WorldWideScience

Sample records for activation decay heat

  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. 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

  3. 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.

  4. 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)

  5. 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

  6. 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.

  7. 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.

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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.

  15. 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)

  16. 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

  17. 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)

  18. 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)

  19. 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...

  20. 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

  1. 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...

  2. 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.)

  3. 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)

  4. 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.

  5. 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.

  6. 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...

  7. 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)

  8. 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)

  9. 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)

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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.)

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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.

  13. 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

  14. 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

  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. 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

  17. 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.

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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)

  3. WAD, a program to calculate the heat produced by alpha decay

    International Nuclear Information System (INIS)

    Jarvis, R.G.; Bretzlaff, C.I.

    1982-09-01

    The FORTRAN program WAD (Watts from Alpha Decay) deals with the alpha and beta decay chains to be encountered in advanced fuel cycles for CANDU reactors. The data library covers all necessary alpha-emitting and beta-emitting nuclides and the program calculates the heat produced by alpha decay. Any permissible chain can be constructed very simply

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. 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.

  19. 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.

  20. 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

  1. 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)

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. Parametric decay instabilities in ECR heated plasmas

    International Nuclear Information System (INIS)

    Porkolab, M.

    1982-01-01

    The possibility of parametric excitation of electron Bernstein waves and low frequency ion oscillations during ECR heating at omega/sub o/ approx. = l omega/sub ce/, l = 1,2 is examined. In particular, the thresholds for such instabilities are calculated. It is found that Bernstein waves and lower hybrid quasi-modes have relatively low homogeneous where T/sub e/ approx. = T/sub i/. Thus, these processes may lead to nonlinear absorption and/or scattering of the incident pump wave. The resulting Bernstein waves may lead to either more effective heating (especially during the start-up phase) or to loss of microwave energy if the decay waves propagate out of the system before their energy is absorbed by particles. While at omega/sub o/ = omega/sub UH/ the threshold is reduced due to the WKB enhancement of the pump wave, (and this instability may be important in tokamaks) in EBT's and tandem mirrors the instability at omega /sub o/ greater than or equal to 2 omega/sub ce/ may be important. The instability may persist even if omega > 2 omega/sub ce/ and this may be the case during finite beta depression of the magnetic field in which case the decay waves may be trapped in the local magnetic well so that convective losses are minimized. The excited fluctuations may lead to additional scattering of the ring electrons and the incident microwave fields. Application of these calculations to ECR heating of tokamaks, tandem mirrors, and EBT's will be examined

  9. 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)

  10. 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.

  11. 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.

  12. 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%

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  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. 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

  20. 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.)

  1. 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

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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.

  17. Effect of heat-treatment with raw cotton seed oil on decay resistance and dimensional stability of Beech (Fagus orientalis

    Directory of Open Access Journals (Sweden)

    مریم قربانی

    2015-05-01

    Full Text Available This research was conducted to determine the effect of heat-treatment with raw cotton seed oil on decay resistance and dimensional stability of beech according to EN113 and ASTM-D1037 standards respectively. The heat treatment with raw cotton seed oil was carried out in the cylinder at the temperatures of 130 and 170oC for 30 and 60 minutes. Oil uptake, density, volumetric swelling, water absorption and weight loss exposed to decay were measured. Oil uptake at 30 and 60 min were determined 10.5 and 13.3 Kg/cm3 respectively. Oil-heat treated samples at 30min and 130°C indicated the maximum density with 87.7% increase. According to results, oil-heat treatment improved water repellency and dimensional stability. Water absorption in 130°C and 60 minutes decreased 76% in comparison with control. Decay resistance of oil soaked samples for 60minutes was 80.2% more than control samples. Oil-heat treatment compared with oil treatment improved decay resistance, this effect was significant at 30 min. The temperature rise of oil–heat treatment at 30 minutes improved decay resistance, but the improvement under same level of temperature with increase time was not significant.

  18. 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.

  19. 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)

  20. 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.

  1. 'Thermal ghosts': apparent decay of fixed surfaces caused by heat diffusion

    International Nuclear Information System (INIS)

    Livadiotis, George

    2007-01-01

    The behaviour concerning classical heat diffusion on fixed thermal surfaces, studied by observations, still holds surprises. As soon as convective and radiative processes are negligible within the medium, this is considered to be free from energy sources and sinks. Then, the heat diffusion equation is conveniently solved using standard Fourier methods. Some considerations about the contrast effect suggest that the surface boundary would rather be observed to follow specific area decay dynamics than remaining fixed and static. Here it is shown that the apparent boundary lies on a specific isothermal spatiotemporal curve, which depends on the observing device. This is characterized by a slight, though determinative, difference between its radiance and that of the ambient background. Thereafter, the heat diffusion yields apparent boundary shrinkage with the passing of time. This phenomenon is particularly notable for two reasons: its lifetime and final decay rate depend only on the medium thermal properties, while being independent of the apparent boundary spatiotemporal curve. Thus, the former provides a suitable method for measuring the medium thermal properties via the observational data. The latter strongly reveal a kind of universality of some characteristic properties of the phenomenon, common to all observers

  2. 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

  3. 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

  4. 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

  5. Heat generation and heating limits for the IRUS LLRW disposal facility

    International Nuclear Information System (INIS)

    Donders, R.E.; Caron, F.

    1995-10-01

    Heat generation from radioactive decay and chemical degradation must be considered when implementing low-level radioactive waste (LLRW) disposal. This is particularly important when considering the management of spent radioisotope sources. Heating considerations and temperature calculations for the proposed IRUS (Intrusion Resistant Underground Structure) near-surface disposal facility are presented. Heat transfer calculations were performed using a finite element code with realistic but somewhat conservative heat transfer parameters and environmental boundary conditions. The softening-temperature of the bitumen waste-form (38 deg C) was found to be the factor that limits the heat generation rate in the facility. This limits the IRUS heat rate, assuming a uniform source term, to 0.34 W/m 3 . If a reduced general heat-limit is considered, then some higher-heat packages can be accepted with restrictions placed on their location within the facility. For most LLRW, heat generation from radioactive decay and degradation are a small fraction of the IRUS heating limits. However, heating restrictions will impact on the disposal of higher-activity radioactive sources. High activity 60 Co sources will require decay-storage periods of about 70 years, and some 137 Cs will need to bed disposed of in facilities designed for higher-heat waste. (author). 21 refs., 8 tabs., 2 figs

  6. 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)

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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)

  19. 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)

  20. 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).

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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).

  7. Safety aspects of targets for ADTT: Activity, volatile products, residual heat release

    International Nuclear Information System (INIS)

    Gai, E.V.; Ignatyuk, A.V.; Lunev, V.P.; Shubin, Yu.N.

    1999-01-01

    Safety aspects of heavy metal liquid targets for the accelerator driven systems connected with the activity accumulation and residual energy release due to the irradiation with high energy proton beam are discussed. The results obtained for the lead-bismuth target that are under construction in IPPE now in the frame of ISTC Project No. 559 are briefly presented. The calculations and the analysis of the accumulation of the spallation reaction products, activity and energy release at various moments after the accelerator shutdown are presented. The concentrations of the reaction products, the total and partial activities, the activities of volatile products are determined. The contributions of the short-lived nuclides important for the prediction of the facility behaviour in regimes with the accelerator beam trips. The calculations and analysis of the residual energy release due to different decay type have been performed. The conclusions are as follows. The obtained results showed that long lived radioactivity accumulates mainly due to primary nuclear reactions. Secondary reactions are responsible for the production of small number of long-lived isotopes Bi-207, Po-210 and some others, being generated by radiative capture of low energy neutrons. It is possible to make a conclusion that neutrons in the energy range 20 - 800 MeV and protons with energy above 100 MeV give main contribution to the total activity generation although these parts of spectra inside the target give comparatively small contribution to the total flux. The correct consideration of short-lived nuclides contribution is the main problem in the analysis of the target behaviour in the case of short accelerator shutdowns. They make the determining contribution to the both activity and the heat release at the first moments after the accelerator shutdown, creating the intermediate links and additional channels for the long-lived nuclides accumulation chains. The strong dependence of calculated

  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. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  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. Understanding decay resistance, dimensional stability and strength changes in heat treated and acetylated wood

    Science.gov (United States)

    Roger M. Rowell; Rebecca E. Ibach; James McSweeny; Thomas Nilsson

    2009-01-01

    Reductions in hygroscopicity, increased dimensional stability and decay resistance of heat-treated wood depend on decomposition of a large portion of the hemicelluloses in the wood cell wall. In theory, these hemicelluloses are converted to small organic molecules, water and volatile furan-type intermediates that can polymerize in the cell wall. Reductions in...

