater, in which it is attempted to convert sulphide into elemental sulphur by colourless sulphur bacteria. The toxicity, corrosive properties, unpleasant odor and high oxygen demand of sulphide

  13. The study of hydrogen removal

    International Nuclear Information System (INIS)

    Yasufuku, Katsumi; Fukuhara, Masashi; Izaki, Takashi; Nakase, Takeshi

    1979-01-01

    Two methods of hydrogen removal from the helium coolant for high temperature helium gas-cooled nuclear reactor plants were investigated; the one is the process absorbing hydrogen with titanium sponges and the other is the water removal with zeolite, after hydrogen is converted to water utilizing copper oxide (CuO). The special feature of these two hydrogen removal methods is to treat the very low hydrogen concentration in helium about 0.06 mm Hg (2 Vpm, 41 ata). As for the titanium sponge method, a preliminary experimental facility was constructed to test the temperature dependences of the quantity of equilibrium absorption of hydrogen and the diffusion velocity inside titanium sponge by the batch type constant volume process. The temperature of titanium sponge was 800 deg C, the vacuum was from 2 to 3 x 10 -7 mm Hg and hydrogen partial pressure was from 1.0 to 10 -4 mm Hg in the experiment. The measured hydrogen absorption rate and the diffusion velocity data are presented, and the experimental conditions were evaluated. After the preliminary experiment, a mini-loop was constructed to confirm the temperature and velocity dependences of overall capacity factor, and the overall capacity factor and the regenerating characteristics of titanium sponge were tested. These experimental data are shown, and were evaluated. Concerning the hydrogen removal method utilizing CuO, the experiment was carried out under the following test conditions: the temperature from 400 to 265 deg C, the linear velocity from 50.3 to 16.7 cm/sec and the hydrogen concentration from 12.0 to 1.93 mm/Hg. The hydrogen removal rate and capacity were obtained in this experiment, and the data are presented and explained. (Nakai, Y.)

  14. 27 CFR 25.251 - Authorized removals.

    Science.gov (United States)

    2010-04-01

    ..., DEPARTMENT OF THE TREASURY LIQUORS BEER Removal of Brewer's Yeast and Other Articles § 25.251 Authorized removals. (a) Brewer's yeast. A brewer may remove brewer's yeast, in liquid or solid form containing not... including the words “Brewer's Yeast.” (c) Pipeline. If brewer's yeast is removed by pipeline, the pipeline...

  15. VDTT removal system functional design criteria

    International Nuclear Information System (INIS)

    Legare, D.E.

    1996-01-01

    Two Velocity Density Temperature Trees (H-2-815016) are to be removed from risers 14A and 1B of tank 241-SY-101. This document provides functional design criteria for the removal system. The removal system consists of a Liquid Removal Tool, Flexible Receiver (H-2-79216), Burial Container, Transport Trailers, and associated equipment

  16. Descriptive Analyses of Mechanical Systems

    DEFF Research Database (Denmark)

    Andreasen, Mogens Myrup; Hansen, Claus Thorp

    2003-01-01

    Forord Produktanalyse og teknologianalyse kan gennmføres med et bredt socio-teknisk sigte med henblik på at forstå kulturelle, sociologiske, designmæssige, forretningsmæssige og mange andre forhold. Et delområde heri er systemisk analyse og beskrivelse af produkter og systemer. Nærværende kompend...

  17. Analysing and Comparing Encodability Criteria

    Directory of Open Access Journals (Sweden)

    Kirstin Peters

    2015-08-01

    Full Text Available Encodings or the proof of their absence are the main way to compare process calculi. To analyse the quality of encodings and to rule out trivial or meaningless encodings, they are augmented with quality criteria. There exists a bunch of different criteria and different variants of criteria in order to reason in different settings. This leads to incomparable results. Moreover it is not always clear whether the criteria used to obtain a result in a particular setting do indeed fit to this setting. We show how to formally reason about and compare encodability criteria by mapping them on requirements on a relation between source and target terms that is induced by the encoding function. In particular we analyse the common criteria full abstraction, operational correspondence, divergence reflection, success sensitiveness, and respect of barbs; e.g. we analyse the exact nature of the simulation relation (coupled simulation versus bisimulation that is induced by different variants of operational correspondence. This way we reduce the problem of analysing or comparing encodability criteria to the better understood problem of comparing relations on processes.

  18. Analysing Children's Drawings: Applied Imagination

    Science.gov (United States)

    Bland, Derek

    2012-01-01

    This article centres on a research project in which freehand drawings provided a richly creative and colourful data source of children's imagined, ideal learning environments. Issues concerning the analysis of the visual data are discussed, in particular, how imaginative content was analysed and how the analytical process was dependent on an…

  19. Impact analyses after pipe rupture

    International Nuclear Information System (INIS)

    Chun, R.C.; Chuang, T.Y.