  17. 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

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. JENDL FP decay data file 2000 and the beta-decay theory

    International Nuclear Information System (INIS)

    Yoshida, Tadashi; Katakura, Jun Ichi; Tachibana, Takahiro

    2002-01-01

    JENDL FP Decay Data File 2000 has been developed as one of the special purpose files of the Japanese Evaluated Nuclear Data Library (JENDL), which constitutes a versatile nuclear data basis for science and technology. In the format of ENDF-6 this file includes the decay data for 1087 unstable fission product (FP) nuclides and 142 stable nuclides as their daughters. The primary purpose of this file is to use in the summation calculation of FP decay heat, which plays a critical role in nuclear safety analysis; the loss-of-coolant accident analysis of reactors, for example. The data for a given nuclide are its decay modes, the Q value, the branching ratios, the average energies released in the form of beta- and gamma-rays per decay, and their spectral data. The primary source of the decay data adopted here is the ENSDF (Evaluated Nuclear Structure Data File). The data in ENSDF, however, cover only the measured values. The data of the short-lived nuclides, which are essential for the decay heat calculations at short cooling times, are often fully lacking or incomplete even if they exist. This is mainly because of their short half-life nature. For such nuclides a theoretical model calculation is applied in order to fill the gaps between the true and the experimentally known decay schemes. In practice we have to predict the average decay energies and the spectral data for a lot of short-lived FPs by use of beta-decay theories. Thus the beta-decay theory plays a very important role in generating the FP decay data file

  9. Physical distributions of radon decay chain activities in air

    International Nuclear Information System (INIS)

    Rolle, R.

    2004-01-01

    The distribution of short-lived radon decay chain activities in air - in time, space and on aerosols - determines their exposure potential and measurement thereof. The radioactive decay constants and flow variables in a flow system combine, yielding activity concentration distributions and ratios of concentrations characteristic of the flow scheme, its source(s) and sink(s). The clock of 'internal' decay constants allows the unraveling of characteristics of the flow scheme from activity concentration measurements of individual members of a decay chain. Basic flow string calculations are shown. These can be assembled to define or simulate concentrations in a single- or multiple-compartment flow network. Response calculations to single- and multiple-step, or continuous changes in sources and sinks yield time-, spatial- and attachment-distributions. For the short-lived 222 Rn and 212 Pb decay chains the decay constants of the shorter-lived progeny in relation to the parent impose air activity ratios on successive chain members. Ratio limits had been used in the past to improve older grab-sampling- or integral gross-alpha measurement procedures for assessing exposure level. Assessment of individual concentrations, ratios and their distributions enables unravelling of dynamic flow systems, with restriction from the range of the parameters of flow and decay. An activity measuring instrument by itself represents a flow system with a response time distribution. Instrument response correction during continuous or quasi-continuous sampling and continuous spectrometric measurement allows far more accurate time-resolved measurement evaluation of continuously varying air concentrations, than previously attainable. Strong diurnal or even shorter (≤ 1 hr) changes probably are the norm in indoor and outdoor air activity concentrations. A mere average response evaluation, as used in steady state instrument calibration, and using less efficient instruments, is usually inadequate

  10. 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)

  11. 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

  12. 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

  13. Parametric Decay during HHFW on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bernabei, S.; Biewer, T.; Diem, S.; Hosea, J.; LeBlanc, B.; Phillips, C.K.; Ryan, P.; Swain, D.W.

    2005-01-01

    High Harmonic Fast Wave (HHFW) heating experiments on NSTX have been observed to be accompanied by significant edge ion heating (T i >> T e ). This heating is found to be anisotropic with T perp > T par . Simultaneously, coherent oscillations have been detected with an edge Langmuir probe. The oscillations are consistent with parametric decay of the incident fast wave (ω > 13ω ci ) into ion Bernstein waves and an unobserved ion-cyclotron quasi-mode. The observation of anisotropic heating is consistent with Bernstein wave damping, and the Bernstein waves should completely damp in the plasma periphery as they propagate toward a cyclotron harmonic resonance. The number of daughter waves is found to increase with rf power, and to increase as the incident wave's toroidal wavelength increases. The frequencies of the daughter wave are separated by the edge ion cyclotron frequency. Theoretical calculations of the threshold for this decay in uniform plasma indicate an extremely small value of incident power should be required to drive the instability. While such decays are commonly observed at lower harmonics in conventional ICRF heating scenarios, they usually do not involve the loss of significant wave power from the pump wave. On NSTX an estimate of the power loss can be found by calculating the minimum power required to support the edge ion heating (presumed to come from the decay Bernstein wave). This calculation indicates at least 20-30% of the incident rf power ends up as decay waves

  14. Effect of tin oxide nano particles and heat treatment on decay resistance and physical properties of beech wood (Fagus orientalis

    Directory of Open Access Journals (Sweden)

    Maryam Ghorbani

    2014-11-01

    Full Text Available This research was conducted to investigate the effect of Tin oxide nanoparticles and heat treatment on decay resistance and physical properties of beech wood. Biological and physical test samples were prepared according to EN-113 and ASTM-D4446-05 standards respectively. Samples were classified into 4 groups: control, impregnation with Tin oxide nanoparticles, heat treatment and nano-heat treatment. Impregnation with Tin oxide nano at 5000ppm concentration was carried out in the cylinder according to Bethell method. Then, samples were heated at 140, 160 and 185˚C for 2 and 4 hours. According to results, decay resistance improved with increasing time and temperature of heat treatment. Least weight loss showed 46.39% reduction in nano-heat samples treated at 180˚C for 4 hours in comparison with control at highest weight loss. Nano-heat treated samples demonstrated the maximum amount of water absorption without significant difference with control and nanoparticles treated samples. Increase in heat treatment temperature reduced water absorption so that it is revealed 47.8% reduction in heat treated samples at 180°C for 4h after 24h immersion in water. In nano-heat treated samples at 180˚C for 2h was measured least volume swelling. Volume swelling in nano-treated samples decreased 8.7 and 22.76% after 2 and 24 h immersion in comparison with the control samples respectively.

  15. 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.

  16. 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)

  17. Heating tokamaks by parametric decay of intense extraordinary mode radiation

    International Nuclear Information System (INIS)

    Elder, G.B.; Perkins, F.W.

    1979-08-01

    Intense electron beam technology has developed coherent, very high power (350 megawatts) microwave sources at frequencies which are a modest fraction of the electron cyclotron frequency in tokamaks. Propagation into a plasma occurs via the extraordinary mode which is subject to parametric decay instabilities in the density range ω/sub o/ 2 2 < ω/sub o/(ω/sub o/ + Ω/sub e/). For an incident wave focused onto a hot spot by a dish antenna of radius rho, the effective threshold power P/sub o/ required to induced effective parametric heating is P/sub o/ approx. = 10 MW x/rho Ω/sub e//ω/sub o/ (T/sub e//1 keV)/sup 3/2/ where x denotes the distance to the hot spot

  18. Phase coherence of parametric-decay modes during high-harmonic fast-wave heating in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, J. A., E-mail: carlsson@pppl.gov [Crow Radio and Plasma Science, Princeton, New Jersey 08540 (United States); Wilson, J. R.; Hosea, J. C.; Greenough, N. L.; Perkins, R. J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States)

    2016-06-15

    Third-order spectral analysis, in particular, the auto bicoherence, was applied to probe signals from high-harmonic fast-wave heating experiments in the National Spherical Torus Experiment. Strong evidence was found for parametric decay of the 30 MHz radio-frequency (RF) pump wave, with a low-frequency daughter wave at 2.7 MHz, the local majority-ion cyclotron frequency. The primary decay modes have auto bicoherence values around 0.85, very close to the theoretical value of one, which corresponds to total phase coherence with the pump wave. The threshold RF pump power for onset of parametric decay was found to be between 200 kW and 400 kW.

  19. The effects of moderately high temperature on zeaxanthin accumulation and decay.

    Science.gov (United States)

    Zhang, Ru; Kramer, David M; Cruz, Jeffrey A; Struck, Kimberly R; Sharkey, Thomas D

    2011-09-01

    Moderately high temperature reduces photosynthetic capacities of leaves with large effects on thylakoid reactions of photosynthesis, including xanthophyll conversion in the lipid phase of the thylakoid membrane. In previous studies, we have found that leaf temperature of 40°C increased zeaxanthin accumulation in dark-adapted, intact tobacco leaves following a brief illumination, but did not change the amount of zeaxanthin in light-adatped leaves. To investigate heat effects on zeaxanthin accumulation and decay, zeaxanthin level was monitored optically in dark-adapted, intact tobacco and Arabidopsis thaliana leaves at either 23 or 40°C under 45-min illumination. Heated leaves had more zeaxanthin following 3-min light but had less or comparable amounts of zeaxanthin by the end of 45 min of illumination. Zeaxanthin accumulated faster at light initiation and decayed faster upon darkening in leaves at 40°C than leaves at 23°C, indicating that heat increased the activities of both violaxanthin de-epoxidase (VDE) and zeaxanthin epoxidase (ZE). In addition, our optical measurement demonstrated in vivo that weak light enhances zeaxanthin decay relative to darkness in intact leaves of tobacco and Arabidopsis, confirming previous observations in isolated spinach chloroplasts. However, the maximum rate of decay is similar for weak light and darkness, and we used the maximum rate of decay following darkness as a measure of the rate of ZE during steady-state light. A simulation indicated that high temperature should cause a large shift in the pH dependence of the amount of zeaxanthin in leaves because of differential effects on VDE and ZE. This allows for the reduction in ΔpH caused by heat to be offset by increased VDE activity relative to ZE.

  20. Improvement of decay and cross-section data libraries for activation calculations

    International Nuclear Information System (INIS)

    Attaya, H.

    1993-01-01

    A new decay data library has been completed. The new library contains up-to-date decay information (half-lives, branching ratios, decay energies, γ's energies and intensities). Activation responses such as the air and water biological hazard potentials, the waste disposal rating, and the biological dose are also included in this library. Recently developed cross-section libraries have been acquired to be used together with the decay data library

  1. 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)

  2. Heat sink considerations during SLAR defuelling activities of the 1995 maintenance outage at the Point Lepreau G.S

    International Nuclear Information System (INIS)

    Richard, R.A.; Edgar, D.J.; Gibb, R.A.

    1996-01-01

    In order to conduct the Spacer Locate and Reposition (SLAR) activities, individual fuel channels were de-fuelled using the SLAR tool itself. The presence of the tool in the channel containing fuel caused concerns over channel flow interruption. A methodology was devised to assess the limited fuel heat up during de-fuelling and during possible prolonged periods of complete flow interruption. Once established, the approach could be used to assess the effect of varying parameters which could effect channel heat-up. Costly delays in SLAR maintenance activities were avoided by ensuring the sequence of channel visits was such that the individual channel powers decayed sufficiently to prevent channel heat up beyond acceptable temperature limits. (author)

  3. Study of the influence of decay data in activation reaction cross section

    International Nuclear Information System (INIS)

    Huang Xiaolong; Lu Hanlin

    2002-01-01

    The effect of the decay data on the measured activation cross section is investigated carefully and testified by several examples. These decay data include the half-life of the product, γ branching ratio of the product and decay scheme. Present work shows that these effects must be considered carefully when evaluating the activation reaction cross section. Sometimes they are main reason for causing the discrepancies among the experimental data

  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. Experimental observation of microwave absorption and electron heating due to the two plasmon decay instability and resonance absorption

    International Nuclear Information System (INIS)

    Rasmussen, D.A.

    1981-01-01

    The interaction of intense microwaves with an inhomogeneous plasma is studied in two experimental devices. In the first device an investigation was made of microwave absorption and electron heating due to the parametric decay of microwaves into electron plasma waves (Two Plasmon Decay instability, TPDI), modeling a process which can occur near the quarter critical surface in laser driven pellets. P-polarized microwave (f = 1.2 GHz, P 0 less than or equal to 12 kW) are applied to an essentially collisionless, inhomogeneous plasma, in an oversized waveguide, in the U.C. Davis Prometheus III device. The initial density scale length near the quarter critical surface is quite long (L/lambda/sub De/ approx. = 3000 or k 0 L approx. = 15). The observed threshold power for the TPDI is quite low (P/sub T/approx. = 0.1 kW or v/sub os//v/sub e/ approx. = 0.1). Near the threshold the decay waves only occur near the quarter critical surface. As the incident power is increased above threshold, the decay waves spread to lower densities, and for P 0 greater than or equal to lkW, (v/sub os//v/sub e/ greater than or equal to 0.3) suprathermal electron heating is strong for high powers (T/sub H/ less than or equal to 12 T/sub e/ for P 0 less than or equal to 8 kW or v/sub os//v/sub e/ less than or equal to 0.9)

  6. JNDC FP decay data file

    International Nuclear Information System (INIS)

    Yamamoto, Tohru; Akiyama, Masatsugu

    1981-02-01

    The decay data file for fission product nuclides (FP DECAY DATA FILE) has been prepared for summation calculation of the decay heat of fission products. The average energies released in β- and γ-transitions have been calculated with computer code PROFP. The calculated results and necessary information have been arranged in tabular form together with the estimated results for 470 nuclides of which decay data are not available experimentally. (author)

  7. 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)

  8. 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)

  9. The file of evaluated decay data in ENDF/B

    International Nuclear Information System (INIS)

    Reich, C.W.

    1991-01-01

    One important application of nuclear decay data is the Evaluated Nuclear Data File/B (ENDF/B), the base of evaluated nuclear data used in reactor research and technology activities within the United States. The decay data in the Activation File (158 nuclides) and the Actinide File (108 nuclides) excellently represent the current status of this information. In particular, the half-lives and gamma and alpha emission probabilities, quantities that are so important for many applications, of the actinide nuclides represent a significant improvement over those in ENDF/B-V because of the inclusion of data produced by an International Atomic Energy Agency Coordinated Research Program. The Fission Product File contains experimental decay data on ∼510 nuclides, which is essentially all for which a meaningful number of data are available. For the first time, delayed-neutron spectra for the precursor nuclides are included. Some hint of problems in the fission product data base is provided by the gamma decay heat following a burst irradiation of 239 Pu

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. Evaluating death and activity decay of Anammox bacteria during anaerobic and aerobic starvation.

    Science.gov (United States)

    Wang, Qilin; Song, Kang; Hao, Xiaodi; Wei, Jing; Pijuan, Maite; van Loosdrecht, Mark C M; Zhao, Huijun

    2018-06-01

    The decreased activity (i.e. decay) of anaerobic ammonium oxidation (Anammox) bacteria during starvation can be attributed to death (i.e. decrease in the amount of viable bacteria) and activity decay (i.e. decrease in the specific activity of viable bacteria). Although they are crucial for the operation of the Anammox process, they have never been comprehensively investigated. This study for the first time experimentally assessed death and activity decay of the Anammox bacteria during 84 days' starvation stress based on ammonium removal rate, Live/Dead staining and fluorescence in-situ hybridization. The anaerobic and aerobic decay rates of Anammox bacteria were determined as 0.015 ± 0.001 d -1 and 0.028 ± 0.001 d -1 , respectively, indicating Anammox bacteria would lose their activity more quickly in the aerobic starvation than in the anaerobic starvation. The anaerobic and aerobic death rates of Anammox bacteria were measured at 0.011 ± 0.001 d -1 and 0.025 ± 0.001 d -1 , respectively, while their anaerobic and aerobic activity decay rates were determined at 0.004 ± 0.001 d -1 and 0.003 ± 0.001 d -1 , respectively. Further analysis revealed that death accounted for 73 ± 4% and 89 ± 5% of the decreased activity of Anammox bacteria during anaerobic and aerobic starvations, and activity decay was only responsible for 27 ± 4% and 11 ± 5% of the decreased Anammox activity, respectively, over the same starvation periods. These deeply shed light on the response of Anammox bacteria to the starvation stress, which would facilitate operation and optimization of the Anammox process. Copyright © 2018 Elsevier Ltd. All rights reserved.

  15. 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)

  16. Electron heating caused by the ion-acoustic decay instability in a finite-length system

    International Nuclear Information System (INIS)

    Rambo, P.W.; Woo, W.; DeGroot, J.S.; Mizuno, K.

    1984-01-01

    The ion-acoustic decay instability is investigated for a finite-length plasma with density somewhat below the cutoff density of the electromagnetic driver (napprox.0.7n/sub c/). For this regime, the heating in a very long system can overpopulate the electron tail and cause linear saturation of the low phase velocity electron plasma waves. For a short system, the instability is nonlinearly saturated at larger amplitude by ion trapping. Absorption can be significantly increased by the large-amplitude ion waves. These results compare favorably with microwave experiments

  17. 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.)

  18. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    Directory of Open Access Journals (Sweden)

    Frisoni Manuela

    2016-01-01

    Full Text Available ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1 containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2 containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E are shown and discussed in this paper.