    1983-01-01

    Two of the French pipe whip experiments are reproduced with the computer code WIPS. The WIPS results are in good agreement with the experimental data and the French computer code TEDEL. This justifies the use of its pipe element in conjunction with its U-bar element in a simplified method of impact analyses

  20. Millifluidic droplet analyser for microbiology

    NARCIS (Netherlands)

    Baraban, L.; Bertholle, F.; Salverda, M.L.M.; Bremond, N.; Panizza, P.; Baudry, J.; Visser, de J.A.G.M.; Bibette, J.

    2011-01-01

    We present a novel millifluidic droplet analyser (MDA) for precisely monitoring the dynamics of microbial populations over multiple generations in numerous (=103) aqueous emulsion droplets (100 nL). As a first application, we measure the growth rate of a bacterial strain and determine the minimal

  1. Analyser of sweeping electron beam

    International Nuclear Information System (INIS)

    Strasser, A.

    1993-01-01

    The electron beam analyser has an array of conductors that can be positioned in the field of the sweeping beam, an electronic signal treatment system for the analysis of the signals generated in the conductors by the incident electrons and a display for the different characteristics of the electron beam

  2. Present municipal water treatment and potential removal methods

    International Nuclear Information System (INIS)

    Lee, S.Y.; White, S.K.; Bondietti, E.A.

    1982-01-01

    Uranium analyses of raw water, intermediate stage, and treated water samples from 20 municipal water treatment plants indicated that the present treatment practices were not effective in removing uranium from raw waters when the influent concentration was in the range of 0.1 to 16 μg/L uranium. Laboratory batch tests revealed that the water softening and coagulant chemicals commonly used were able to remove more than 90% of the dissolved uranium ( < 100 μg/L) in waters if an optimum pH and dosage were provided. Absorbents, titanium oxide and activated charcoal, were also effective in uranium removal under specific conditions. Strong base anion exchange resin was the most efficient uranium adsorbent, and an anion exchange column is a recommended option for the treatment of private well waters containing uranium at higher than desirable levels

  3. ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses

    International Nuclear Information System (INIS)

    2008-01-01

    problems: calculation of perturbation integrals, of cross-section variations, sensitivity analysis, perturbation analysis. As a matter of fact, sensitivity analyses, and first-order or exact perturbation analyses can be performed for the multiplication factor (standard perturbation theory, SPT), ratios of linear or bilinear integrals (generalized perturbation theory, GPT), and reactivity effects (equivalent generalized perturbation theory, EGPT). If a dispersion (variance/covariance) matrix is provided, a specific module can be used to perform uncertainty and representativeness calculations. Core follow-up: Specific ERANOS modules and appropriate complex subroutines written in the LU user's language (the PROJERIX package) are available to perform a detailed core follow-up. Each individual sub-assembly can be followed through its entire life (moves during shuffles and batch re-loadings, time spent in internal storage, etc.). Fine burn-up: For sub-assemblies burnt in significant flux gradients (e.g. fertile sub-assemblies) a detailed burn-up capability is available through specific ERANOS modules. Other topics: Several other features are available: - Coupled neutron/gamma Sn transport calculations (with specific libraries), - Detailed treatment of damage and kerma (with specific libraries), - Detailed burn-up with computation of decay (alpha, beta, gamma and neutron particles) activities, energies, energy spectra of emitted particles, dose rates (for simple geometries), decay heat (the MECCYCO package, with specific libraries). 2 - Methods: The methods used in the ERANOS modules have been mentioned shortly above. The user can feed and connect these modules in a variety of ways to produce specific analytic sequences. Conditional chaining (IF, FOR, WHILE instructions) is possible with the user's language. This allows a great deal of flexibility in the use of the code system

  4. Workload analyse of assembling process

    Science.gov (United States)

    Ghenghea, L. D.

    2015-11-01

    The workload is the most important indicator for managers responsible of industrial technological processes no matter if these are automated, mechanized or simply manual in each case, machines or workers will be in the focus of workload measurements. The paper deals with workload analyses made to a most part manual assembling technology for roller bearings assembling process, executed in a big company, with integrated bearings manufacturing processes. In this analyses the delay sample technique have been used to identify and divide all bearing assemblers activities, to get information about time parts from 480 minutes day work time that workers allow to each activity. The developed study shows some ways to increase the process productivity without supplementary investments and also indicated the process automation could be the solution to gain maximum productivity.

  5. Mitogenomic analyses from ancient DNA

    DEFF Research Database (Denmark)

    Paijmans, Johanna L. A.; Gilbert, Tom; Hofreiter, Michael

    2013-01-01

    The analysis of ancient DNA is playing an increasingly important role in conservation genetic, phylogenetic and population genetic analyses, as it allows incorporating extinct species into DNA sequence trees and adds time depth to population genetics studies. For many years, these types of DNA...... analyses (whether using modern or ancient DNA) were largely restricted to the analysis of short fragments of the mitochondrial genome. However, due to many technological advances during the past decade, a growing number of studies have explored the power of complete mitochondrial genome sequences...... yielded major progress with regard to both the phylogenetic positions of extinct species, as well as resolving population genetics questions in both extinct and extant species....