  19. 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

  20. 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

  1. Determination of activation energies of K2YF5: Tb by the isothermal decay method

    International Nuclear Information System (INIS)

    Gallegos, A.A.; Azorin, J.; Khaidukov, M.

    2007-01-01

    Full text: In the search of thermoluminescent materials (TL) for use in personal dosimetry, were studied different materials such as the lithium fluoride doped with magnesium and titanium (LiF: Mg, Ti), the one which had been studied and used as reference material by their good equivalence with the tissue although has poor sensibility, this has generated the necessity to look for materials that show a considerable improvement as for their TL response. In this search the double fluorides as the double lithium potassium fluorides and yttrium doped with praseodymium (LiKYF 5 : Pr) and the double fluoride of potassium and yttrium doped with terbium (K 2 YF 5 : Tb) they show an alternative with many advantages, for their increase of the sensibility to the being doped with those rare earths in different percentages. In this investigation work was determined some dosimetric characteristics like they are the form of TL glow curve to be able to determine the TL property expressed by the kinetic parameters of the TL glow curve: Activation energy E t of the trapper states and the time of permanency in these metastates given by the preexponential factor s of the TL material: double fluoride of potassium and yttrium doped with terbium (K 2 YF 5 : Tb) to 0.8% and 0.99%. Also to this TL material it is characterized studying the TL response in function of the dose and the time of decay of the TL material. To determine the kinetic parameters of the TL glow curve of the TL materials these went one to one irradiated with a beta radiation source of 90 Sr/ 90 Y giving them a dose of 233 mGy for later to be heated in a lineal way during a certain time, with this the trapped electrons were liberated and the energy excess is emitted in light form. The registration of this light is known like it glow curves, those that were obtained with an analyzer TL Harshaw 4000, with interface to CPU for the handling of the data of the curves. The kinetic parameters of TL glow curve were determined for

  2. MHTGR inherent heat transfer capability

    International Nuclear Information System (INIS)

    Berkoe, J.M.

    1992-01-01

    This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

  3. Decay Process in an Active Medium. An Example

    International Nuclear Information System (INIS)

    Atamanuk, B.; Volokitin, A.S.

    1999-01-01

    In the background of many plasma phenomena are wave-wave interactions, wave-particle interactions, energy transfers, and so on. In natural and laboratory plasma, very often there are situations when a plasma is weakly unstable, that is not far above a threshold of instability. In these cases, we can describe a perturbed state in terms of a small finite number of plasma waves. In the present work, we will consider non-linear stabilization of the current instability in isothermal magnetized plasma by a three-wave decay process. This problem was studied in the many works, but some questions remain unclear, e.g. the existence and properties of chaotic regimes for certain parameters. We will consider an interesting, realistic plasma system with current instability in isothermal plasmas. To assume that only three waves are involved in the process of stabilization we have to consider instability in the conditions very close to the threshold, when only one low hybrid mode is unstable. This wave decays on two other strongly damped low hybrids waves. The dynamic stabilization of instability and swapping of energy from a radiant in area of damping (heat of plasma) thus happens. This example shows a common situation and that represents practical interest at the analysis of natural processes in ionosphere. (author)

  4. 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

  5. Modeling effects of DO and SRT on activated sludge decay and production.

    Science.gov (United States)

    Liu, Guoqiang; Wang, Jianmin

    2015-09-01

    The effect of dissolved oxygen (DO) on the endogenous decay of active heterotrophic biomass and the hydrolysis of cell debris were studied. With the inclusion of a hydrolysis process for the cell debris, mathematical models that are capable of quantifying the effects of DO and sludge retention time (SRT) on concentrations of active biomass and cell debris in activated sludge are presented. By modeling the biomass cultivated with unlimited DO, the values of endogenous decay coefficient for heterotrophic biomass, the hydrolysis constant of cell debris, and the fraction of decayed biomass that became cell debris were determined to be 0.38 d(-1), 0.013 d(-1), and 0.28, respectively. Results from modeling the biomass cultivated under different DO conditions suggested that the experimental low DO (∼0.2 mg/L) did not inhibit the endogenous decay of heterotrophic biomass, but significantly inhibited the hydrolysis of cell debris with a half-velocity constant value of 2.1 mg/L. Therefore, the increase in sludge production with low DO was mainly contributed by cell debris rather than the active heterotrophic biomass. Maximizing sludge production during aerobic wastewater treatment could reduce aeration energy consumption and improve biogas energy recovery potential. Copyright © 2015 Elsevier Ltd. All rights reserved.

  6. The effect of the decay data on activation cross section

    International Nuclear Information System (INIS)

    Huang Xiaolong

    2002-01-01

    The effect of the decay data on evaluation of activation cross section is investigated. Present work shows that these effects must be considered carefully when activation cross section is evaluated. Sometime they are main reason for causing the discrepancies among the experimental data

  7. Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

    International Nuclear Information System (INIS)

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

    2015-01-01

    Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

  8. Global, decaying solutions of a focusing energy-critical heat equation in R4

    Science.gov (United States)

    Gustafson, Stephen; Roxanas, Dimitrios

    2018-05-01

    We study solutions of the focusing energy-critical nonlinear heat equation ut = Δu - | u|2 u in R4. We show that solutions emanating from initial data with energy and H˙1-norm below those of the stationary solution W are global and decay to zero, via the "concentration-compactness plus rigidity" strategy of Kenig-Merle [33,34]. First, global such solutions are shown to dissipate to zero, using a refinement of the small data theory and the L2-dissipation relation. Finite-time blow-up is then ruled out using the backwards-uniqueness of Escauriaza-Seregin-Sverak [17,18] in an argument similar to that of Kenig-Koch [32] for the Navier-Stokes equations.

  9. Presentation of the decay data group activities at LPRI. (Links with international evaluation works)

    International Nuclear Information System (INIS)

    Be, M.M.

    1999-01-01

    The activities of the Primary Ionizing Radiation Laboratory (LPRI) which is affiliated with the French National Bureau of Metrology (BNM), include evaluation of decay data, absolute activity measurements, X- and gamma-ray spectrometry, development of new measuring techniques, etc. The following three activities in the field of decay data evaluation are briefly presented: NUCLEIDE software, internal conversion coefficients, and nuclide evaluations

  10. Approximation of the decay of fission and activation product mixtures

    International Nuclear Information System (INIS)

    Henderson, R.W.

    1991-01-01

    The decay of the exposure rate from a mixture of fission and activation products is a complex function of time. The exact solution of the problem involves the solution of more than 150 tenth order Bateman equations. An approximation of this function is required for the practical solution of problems involving multiple integrations of this function. Historically this has been a power function, or a series of power functions, of time. The approach selected here has been to approximate the decay with a sum of exponential functions. This produces a continuous, single valued function, that can be made to approximate the given decay scheme to any desired degree of closeness. Further, the integral of the sum is easily calculated over any period. 3 refs

  11. 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)

  12. 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)

  13. Heat-equilibrium low-temperature plasma decay in synthesis of ammonia via transient components N2H6

    International Nuclear Information System (INIS)

    Cao Guobin; Song Youqun; Chen Qing; Zhou Qiulan; Cao Yun; Wang Chunhe

    2001-01-01

    The author introduced a new method of heat-equilibrium low-temperature plasma in ammonia synthesis and a technique of continuous real-time inlet sampling mass-spectrometry to detect the reaction channel and step of the decay of transient component N 2 H 6 into ammonia. The experimental results indicated that in the process of ammonia synthesis by discharge of N 2 and H 2 mixture, the transient component N 2 H 6 is a necessary step

  14. One-Loop Operation of Primary Heat Transport System in MONJU During Heat Transport System Modifications

    International Nuclear Information System (INIS)

    Goto, T.; Tsushima, H.; Sakurai, N.; Jo, T.