  6. Effectiveness of pollutants removal in hybrid constructed wetlands – different configurations case study

    Directory of Open Access Journals (Sweden)

    Gajewska Magdalena

    2017-01-01

    Full Text Available In recent years, an increase in interest in hybrid constructed wetland systems (HCWs has been observed. The aim of the paper is to compare different HCW configurations in terms of mass removal rates and efficiency of pollutants removal. Analysed data have been collected at multistage constructed wetlands in Poland, which are composed by at least two beds: horizontal subsurface flow (SSHF and vertical subsurface flow (SSVF. The evaluation was focused on hybrid constructed wetlands performance with HF+VF vs. VF+HF configuration, where influent wastewater of the same composition was treated. In analysed HCWs, the effective removal of organic matter from 75.2 to 91.6% COD was confirmed. Efficiency of total nitrogen removal varied from 47.3 to 91.7%. The most effective removal of TN (8.3 g m−2 d−1 occurred in the system with VF+VF+HF configuration.

  7. Recriticality analyses for CAPRA cores

    International Nuclear Information System (INIS)

    Maschek, W.; Thiem, D.

    1995-01-01

    The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

  8. Recriticality analyses for CAPRA cores

    Energy Technology Data Exchange (ETDEWEB)

    Maschek, W.; Thiem, D.

    1995-08-01

    The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

  9. Technical center for transportation analyses

    International Nuclear Information System (INIS)

    Foley, J.T.

    1978-01-01

    A description is presented of an information search/retrieval/research activity of Sandia Laboratories which provides technical environmental information which may be used in transportation risk analyses, environmental impact statements, development of design and test criteria for packaging of energy materials, and transportation mode research studies. General activities described are: (1) history of center development; (2) environmental information storage/retrieval system; (3) information searches; (4) data needs identification; and (5) field data acquisition system and applications

  10. Methodology of cost benefit analyses

    International Nuclear Information System (INIS)

    Patrik, M.; Babic, P.

    2000-10-01

    The report addresses financial aspects of proposed investments and other steps which are intended to contribute to nuclear safety. The aim is to provide introductory insight into the procedures and potential of cost-benefit analyses as a routine guide when making decisions on costly provisions as one of the tools to assess whether a particular provision is reasonable. The topic is applied to the nuclear power sector. (P.A.)

  11. Soil washing for brine removal

    International Nuclear Information System (INIS)

    Ayyachamy, J.S.; Atalay, A.; Zaman, M.

    1992-01-01

    During the exploration for oil and thereafter, brine transfer lines get ruptured releasing the brine which contaminates the surrounding soil. The salinity level in brine is very high, sometimes approaching or exceeding that of sea water. Soils contaminated with brine are unproductive and unsuitable for plant growth. Several investigators have documented the pollution of surface water and groundwater due to brine disposal from oil and needed to clean up such sites. The objective of this study is to develop a soil washing technique that can be used to remove brine sites were collected and used in the study. This paper reports on results which indicate that soil washing using various surface active agents is effective in removing the brine

  12. Chloride removal from plutonium alloy

    International Nuclear Information System (INIS)

    Holcomb, H.P.

    1983-01-01

    SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP

  13. After-heat removal system

    International Nuclear Information System (INIS)

    Yamamoto, Michiyoshi; Mitani, Shinji.

    1982-01-01

    Purpose: To prevent contamination of suppression pool water and intrusion of corrosion products into a nuclear reactor. Constitution: Upon stop of an after-heat removing system, reactor water contained in pipelines is drained out to a radioactive wastes processing facility at the time the cooling operation mode has been completed. At the same time, water is injected from a pure water supply system to the after-heat removing system to discharge corrosion product and activated materials while cleaning the inside of the pipelines. Then, pure water is held in the pipelines and it is discharged again and replaced with pure water before entering the cooling mode operation. Thereafter, the cooling mode operation upon reactor shutdown is performed. (Yoshino, Y.)

  14. Microalgae removal with Moringa oleifera.

    Science.gov (United States)

    Barrado-Moreno, M M; Beltran-Heredia, J; Martín-Gallardo, J

    2016-02-01

    Moringa oleifera seed extract was tested for algae (Chlorella, Microcystis, Oocystis and Scenedesmus) removal by Jar-test technique. This coagulant can be used in drinking water treatment. Jar-test has been carried out in order to evaluate the efficiency of this natural coagulant agent inside real surface water matrix. The influence of variables has been studied in this process, including operating parameters such as coagulant dosage, initial algae concentration, pH, agitation time and water matrix. Removal capacity is verified for water with high contamination of algae while the process is not affected by the pH and water matrix. Coagulation process may be modelling through Langmuir and Freundlich adsorption hypothesis, so acceptable r2 coefficients are obtained. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Hydrocarbon removal with constructed wetlands

    OpenAIRE

    Eke, Paul Emeka

    2008-01-01

    Wetlands have long played a significant role as natural purification systems, and have been effectively used to treat domestic, agricultural and industrial wastewater. However, very little is known about the biochemical processes involved, and the use of constructed treatment wetlands in the removal of petroleum aromatic hydrocarbons from produced and/or processed water. Wastewaters from the oil industry contain aromatic hydrocarbons such as benzene, toluene, ethylbenzene and x...