    2006-01-01

    MONJU is a prototype fast breeder reactor (FBR). Modification work commenced in March 2005. Since June 2004, MONJU has changed to one-loop operation of the primary heat transport system (PHTS) with all of the secondary heat transport systems (SHTS) drained of sodium. The purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the decay heat by an acceptable margin but the capacity of pre-heaters to keep the sodium within the primary vessel at about 200 deg. C must be maintained. With regard to the heat loss and the decay heat, the estimated heat loss in the primary system was in the range of 90-170 kW in one-loop operation, and the calculated decay heat was 21.2 kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the decay heat. As for pre-heaters, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the PHTS was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the PHTS and the modification period was shortened. (authors)

  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. Gamma-ray constraints on hadronic and leptonic activities of decaying dark matter

    International Nuclear Information System (INIS)

    Chen, Chuan-Ren; Mandal, Sourav K.; Takahashi, Fuminobu

    2010-01-01

    While the excess in cosmic-ray electrons and positrons reported by PAMELA and Fermi may be explained by dark matter decaying primarily into charged leptons, this does not necessarily mean that dark matter should not have any hadronic decay modes. In order to quantify the allowed hadronic activities, we derive constraints on the decay rates of dark matter into WW, ZZ, hh, q q-bar and gg using the Fermi and HESS gamma-ray data. We also derive gamma-ray constraints on the leptonic e + e − , μ + μ − and τ + τ − final states. We find that dark matter must decay primarily into μ + μ − or τ + τ − in order to simultaneously explain the reported excess and meet all gamma-ray constraints

  17. Novel measurement method of heat and light detection for neutrinoless double beta decay

    Science.gov (United States)

    Kim, G. B.; Choi, J. H.; Jo, H. S.; Kang, C. S.; Kim, H. L.; Kim, I.; Kim, S. R.; Kim, Y. H.; Lee, C.; Lee, H. J.; Lee, M. K.; Li, J.; Oh, S. Y.; So, J. H.

    2017-05-01

    We developed a cryogenic phonon-scintillation detector to search for 0νββ decay of 100Mo. The detector module, a proto-type setup of the AMoRE experiment, has a scintillating 40Ca100MoO4 absorber composed of 100Mo-enriched and 48Ca-depleted elements. This new detection method employs metallic magnetic calorimeters (MMCs) as the sensor technology for simultaneous detection of heat and light signals. It is designed to have high energy and timing resolutions to increase sensitivity to probe the rare event. The detector, which is composed of a 200 g 40Ca100MoO4 crystal and phonon/photon sensors, showed an energy resolution of 8.7 keV FWHM at 2.6 MeV, with a weak temperature dependence in the range of 10-40 mK. Using rise-time and mean-time parameters and light/heat ratios, the proposed method showed a strong capability of rejecting alpha-induced events from electron events with as good as 20σ separation. Moreover, we discussed how the signal rise-time improves the rejection efficiency for random coincidence signals.

  18. Radioactive decay and labeled compounds

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    This chapter on radioactive decay and labeled compounds has numerous intext equations and worked, sample problems. Topics covered include the following: terms and mathematics of radioactive decay; examples of calculations; graphs of decay equations; radioactivity or activity; activity measurements; activity decay; half-life determinations; labeled compounds. A 20 problem set is also included. 1 ref., 4 figs., 1 tab

  19. 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

  20. Active transport and heat.

    Science.gov (United States)

    Tait, Peter W

    2011-07-01

    Increasing heat may impede peoples' ability to be active outdoors thus limiting active transport options. Co-benefits from mitigation of and adaptation to global warming should not be assumed but need to be actively designed into strategies.

  1. Comparison of decay and yield data between JNDC2 and ENDF/B-VI

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, K.; Sagosaka, M.; Miyazono, T. [Nagoya Univ. (Japan)

    1997-03-01

    This work is intended to be our first step to solve disagreements of the decay heat powers between measurements and summation calculations. We examine differences between nuclear data libraries to complement our uncertainty evaluation of the decay heat summation calculations only with ENDF/B-VI. The comparison is made mainly between JNDC2 and ENDF/B-VI while JEF2.2 decay data is also discussed. In this study, we propose and use a simple method which is an analogue of the overlap integral of two wave functions in quantum mechanics. As the first step, we compare the whole input nuclear data for the summation calculations as a whole. We find a slight difference of the fission yields especially for high-energy neutron induced fissions between JNDC2 and ENDF/B-VI. As for the decay energies, JNDC2, ENDF/B-VI are quite similar while JEF2.2 is found significantly different from these two libraries. We find substantial differences in the decay constant values among the three libraries. As the second step, we calculate the decay heat powers with FPGS90 using JNDC2 and ENDF/B-VI. The total decay heat powers with the two libraries differ by more than 10% at short cooling times while they agree well on the average at cooling times longer that 100 (s). We also point out nuclides whose contributions are significantly different between the two libraries even though the total decay heats agree well. These nuclides may cause some problems in predicting aggregate spectra of {beta} and {gamma} rays as well as delayed neutrons, and are to be reviewed in the future revision of decay and yield data. (author)

  2. Active heat exchange system development for latent heat thermal energy storage

    Science.gov (United States)

    Lefrois, R. T.; Mathur, A. K.

    1980-01-01

    Five tasks to select, design, fabricate, test and evaluate candidate active heat exchanger modules for future applications to solar and conventional utility power plants were discussed. Alternative mechanizations of active heat exchange concepts were analyzed for use with heat of fusion phase change materials (PCMs) in the temperature range of 250 to 350 C. Twenty-six heat exchange concepts were reviewed, and eight were selected for detailed assessment. Two candidates were selected for small-scale experimentation: a coated tube and shell heat exchanger and a direct contact reflux boiler. A dilute eutectic mixture of sodium nitrate and sodium hydroxide was selected as the PCM from over 50 candidate inorganic salt mixtures. Based on a salt screening process, eight major component salts were selected initially for further evaluation. The most attractive major components in the temperature range of 250 to 350 C appeared to be NaNO3, NaNO2, and NaOH. Sketches of the two active heat exchange concepts selected for test are given.

  3. Bubble chamber: Omega production and decay

    CERN Multimedia

    1973-01-01

    This image is of real particle tracks taken from the CERN 2 m liquid hydrogen bubble chamber and shows the production and decay of a negative omega particle. A negative kaon enters the chamber which decays into many particles, including a negative omega that travels a short distance before decaying into more particles. The invention of bubble chambers in 1952 revolutionized the field of particle physics, allowing real tracks left by particles to be seen and photographed by expanding liquid that had been heated to boiling point.

  4. Effects of proliferation on the decay of thermotolerance in Chinese hamster cells.

    Science.gov (United States)

    Armour, E P; Li, G C; Hahn, G M

    1985-09-01

    Development and decay of thermotolerance were observed in Chinese hamster HA-1 cells. The thermotolerance kinetics of exponentially growing and fed plateau-phase cells were compared. Following a 10-min heat exposure at 45 degrees C, cells in both growth states had similar rates of development of tolerance to a subsequent 45-min exposure at 45 degrees C. This thermotolerant state started to decay between 12 and 24 hr after the initial heat exposure. The decay appeared to initiate slightly sooner in the exponentially growing cells when compared to the fed plateau-phase cells. During the decay phase, the rate of thermotolerance decay was similar in the two growth conditions. In other experiments, cells were induced to divide at a slower rate by chronic growth (3 months) in a low concentration of fetal calf serum. Under these low serum conditions cells became more sensitive to heat and the rate of decay of thermotolerance remained the same for exponentially growing cells. Plateau-phase cells were also more sensitive, but thermotolerance decayed more rapidly in these cells. Although dramatic cell cycle perturbations were seen in the exponentially growing cells, these changes appeared not to be related to thermotolerance kinetics.

  5. 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

  6. 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

  7. 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

  8. Device-dependent activity estimation and decay correction of radionuclide mixtures with application to Tc-94m PET studies

    International Nuclear Information System (INIS)

    Smith, Mark F.; Daube-Witherspoon, Margaret E.; Plascjak, Paul S.; Szajek, Lawrence P.; Carson, Richard E.; Everett, James R.; Green, Shielah L.; Territo, Paul R.; Balaban, Robert S.; Bacharach, Stephen L.; Eckelman, William C.

    2001-01-01

    Multi-instrument activity estimation and decay correction techniques were developed for radionuclide mixtures, motivated by the desire for accurate quantitation of Tc-94m positron emission tomography (PET) studies. Tc-94m and byproduct Tc isotopes were produced by proton irradiation of enriched Mo-94 and natural Mo targets. Mixture activities at the end of bombardment were determined with a calibrated high purity germanium detector. The activity fractions of the greatest mixture impurities relative to 100% for Tc-94m averaged 10.0% (Tc-94g) and 3.3% (Tc-93) for enriched targets and 10.1% (Tc-94g), 11.0% (Tc-95), 255.8% (Tc-96m), and 7.2% (Tc-99m) for natural targets. These radioisotopes have different half-lives (e.g., 52.5 min for Tc-94m, 293 min for Tc-94g), positron branching ratios (e.g., 0.72 for Tc-94m, 0.11 for Tc-94g) and gamma ray emissions for themselves and their short-lived, excited Mo daughters. This complicates estimation of injected activity with a dose calibrator, in vivo activity with PET and blood sample activity with a gamma counter. Decay correction using only the Tc-94m half-life overestimates activity and is inadequate. For this reason analytic formulas for activity estimation and decay correction of radionuclide mixtures were developed. Isotope-dependent sensitivity factors for a PET scanner, dose calibrator, and gamma counter were determined using theoretical sensitivity models and fits of experimental decay curves to sums of exponentials with fixed decay rates. For up to 8 h after the end of bombardment with activity from enriched and natural Mo targets, decay-corrected activities were within 3% of the mean for three PET studies of a uniform cylinder, within 3% of the mean for six dose calibrator decay studies, and within 6% of the mean for four gamma counter decay studies. Activity estimation and decay correction for Tc-94m mixtures enable routine use of Tc-94m in quantitative PET, as illustrated by application to a canine Tc-94m sestamibi

  9. 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)

  10. Monopole heat

    International Nuclear Information System (INIS)

    Turner, M.S.