  16. Adsorptive Iron Removal from Groundwater

    OpenAIRE

    Sharma, S.K.

    2001-01-01

    Iron is commonly present in groundwater worldwide. The presence of iron in the water supply is not harmful to human health, however it is undesirable. Bad taste, discoloration, staining, deposition in the distribution system leading to aftergrowth, and incidences of high turbidity are some of the aesthetic and operational problems associated with iron in water supplies. Iron removal from groundwater is, therefore, a major concern for water supply companies using groundwater sources....

  17. Whole blood analysis rotor assembly having removable cellular sedimentation bowl

    Science.gov (United States)

    Burtis, C.A.; Johnson, W.F.

    1975-08-26

    A rotor assembly for performing photometric analyses using whole blood samples is described. Following static loading of a gross blood sample within a centrally located, removable, cell sedimentation bowl, the red blood cells in the gross sample are centrifugally separated from the plasma, the plasm displaced from the sedimentation bowl, and measured subvolumes of plasma distributed to respective sample analysis cuvettes positioned in an annular array about the rotor periphery. Means for adding reagents to the respective cuvettes are also described. (auth)

  18. Comparison of auxiliary feedwater and EDRS operation during natural circulation of MRX

    International Nuclear Information System (INIS)

    Kim, Jae Hak; Park, Goon Cherl

    1997-01-01

    The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control rod drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation

  19. Device for removing impurities from liquid metals

    International Nuclear Information System (INIS)

    Naito, Kesahiro; Yokota, Norikatsu; Shimoyashiki, Shigehiro; Takahashi, Kazuo; Ishida, Tomio.

    1984-01-01

    Purpose: To attain highly reliable and efficient impurity removal by forming temperature distribution the impurity removing device thereby providing the function of corrosion product trap, nuclear fission product trap and cold trap under the conditions suitable to the impurity removing materials. Constitution: The impurity removing device comprises a container containing impurity removing fillers. The fillers comprise material for removing corrosion products, material for removing nuclear fission products and material for removing depositions from liquid sodium. The positions for the respective materials are determined such that the materials are placed under the temperature conditions easy to attain their function depending on the temperature distribution formed in the removing device, whereby appropriate temperature condition is set to each of the materials. (Yoshino, Y.)

  20. Mechanism of nitrogen removal in wastewater lagoon: a case study.

    Science.gov (United States)

    Vendramelli, Richard A; Vijay, Saloni; Yuan, Qiuyan

    2017-06-01

    Ammonia being a nutrient facilitates the growth of algae in wastewater and causes eutrophication. Nitrate poses health risk if it is present in drinking water. Hence, nitrogen removal from wastewater is required. Lagoon wastewater treatment systems have become common in Canada these days. The study was conducted to understand the nitrogen removal mechanisms from the existing wastewater treatment lagoon system in the town of Lorette, Manitoba. The lagoon system consists of two primary aerated cells and two secondary unaerated cells. Surface samples were collected periodically from lagoon cells and analysed from 5 May 2015 to 9 November 2015. The windward and leeward sides of the ponds were sampled and the results were averaged. It was found that the free ammonia volatilization to the atmosphere is responsible for most of the ammonia removal. Ammonia and nitrate assimilation into biomass and biological growth in the cells appears to be the other mechanisms of nitrogen removal over the monitoring period. Factors affecting the nitrogen removal efficiency were found to be pH, temperature and hydraulic residence time. Also, the ammonia concentration in the effluent from the wastewater treatment lagoon was compared with the regulatory standard.

  1. Prehospital emergency removal of football helmets using two techniques.

    Science.gov (United States)

    Swartz, Erik E; Hernandez, Adam E; Decoster, Laura C; Mihalik, Jason P; Burns, Matthew F; Reynolds, Cathryn

    2011-01-01

    To compare the Eject Helmet Removal (EHR) System with manual football helmet removal. This quasiexperimental counterbalanced study was conducted in a controlled laboratory setting. Thirty certified athletic trainers (17 men and 13 women; mean ± standard deviation age: 33.03 ± 10.02 years; height: 174.53 ± 12.04 cm; mass: 85.19 ± 19.84 kg) participated after providing informed consent. Participants removed a Riddell Revolution IQ football helmet from a healthy model two times each under two conditions: manual helmet removal (MHR) and removal with the EHR system. A six-camera, three-dimensional motion capture system was used to record range of motion (ROM) of the head. A digital stopwatch was used to time trials and to record a split time associated with EHR system bladder insertion. A modified Borg CR10 scale was used to measure the rating of perceived exertion (RPE). Mean values were created for each variable. Three pairwise t-tests with Bonferroni-corrected alpha levels tested for differences between time for removal, split time, and RPE. A 2 x 3 (condition x plane) totally within-subjects repeated-measures design analysis of variance (ANOVA) tested for differences in head ROM between the sagittal, frontal, and transverse planes. Analyses were performed using SPSS (version 18.0) (alpha = 0.05). There was no statistically significant difference in perceived difficulty between EHR (RPE = 2.73) and MHR (RPE = 2.55) (t(29) = 0.76; p = 0.45; d = 0.20). Manual helmet removal was, on average, 28.95 seconds faster than EHR (t(29) = 11.44; p football helmets and in helmets used in other sports such as lacrosse, motorsports, and ice hockey.