    1983-01-01

    Upper bounds on the flux of monopoles incident on the Earth with velocity -5 c(10 16 GeV m -1 ) and on the flux of monopoles incident on Jupiter with velocity -3 c(10 16 GeV m -1 ), are derived. Monopoles moving this slowly lose sufficient energy to be stopped, and then catalyse nucleon decay, releasing heat. The limits are obtained by requiring the rate of energy release from nucleon decay to be less than the measured amount of heat flowing out from the surface of the planet. (U.K.)

  11. 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)

  12. 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)

  13. Diffusion of heat from a finite, rectangular, plane heat source

    International Nuclear Information System (INIS)

    Ferreri, J.C.; Caballero, C.H.

    1985-01-01

    Non-dimensional results for the temperature field originating in a rectangular, finite, plane heat source with infinitesimal thickness are introduced. The source decays in time, zero decay being a particular case. Results are useful for obtaining an aproximation of the maximum temperature of a system holding an internal heat source. The range selected for the parameters is specially useful in the case of a nuclear waste repository. The application to the case of mass diffussion arises from analogy. (Author) [es

  14. 2.5 MWT Heat Exchanger Designs for Passive DHRS in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dehee; Eoh, Jaehyuk; Lee, Tae-Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Decay Heat Removal System (DHRS) of PGSFR consists of two passive DHRS (PDHRS) trains and two active DHRS (ADHRS) trains. Recently, total heat removal capacity of the DHRS in the PGSFR has increased to 10 MWT from 4 MWT reflecting safety analysis results. Consequently, DHRS components including heat exchangers, dampers, electro-magnetic pump, fan, piping, expansion tank and stack have been newly designed. In this work, physical models and correlations to design two main components of the PDHRS, decay heat exchanger (DHX) and natural-draft sodium-to-air heat exchanger (AHX), are introduced and designed data are presented. Physical models and correlations applied for heat exchangers in the PDHRS design were introduced and design works using the SHXSA and AHXSA codes has been completed for 2.5 MWT decay heat removal capability. DHX and AHX are designed utilizing SHXSA and AHXSA codes, respectively. Those design codes have capability of thermal sizing and performance analysis for the shell-and-tube type and counter-current flow heat exchanger unit. Since both SHXSA and AHXSA codes are similar, following description is focused on the SHXSA code. A single flow channel associated with an individual heat transfer tube is basically considered for thermal sizing and then the calculation results and design variables regarding heat transfer and pressure drop, etc. are extended to whole tubes. Various correlations of heat transfer and pressure loss for the shell- and tubeside flows were implemented in the computer codes. The analysis domain is discretized into several control volumes and heat transfer and pressure losses are calculated in each control volume.

  15. Utilising heat from nuclear waste for space heating

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A heating unit utilising the decay heat from irradiated material comprises a storage envelope for the material associated with a heat exchange system, means for producing a flow of air over the heat exchange system to extract heat from the material, an exhaust duct capable of discharging the heated air to the atmosphere, and means for selectively diverting at least some of the heated air to effect the required heating. With the flow of air over the heat exchange system taking place by a natural thermosyphon process the arrangement is self regulating and inherently reliable. (author)

  16. Thermal decay of Lennard-Jones clusters

    International Nuclear Information System (INIS)

    Garzon, I.L.; Avalos-Borja, M.

    1989-01-01

    The decay mechanisms of argon clusters have been studied using molecular dynamics simulations and Lennard-Jones potentials. Heating up processes were applied to Ar 13 up to temperatures in the melting region. In this range of temperatures large fluctuations in the mean kinetic energy of the system are present and a sequential evaporation is observed. The thermal decay of these aggregates occurs in a time scale of nanoseconds. (orig.)

  17. Bubble chamber: Omega production and decay

    CERN Document Server

    1973-01-01

    This image is taken from one of CERN's bubble chambers and shows the decay of a positive kaon in flight. The decay products of this kaon can be seen spiraling in the magnetic field of the chamber. The invention of bubble chambers in 1952 revolutionized the field of particle physics, allowing real tracks left by particles to be seen and photographed by expanding liquid that has been heated to boiling point.

  18. Droplet condensation in rapidly decaying pressure fields

    International Nuclear Information System (INIS)

    Peterson, P.F.; Bai, R.Y.; Schrock, V.E.; Hijikata, K.

    1992-01-01

    Certain promising schemes for cooling inertial confinement fusion reactors call for highly transient condensation in a rapidly decaying pressure field. After an initial period of condensation on a subcooled droplet, undesirable evaporation begins to occur. Recirculation within the droplet strongly impacts the character of this condensation-evaporation cycle, particularly when the recirculation time constant is of the order of the pressure decay time constant. Recirculation can augment the heat transfer, delay the onset of evaporation, and increase the maximum superheat inside the drop by as much as an order of magnitude. This numerical investigation identifies the most important parameters and physics characterizing transient, high heat flux droplet condensation. The results can be applied to conceptual designs of inertial confinement fusion reactors, where initial temperature differences on the order of 1,500 K decay to zero over time spans the order of tens of milliseconds

  19. Correlation of Coronal Plasma Properties and Solar Magnetic Field in a Decaying Active Region

    Science.gov (United States)

    Ko, Yuan-Kuen; Young, Peter R.; Muglach, Karin; Warren, Harry P.; Ugarte-Urra, Ignacio

    2016-01-01

    We present the analysis of a decaying active region observed by the EUV Imaging Spectrometer on Hinode during 2009 December 7-11. We investigated the temporal evolution of its structure exhibited by plasma at temperatures from 300,000 to 2.8 million degrees, and derived the electron density, differential emission measure, effective electron temperature, and elemental abundance ratios of Si/S and Fe/S (as a measure of the First Ionization Potential (FIP) Effect). We compared these coronal properties to the temporal evolution of the photospheric magnetic field strength obtained from the Solar and Heliospheric Observatory Michelson Doppler Imager magnetograms. We find that, while these coronal properties all decreased with time during this decay phase, the largest change was at plasma above 1.5 million degrees. The photospheric magnetic field strength also decreased with time but mainly for field strengths lower than about 70 Gauss. The effective electron temperature and the FIP bias seem to reach a basal state (at 1.5 x 10(exp 6) K and 1.5, respectively) into the quiet Sun when the mean photospheric magnetic field (excluding all areas correlated with each other and the correlation is the strongest in the high-temperature plasma. Such correlation properties should be considered in the quest for our understanding of how the corona is heated. The variations in the elemental abundance should especially be considered together with the electron temperature and density.

  20. Excitation of half-integer up-shifted decay channel and quasi-mode in plasma edge for high power electron Bernstein wave heating scenario

    Directory of Open Access Journals (Sweden)

    M. Ali Asgarian

    2018-04-01

    Full Text Available Electron Bernstein waves (EBW consist of promising tools in driving localized off-axis current needed for sustained operation as well as effective selective heating scenarios in advanced over dense fusion plasmas like spherical tori and stellarators by applying high power radio frequency waves within the range of Megawatts. Here some serious non-linear effects like parametric decay modes are highly expect-able which have been extensively studied theoretically and experimentally. In general, the decay of an EBW depends on the ratio of the incident frequency and electron cyclotron frequency. At ratios less than two, parametric decay leads to a lower hybrid wave (or an ion Bernstein wave and EBWs at a lower frequency. For ratios more than two, the daughter waves constitute either an electron cyclotron quasi-mode and another EBW or an ion wave and EBW. However, in contrast with these decay patterns, the excitation of an unusual up-shifted frequency decay channel for the ratio less than two is demonstrated in this study which is totally different as to its generation and persistence. It is shown that this mode varies from the conventional parametric decay channels which necessarily satisfy the matching conditions in frequency and wave-vector. Moreover, the excitation of some less-known local non-propagating quasi-modes (virtual modes through weak-turbulence theory and their contributions to energy leakage from conversion process leading the reduction in conversion efficiency is assessed.