  2. Equilibria in social belief removal [Journal article

    CSIR Research Space (South Africa)

    Booth, R

    2010-08-01

    Full Text Available removal function >i, which tells it how to remove any given sentence from its belief set. In this paper we view >i as a unary function on the set L of non- tautologous sentences, i.e., agents are never required to remove >. The result of removing 2 L... from i?s belief set is denoted by >i( ). We assume i?s initial belief set can always be recaptured from >i alone by just removing the (b) (1) (A) contradiction, i.e., i?s initial belief set is >i(?). We call any n-tuple (>i)i2A of removal functions a...

  3. Stochastic methods for the quantification of sensitivities and uncertainties in criticality analyses; Stochastische Methoden zur Quantifizierung von Sensitivitaeten und Unsicherheiten in Kritikalitaetsanalysen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Bock, Matthias; Stuke, Maik; Wagner, Markus

    2014-06-15

    This work describes statistical analyses based on Monte Carlo sampling methods for criticality safety analyses. The methods analyse a large number of calculations of a given problem with statistically varied model parameters to determine uncertainties and sensitivities of the computed results. The GRS development SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool) is a modular, easily extensible abstract interface program, designed to perform such Monte Carlo sampling based uncertainty and sensitivity analyses in the field of criticality safety. It couples different criticality and depletion codes commonly used in nuclear criticality safety assessments to the well-established GRS tool SUSA for sensitivity and uncertainty analyses. For uncertainty analyses of criticality calculations, SunCISTT couples various SCALE sequences developed at Oak Ridge National Laboratory and the general Monte Carlo N-particle transport code MCNP from Los Alamos National Laboratory to SUSA. The impact of manufacturing tolerances of a fuel assembly configuration on the neutron multiplication factor for the various sequences is shown. Uncertainties in nuclear inventories, dose rates, or decay heat can be investigated via the coupling of the GRS depletion system OREST to SUSA. Some results for a simplified irradiated Pressurized Water Reactor (PWR) UO{sub 2} fuel assembly are shown. SUnCISTT also combines the two aforementioned modules for burnup credit criticality analysis of spent nuclear fuel to ensures an uncertainty and sensitivity analysis using the variations of manufacturing tolerances in the burn-up code and criticality code simultaneously. Calculations and results for a storage cask loaded with typical irradiated PWR UO{sub 2} fuel are shown, including Monte Carlo sampled axial burn-up profiles. The application of SUnCISTT in the field of code validation, specifically, how it is applied to compare a simulation model to available benchmark

  4. Chapter No.4. Safety analyses

    International Nuclear Information System (INIS)

    2002-01-01

    In 2001 the activity in the field of safety analyses was focused on verification of the safety analyses reports for NPP V-2 Bohunice and NPP Mochovce concerning the new profiled fuel and probabilistic safety assessment study for NPP Mochovce. The calculation safety analyses were performed and expert reviews for the internal UJD needs were elaborated. An important part of work was performed also in solving of scientific and technical tasks appointed within bilateral projects of co-operation between UJD and its international partnership organisations as well as within international projects ordered and financed by the European Commission. All these activities served as an independent support for UJD in its deterministic and probabilistic safety assessment of nuclear installations. A special attention was paid to a review of probabilistic safety assessment study of level 1 for NPP Mochovce. The probabilistic safety analysis of NPP related to the full power operation was elaborated in the study and a contribution of the technical and operational improvements to the risk decreasing was quantified. A core damage frequency of the reactor was calculated and the dominant initiating events and accident sequences with the major contribution to the risk were determined. The target of the review was to determine the acceptance of the sources of input information, assumptions, models, data, analyses and obtained results, so that the probabilistic model could give a real picture of the NPP. The review of the study was performed in co-operation of UJD with the IAEA (IPSART mission) as well as with other external organisations, which were not involved in the elaboration of the reviewed document and probabilistic model of NPP. The review was made in accordance with the IAEA guidelines and methodical documents of UJD and US NRC. In the field of calculation safety analyses the UJD activity was focused on the analysis of an operational event, analyses of the selected accident scenarios

  5. Analysing the Wrongness of Killing

    DEFF Research Database (Denmark)

    Di Nucci, Ezio

    2014-01-01

    This article provides an in-depth analysis of the wrongness of killing by comparing different versions of three influential views: the traditional view that killing is always wrong; the liberal view that killing is wrong if and only if the victim does not want to be killed; and Don Marquis‟ future...... of value account of the wrongness of killing. In particular, I illustrate the advantages that a basic version of the liberal view and a basic version of the future of value account have over competing alternatives. Still, ultimately none of the views analysed here are satisfactory; but the different...

  6. Methodological challenges in carbohydrate analyses

    Directory of Open Access Journals (Sweden)

    Mary Beth Hall

    2007-07-01

    Full Text Available Carbohydrates can provide up to 80% of the dry matter in animal diets, yet their specific evaluation for research and diet formulation is only now becoming a focus in the animal sciences. Partitioning of dietary carbohydrates for nutritional purposes should reflect differences in digestion and fermentation characteristics and effects on animal performance. Key challenges to designating nutritionally important carbohydrate fractions include classifying the carbohydrates in terms of nutritional characteristics, and selecting analytical methods that describe the desired fraction. The relative lack of information on digestion characteristics of various carbohydrates and their interactions with other fractions in diets means that fractions will not soon be perfectly established. Developing a system of carbohydrate analysis that could be used across animal species could enhance the utility of analyses and amount of data we can obtain on dietary effects of carbohydrates. Based on quantities present in diets and apparent effects on animal performance, some nutritionally important classes of carbohydrates that may be valuable to measure include sugars, starch, fructans, insoluble fiber, and soluble fiber. Essential to selection of methods for these fractions is agreement on precisely what carbohydrates should be included in each. Each of these fractions has analyses that could potentially be used to measure them, but most of the available methods have weaknesses that must be evaluated to see if they are fatal and the assay is unusable, or if the assay still may be made workable. Factors we must consider as we seek to analyze carbohydrates to describe diets: Does the assay accurately measure the desired fraction? Is the assay for research, regulatory, or field use (affects considerations of acceptable costs and throughput? What are acceptable accuracy and variability of measures? Is the assay robust (enhances accuracy of values? For some carbohydrates, we

  7. Theorising and Analysing Academic Labour

    Directory of Open Access Journals (Sweden)

    Thomas Allmer

    2018-01-01

    Full Text Available The aim of this article is to contextualise universities historically within capitalism and to analyse academic labour and the deployment of digital media theoretically and critically. It argues that the post-war expansion of the university can be considered as medium and outcome of informational capitalism and as a dialectical development of social achievement and advanced commodification. The article strives to identify the class position of academic workers, introduces the distinction between academic work and labour, discusses the connection between academic, information and cultural work, and suggests a broad definition of university labour. It presents a theoretical model of working conditions that helps to systematically analyse the academic labour process and to provide an overview of working conditions at universities. The paper furthermore argues for the need to consider the development of education technologies as a dialectics of continuity and discontinuity, discusses the changing nature of the forces and relations of production, and the impact on the working conditions of academics in the digital university. Based on Erik Olin Wright’s inclusive approach of social transformation, the article concludes with the need to bring together anarchist, social democratic and revolutionary strategies for establishing a socialist university in a commons-based information society.

  8. CFD analyses in regulatory practice

    International Nuclear Information System (INIS)

    Bloemeling, F.; Pandazis, P.; Schaffrath, A.

    2012-01-01

    Numerical software is used in nuclear regulatory procedures for many problems in the fields of neutron physics, structural mechanics, thermal hydraulics etc. Among other things, the software is employed in dimensioning and designing systems and components and in simulating transients and accidents. In nuclear technology, analyses of this kind must meet strict requirements. Computational Fluid Dynamics (CFD) codes were developed for computing multidimensional flow processes of the type occurring in reactor cooling systems or in containments. Extensive experience has been accumulated by now in selected single-phase flow phenomena. At the present time, there is a need for development and validation with respect to the simulation of multi-phase and multi-component flows. As insufficient input by the user can lead to faulty results, the validity of the results and an assessment of uncertainties are guaranteed only through consistent application of so-called Best Practice Guidelines. The authors present the possibilities now available to CFD analyses in nuclear regulatory practice. This includes a discussion of the fundamental requirements to be met by numerical software, especially the demands upon computational analysis made by nuclear rules and regulations. In conclusion, 2 examples are presented of applications of CFD analysis to nuclear problems: Determining deboration in the condenser reflux mode of operation, and protection of the reactor pressure vessel (RPV) against brittle failure. (orig.)