  1. Excitation of half-integer up-shifted decay channel and quasi-mode in plasma edge for high power electron Bernstein wave heating scenario

    Science.gov (United States)

    Ali Asgarian, M.; Abbasi, M.

    2018-04-01

    Electron Bernstein waves (EBW) consist of promising tools in driving localized off-axis current needed for sustained operation as well as effective selective heating scenarios in advanced over dense fusion plasmas like spherical tori and stellarators by applying high power radio frequency waves within the range of Megawatts. Here some serious non-linear effects like parametric decay modes are highly expect-able which have been extensively studied theoretically and experimentally. In general, the decay of an EBW depends on the ratio of the incident frequency and electron cyclotron frequency. At ratios less than two, parametric decay leads to a lower hybrid wave (or an ion Bernstein wave) and EBWs at a lower frequency. For ratios more than two, the daughter waves constitute either an electron cyclotron quasi-mode and another EBW or an ion wave and EBW. However, in contrast with these decay patterns, the excitation of an unusual up-shifted frequency decay channel for the ratio less than two is demonstrated in this study which is totally different as to its generation and persistence. It is shown that this mode varies from the conventional parametric decay channels which necessarily satisfy the matching conditions in frequency and wave-vector. Moreover, the excitation of some less-known local non-propagating quasi-modes (virtual modes) through weak-turbulence theory and their contributions to energy leakage from conversion process leading the reduction in conversion efficiency is assessed.

  2. 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

  3. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  4. Comparison of the effect of gamma irradiation, heat-radiation combination, and sulphur dioxide generating pads on decay and quality of grapes

    International Nuclear Information System (INIS)

    Thomas, Paul; Brij Bhushan; Joshi, M.R.

    1995-01-01

    Effect of gamma irradiation, heat-radiation combination and in-package sulphur dioxide fumigation on fungal decay, and quality of seedless grape cultivars, Thompson, Sonaka and Tas-A-Ganesh was evaluated under different storage regimes. Irradiation at 2 kGy or a combination of hot water dip (50 degC, 5 min), plus irradiation (1 kGy) showed less spoilage due to Rhizopus spp. and Botrytis spp. in grapes packaged in tissue paper lined boxes and stored at 4 deg, 15 deg and 25 degC. Storage in polyethylene lined boxes increased the fungal rot. In-package sulphur dioxide generating pad was most effective for control of decay in polyethylene lined boxes stored at 10 deg and 20 degC, but caused berry bleaching. Irradiation at 2.5 or 3.5 kGy controlled decay at 10 degC, but not so effectively at 20 degC. Organoleptic quality, berry firmness, and soluble solids were not affected by irradiation, but decreases in titratable acids and ascorbic acid were recorded. Packaging in polyethylene lined boxes retained berry turgidity, while slight shrivelling occurred in tissue paper lined boxes. The results indicate that gamma irradiation has potential as an alternative to sulphur dioxide fumigation for decay control during shipping and storage. (author). 20 refs., 4 tabs

  5. Fusion decay power: Validation of FISPACT and FENDL/A-2.0

    International Nuclear Information System (INIS)

    Sublet, J.C.; Forrest, R.A.

    1999-01-01

    Integral experiments are a rich source of information with which a wide range of validation and comparison exercises can be made in the activation data field. Materials samples have been irradiated in a wide range of simulated D-T neutron fields at three European laboratories and at JAERI FNS. The later experiment is unique because decay heat rather than activity was measured. Some results from that experiment are reported here with some details of data corrections that have been made for EAF-99. (author)

  6. OPG's approach of crediting natural circulation in outage heat sinks

    International Nuclear Information System (INIS)

    Fung, K.K.; Mackinnon, J.C.

    2001-01-01

    A review of crediting natural circulation as a backup means of removing the reactor core decay heat during an outage in Ontario Power Generation's nuclear stations was completed in 2000. The objective was to define the configurations and conditions under which natural circulation can be confidently credited as an effective heat transport mechanism for use in shutdown heat sink management. The project was an interdisciplinary program, and involved analyses in the areas of heat transport system thermalhydaulics, fuel and fuel channel thermal and mechanical behaviour, radiation physics, and probabilistic risks. The assessment shows that it is economically acceptable to credit natural circulation as a backup means of removing the core decay heat whenever the no fuel failure criteria are met. The economic risks associated with such a potential use decrease with time after shutdown. The waiting times after shutdown when there would be various levels of risks of damaging the pressure tubes and fuel bundles were derived for use in planning maintenance activities so as to minimize the economic risks. (author)

  7. Scattering of lattice solitons and decay of heat-current correlation in the Fermi-Pasta-Ulam-α -β model

    Science.gov (United States)

    Jin, Tao; Yu, Jian; Zhang, Nan; Zhao, Hong

    2017-08-01

    As is well known, solitons can be excited in nonlinear lattice systems; however, whether, and if so, how, this kind of nonlinear excitation can affect the energy transport behavior is not yet fully understood. Here we study both the scattering dynamics of solitons and heat transport properties in the Fermi-Pasta-Ulam-α -β model with an asymmetric interparticle interaction. By varying the asymmetry degree of the interaction (characterized by α ), we find that (i) for each α there exists a momentum threshold for exciting solitons from which one may infer an α -dependent feature of probability of presentation of solitons at a finite-temperature equilibrium state and (ii) the scattering rate of solitons is sensitively dependent on α . Based on these findings, we conjecture that the scattering between solitons will cause the nonmonotonic α -dependent feature of heat conduction. Fortunately, such a conjecture is indeed verified by our detailed examination of the time decay behavior of the heat current correlation function, but it is only valid for an early time stage. Thus, this result may suggest that solitons can have only a relatively short survival time when exposed in a thermal environment, eventually affecting the heat transport in a short time.

  8. 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)

  9. 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

  10. 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

  11. 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

  12. Correlation of Coronal Plasma Properties and Solar Magnetic Field in a Decaying Active Region

    Science.gov (United States)

    Ko, Yuan-Kuen; Young, Peter R.; Muglach, Karin; Warren, Harry P.; Ugarte-Urra, Ignacio

    2016-01-01

    We present the analysis of a decaying active region observed by the EUV Imaging Spectrometer on Hinode during 2009 December 7-11. We investigated the temporal evolution of its structure exhibited by plasma at temperatures from 300,000 to 2.8 million degrees, and derived the electron density, differential emission measure, effective electron temperature, and elemental abundance ratios of Si/S and Fe/S (as a measure of the First Ionization Potential (FIP) Effect). We compared these coronal properties to the temporal evolution of the photospheric magnetic field strength obtained from the Solar and Heliospheric Observatory Michelson Doppler Imager magnetograms. We find that, while these coronal properties all decreased with time during this decay phase, the largest change was at plasma above 1.5 million degrees. The photospheric magnetic field strength also decreased with time but mainly for field strengths lower than about 70 Gauss. The effective electron temperature and the FIP bias seem to reach a basal state (at 1.5 x 10(exp 6) K and 1.5, respectively) into the quiet Sun when the mean photospheric magnetic field (excluding all areas <10 G) weakened to below 35 G, while the electron density continued to decrease with the weakening field. These physical properties are all positively correlated with each other and the correlation is the strongest in the high-temperature plasma. Such correlation properties should be considered in the quest for our understanding of how the corona is heated. The variations in the elemental abundance should especially be considered together with the electron temperature and density.

  13. 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.

  14. 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.

  15. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

  16. 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)

  17. 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)

  18. Experimental investigation on passive heat transfer by long closed two-phase thermosiphons

    Energy Technology Data Exchange (ETDEWEB)

    Grass, Claudia; Kulenovic, Rudi; Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2017-07-15

    The removal of decay heat from spent fuel pools is presently realized by active cooling systems. In case of a station black out, a passive heat removal based on closed two-phase thermosiphons can contribute to the power plant safety. In this paper, the basic laboratory setup for closed two-phase thermosiphons and first experimental results are presented. Depending on the driving temperature difference and the heat input, steady-state and pulsating operation of the thermosiphons are investigated.

  19. Experimental validation of the decay power calculation code and nuclear databases - FISPACT-97 and EAF-97 and FENDL/A-2.0

    International Nuclear Information System (INIS)

    Sublet, J.