  9. Efficient interruption of infection chains by targeted removal of central holdings in an animal trade network.

    Science.gov (United States)

    Büttner, Kathrin; Krieter, Joachim; Traulsen, Arne; Traulsen, Imke

    2013-01-01

    Centrality parameters in animal trade networks typically have right-skewed distributions, implying that these networks are highly resistant against the random removal of holdings, but vulnerable to the targeted removal of the most central holdings. In the present study, we analysed the structural changes of an animal trade network topology based on the targeted removal of holdings using specific centrality parameters in comparison to the random removal of holdings. Three different time periods were analysed: the three-year network, the yearly and the monthly networks. The aim of this study was to identify appropriate measures for the targeted removal, which lead to a rapid fragmentation of the network. Furthermore, the optimal combination of the removal of three holdings regardless of their centrality was identified. The results showed that centrality parameters based on ingoing trade contacts, e.g. in-degree, ingoing infection chain and ingoing closeness, were not suitable for a rapid fragmentation in all three time periods. More efficient was the removal based on parameters considering the outgoing trade contacts. In all networks, a maximum percentage of 7.0% (on average 5.2%) of the holdings had to be removed to reduce the size of the largest component by more than 75%. The smallest difference from the optimal combination for all three time periods was obtained by the removal based on out-degree with on average 1.4% removed holdings, followed by outgoing infection chain and outgoing closeness. The targeted removal using the betweenness centrality differed the most from the optimal combination in comparison to the other parameters which consider the outgoing trade contacts. Due to the pyramidal structure and the directed nature of the pork supply chain the most efficient interruption of the infection chain for all three time periods was obtained by using the targeted removal based on out-degree.

  10. Removing Dams, Constructing Science: Coproduction of Undammed Riverscapes by Politics, Finance, Environment, Society and Technology

    Directory of Open Access Journals (Sweden)

    Zbigniew J. Grabowski

    2017-10-01

    Full Text Available Dam removal in the United States has continued to increase in pace and scope, transitioning from a dam-safety engineering practice to an integral component of many large-scale river restoration programmes. At the same time, knowledge around dam removals remains fragmented by disciplinary silos and a lack of knowledge transfer between communities of practice around dam removal and academia. Here we argue that dam removal science, as a study of large restoration-oriented infrastructure interventions, requires the construction of an interdisciplinary framework to integrate knowledge relevant to decision-making on dam removal. Drawing upon infrastructure studies, relational theories of coproduction of knowledge and social life, and advances within restoration ecology and dam removal science, we present a preliminary framework of dams as systems with irreducibly interrelated political, financial, environmental, social, and technological dimensions (PFESTS. With this framework we analyse three dam removals occurring over a similar time period and within the same narrow geographic region (the Mid-Columbia Region in WA and OR, USA to demonstrate how each PFESTS dimension contributed to the decision to remove the dam, how it affected the process of removing the dam, and how those dimensions continue to operate post removal in each watershed. We conclude with a discussion of a joint research and practice agenda emerging out of the PFESTS framing.

  11. Complexed iron removal from groundwater

    Energy Technology Data Exchange (ETDEWEB)

    Munter, R.; Ojaste, H.; Sutt, J. [Tallinn Technical University, Tallinn (Estonia). Dept. of Environmental & Chemical Technology

    2005-07-01

    The paper demonstrates an intensive work carried out and results obtained on the pilot plant of the City of Kogalym Water Treatment Station (Tjumen, Siberia, Russian Federation) to elaborate on a contemporary nonreagent treatment technology for the local iron-rich groundwater. Several filter materials (Birm, Pyrolox, hydroanthracite, Everzit, granulated activated carbon) and chemical oxidants (ozone, chlorine, hydrogen peroxide, oxygen, and potassium permanganate) were tested to solve the problem with complexed iron removal from groundwater. The final elaborated technology consists of raw water intensive aeration in the gas-degas treatment unit followed by sequential filtration through hydroanthracite and the special anthracite Everzit.

  12. Heat removing under hypersonic conditions

    Directory of Open Access Journals (Sweden)

    Semenov Mikhail E.

    2016-01-01

    Full Text Available In this paper we consider the heat transfer properties of the axially symmetric body with parabolic shape at hypersonic speeds (with a Mach number M > 5. We use the numerical methods based on the implicit difference scheme (Fedorenko method with direct method based on LU-decomposition and iterative method based on the Gauss-Seigel method. Our numerical results show that the heat removing process should be performed in accordance with the nonlinear law of heat distribution over the surface taking into account the hypersonic conditions of motion.

  13. South American Source Removal Project

    International Nuclear Information System (INIS)

    Nader, Alejandro V.

    2017-01-01

    Main objective of the project: •Thanks to Canada funding and IAEA technical assistance the main objective is to remove 29 disused sealed radioactive sources (DSRS), from 5 member states in Latin America region (Bolivia - Ecuador - Paraguay – Peru – Uruguay) to an authorized recipient for their final management. •It includes packaging of the DSRS and the DU working shields, customs arrangement for the export from the respective countries and import to the final destination in the Authorized Recipient’s country, transportation, deposit and hand over to an Authorized Recipient

  14. Removal of bound metal fasteners

    Science.gov (United States)

    Kramer, R. F.

    1981-04-01

    This project explored the removal of bound metal fasteners through the use of ultrasonically assisted wrenches. Two wrenches were designed, fabricated and tested. Previous studies had indicated an increase in thread tension for a given torque application under the influence of ultrasonics. Based on this, the loosening of seized and corroded fasteners with the aid of ultrasonics was explored. Experimental data confirmed our prior analysis of the torque-tension relationship under the influence of ultrasonics; however, our progress did not satisfy the requirements necessary to loosen seized studs in a shipyard environment.

  15. Photodynamic therapy for hair removal

    Directory of Open Access Journals (Sweden)

    Mohamed H. M. Ali

    2013-05-01

    Full Text Available Background: Unwanted hair is one of the most common medical problems affecting women of reproductive age inducing a lot of psychological stress and threatening their femininity and self-esteem. Old methods of removing unwanted hair include shaving, waxing, chemical depilation, and electrolysis, all of which have temporary results. However laser-assisted hair removal is the most efficient method of long-term hair removal currently available. It is desirable to develop a reduced cost photodynamic therapy (PDT system whose properties should include high efficiency and low side-effects. Method: Mice skin tissues were used in this study and divided into six groups such as controls, free methylene blue (MB incubation, liposome methylene blue (MB incubation, laser without methylene blue (MB, free methylene blue (MB for 3 and 4 hrs and laser, liposome methylene blue (MB for 3 hrs and laser. Methylene blue (MBwas applied to wax epilated areas. The areas were irradiated with CW He-Ne laser system that emits orange-red light with wavelength 632.8 nm and 10 mW at energy density of 5 J/ cm2 for 10 minutes. The UV-visible absorption spectrum was collected by Cary spectrophotometer. Results: Methylene blue (MB is selectively absorbed by actively growing hair follicles due to its cationic property. Methylene blue (MBuntreated sections showed that hair follicle and sebaceous gland are intact and there is no change due to the laser exposure. Free methylene blue (MB sections incubated for 3 hrs showed that He:Ne laser induced destruction in hair follicles, leaving an intact epidermis. Treated section with free methylene blue (MB for 4 hrs showed degeneration and necrosis in hair follicle, leaving an intact epidermis. Liposomal methylene blue (MB sections incubated for 3 hrs showed He:Ne laser induced destruction in hair follicles with intradermal leucocytic infiltration. Conclusions: Low power CW He:Ne laser and methylene blue (MB offered a successful PDT system

  16. Radioiodine removal in nuclear facilities

    International Nuclear Information System (INIS)

    1980-01-01

    Technical means are reviewed available for the retention of radioiodine in nuclear power plants and fuel reprocessing plants, its immobilization, storage, and disposal. The removal of iodine species from gaseous effluents of nuclear power plants using impregnated activated charcoal is dealt with. Various scrubbing techniques for trapping iodine from the head-end and dissolver off-gases are discussed as well as solid adsorbents for iodine which may be used to clean up other gaseous streams. Current practices and activities for radioiodine treatment and management in Belgian, Dutch, Swedish, USSR and UK nuclear installations are presented

  17. Laser-based coatings removal

    International Nuclear Information System (INIS)

    Freiwald, J.G.; Freiwald, D.

    1995-01-01

    Over the years as building and equipment surfaces became contaminated with low levels of uranium or plutonium dust, coats of paint were applied to stabilize the contaminants in place. Most of the earlier paint used was lead-based paint. More recently, various non-lead-based paints, such as two-part epoxy, are used. For D ampersand D (decontamination and decommissioning), it is desirable to remove the paints or other coatings rather than having to tear down and dispose of the entire building

  18. Removing Pubic Hair (For Young Men)

    Science.gov (United States)

    ... who has experience with performing laser hair removal. Electrolysis: Electrolysis is the only hair removal method that permanently ... using slow strokes. Rinse your skin with warm water after you are done shaving and then pat ...

  19. THE REMOVAL OF GLYPHOSATE FROM DRINKING WATER

    Science.gov (United States)

    The effectiveness of granulated activated carbon (GAC), packed activated carbon (PAC), conventional treatment, membranes, and oxidation for removing glyphosate from natural waters is evaluated. Results indicate that GAC and PAC are not effective in removing glyphosate, while oxid...

  20. Investigating the removal of body piercings.

    Science.gov (United States)

    Armstrong, Myrna L; Roberts, Alden E; Koch, Jerome R; Saunders, Jana C; Owen, Donna C

    2007-05-01

    Although body piercing procurement continues to increase, 13% to 18% of them are removed. Reasons for piercing removal in college students were examined with three groups: (a) those who kept all their piercings, (b) those who removed some, or (c) those who removed all of their body piercings. Of the sample, 41% were still pierced; 50% in their lifetime. Their major purpose for the body piercing was "helped them feel unique." Females obtained more (in high school) and then removed more, usually as upperclassmen. Males and females reported themselves as risk takers at procedure time and currently; however, only 10% cited deviancy as a reason for the body piercing(s). Only removal elements of "I just got tired of it" and "I just decided to remove it" were present, especially with the Some Removed Group. Further examination of body piercing building personal distinctiveness and self-identity to promote their need of uniqueness is suggested.