    1998-01-01

    The calculation of activation inventories is a key input to virtually all aspects of the safety and environmental assessment of fusion power devices, such as ITER. For the licensing of such devices, regulatory authorities will require proof that the calculations of activation, and calculations to which activation quantities are inputs, are either correct or conservative. An important aspect of activation is decay heat power. In fusion power plants, decay power arises after shutdown from the energy released in the decay of the products of neutron activation, mainly from gamma and beta rays. Computation of the decay power is performed by sophisticated computer codes which solve the large number of coupled differential equations which govern the generation and decay chains for the many nuclides involved. They rely on a large volume of nuclear data, both neutron activation cross-sections and radioactive decay data. Validation of decay power code predictions by means of direct comparison with integral data measurements of sample structural materials under fusion-typical neutron spectra generates confidence in the decay power values calculated. It also permits an assessment of the adequacy of the methods and nuclear data and indicates any inaccuracy or omission that may have led to erroneous results. No experimental data on decay power existed for fusion reactor structural materials and irradiation conditions before a series of experiments were performed using the Fusion Neutron Source FNS facility at the Japan Atomic Energy Research Institute JAERI. Fusion relevant material samples were irradiated in a simulated D-T neutron field for times up to 7 hours and the decay power so generated measured for cooling times up to three months. Using the highly sensitive Whole Energy Absorption Spectrometer (WEAS) method, both β and γ rays decay energies were measured at selected cooling times as early as one minute after the irradiation ended. Coupled to the experiments, and at

  20. Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Bang, In Cheol

    2017-01-01

    Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

  1. I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology with SCALE6.1

    International Nuclear Information System (INIS)

    Matijevic, M.; Pevec, D.; Jecmenica, R.

    2016-01-01

    The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core with 121 fuel assembly (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel is selected to achieve a high thermal power which is extracted with PHEs placed in the annual region between the barrel and the vessel. The complex and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging one, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,gamma) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of

  2. Development of nuclear decay data library JDDL, and nuclear generation and decay calculation code COMRAD

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Ihara, Hitoshi; Katakura, Jun-ichi; Hara, Toshiharu.

    1986-08-01

    For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)

  3. Activated barrier crossing dynamics in the non-radiative decay of NADH and NADPH

    Energy Technology Data Exchange (ETDEWEB)

    Blacker, Thomas S., E-mail: t.blacker@ucl.ac.uk [Centre for Mathematics and Physics in the Life Sciences and Experimental Biology (CoMPLEX), University College London, London WC1E 6BT (United Kingdom); Department of Physics and Astronomy, University College London, London WC1E 6BT (United Kingdom); Research Department of Cell and Developmental Biology, University College London, London WC1E 6BT (United Kingdom); Marsh, Richard J., E-mail: richard.marsh@ucl.ac.uk [Department of Physics and Astronomy, University College London, London WC1E 6BT (United Kingdom); Duchen, Michael R., E-mail: m.duchen@ucl.ac.uk [Research Department of Cell and Developmental Biology, University College London, London WC1E 6BT (United Kingdom); Bain, Angus J., E-mail: a.bain@ucl.ac.uk [Centre for Mathematics and Physics in the Life Sciences and Experimental Biology (CoMPLEX), University College London, London WC1E 6BT (United Kingdom); Department of Physics and Astronomy, University College London, London WC1E 6BT (United Kingdom)

    2013-08-30

    Highlights: ► NADH and NADPH have a high rate of non-radiative excited state decay. ► Conformational relaxation is shown to be a significant non-radiative pathway. ► The Kramers equation describes the barrier crossing dynamics of the relaxation. ► Conformational restriction upon enzyme binding will alter NAD(P)H lifetimes. - Abstract: In live tissue, alterations in metabolism induce changes in the fluorescence decay of the biological coenzyme NAD(P)H, the mechanism of which is not well understood. In this work, the fluorescence and anisotropy decay dynamics of NADH and NADPH were investigated as a function of viscosity in a range of water–glycerol solutions. The viscosity dependence of the non-radiative decay is well described by Kramers and Kramers–Hubbard models of activated barrier crossing over a wide viscosity range. Our combined lifetime and anisotropy analysis indicates common mechanisms of non-radiative relaxation in the two emitting states (conformations) of both molecules. The low frequencies associated with barrier crossing suggest that non-radiative decay is mediated by small scale motion (e.g. puckering) of the nicotinamide ring. Variations in the fluorescence lifetimes of NADH and NADPH when bound to different enzymes may therefore be attributed to differing levels of conformational restriction upon binding.

  4. Radon and its decay product activities in the magmatic area and the adjacent volcano-sedimentary Intrasudetic Basin

    OpenAIRE

    Solecki, A. T.; Puchala, R.; Tchorz, D.

    2007-01-01

    In the magmatic area of Sudetes covering the Karkonosze granite and adjacent volcano-sedimentary Intrasudetic Basin a study of atmospheric radon activity was performed by means of SSNTD Kodak LR-115. The study was completed by gamma spectrometric survey of eU and eTh determined by gamma activity of radon decay products 214Bi and 208Tl respectively. In the case of the western part of the Karkonosze granite area the radon decay products activity in the granitic basement was ...

  5. General Purpose Heat Source Simulator

    Science.gov (United States)

    Emrich, Bill

    2008-01-01

    The General Purpose Heat Source (GPHS) simulator project is designed to replicate through the use of electrical heaters, the form, fit, and function of actual GPHS modules which generate heat through the radioactive decay of Pu238. The use of electrically heated modules rather than modules containing Pu238 facilitates the testing of spacecraft subsystems and systems without sacrificing the quantity and quality of the test data gathered. Previous GPHS activities are centered around developing robust heater designs with sizes and weights that closely matched those of actual Pu238 fueled GPHS blocks. These efforts were successful, although their maximum temperature capabilities were limited to around 850 C. New designs are being pursued which also replicate the sizes and weights of actual Pu238 fueled GPHS blocks but will allow operation up to 1100 C.

  6. Heat transfer performance test of PDHRS heat exchangers of PGSFR using STELLA-1 facility

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jonggan, E-mail: hong@kaeri.re.kr; Yeom, Sujin; Eoh, Jae-Hyuk; Lee, Tae-Ho; Jeong, Ji-Young

    2017-03-15

    Highlights: • Heat transfer performance test of heat exchangers of PGSFR PDHRS is conducted using STELLA-1 facility. • Steady-state test results of DHX and AHX show good agreement with theoretical results of design codes. • Design codes for DHX and AHX are validated by STELLA-1 experimental results. • Heat transport capability of DHX and AHX is turned out to be satisfactory for reliable plant operation. - Abstract: The STELLA-1 facility was designed and constructed to carry out separate effect tests of the decay heat exchanger (DHX) and natural draft sodium-to-air heat exchanger (AHX), which are key components of the safety-grade decay heat removal system in PGSFR. The DHX is a sodium-to-sodium heat exchanger with a straight tube arrangement, and the AHX is a sodium-to-air heat exchanger with a helically coiled tube arrangement. The model heat exchangers in STELLA-1 have been designed to meet their own similitude conditions from the prototype ones, of which scale ratios were set to be unity in height (or length) and 1/2.5 in heat transfer rate. Consequently, the overall heat transfer coefficients and log-mean temperature differences of the prototypes have been preserved as well. The steady-state test results for each model heat exchanger obtained from STELLA-1 showed good agreement with the theoretical results of the computer design codes for thermal-sizing and a performance analysis of the DHX and AHX. In the DHX result comparison, the discrepancies in the heat transfer rate ranged from −4.4% to 2.0%, and in the AHX result comparison, they ranged from −11.1% to 12.6%. Therefore, the first step in thermal design codes validation for sodium heat exchangers, e.g., DHX and AHX, has been successfully completed with the experimental database obtained from STELLA-1. In addition, the heat transfer performance of the DHX and AHX was found to be satisfactory enough to secure a reliable decay heat removal performance.

  7. Few active mechanisms of the neutrinoless double beta-decay and effective mass of Majorana neutrinos

    CERN Document Server

    Simkovic, Fedor; Faessler, Amand

    2010-01-01

    It is well known that there exist many mechanisms that may contribute to neutrinoless double beta decay (0nbb-decay). By exploiting the fact that the associated nuclear matrix elements are target dependent we show that, given definite experimental results on a sufficient number of targets, one can determine or sufficiently constrain all lepton violating parameters including the mass term. As a specific example we show that, assuming the observation of the 0nbb-decay in three different nuclei, e.g., 76Ge, 100Mo and 130Te, and just three lepton number violating mechanisms (light and heavy neutrino mass mechanisms as well as R-parity breaking SUSY mechanism) being active, there are only four different solutions for the lepton violating parameters, provided that they are relatively real. In particular, assuming evidence of the 0nbb-decay of 76Ge, the effective neutrino Majorana mass |m_bb| can be almost uniquely extracted by utilizing other existing constraints (cosmological observations and tritium beta-decay ex...

  8. Decay property of Timoshenko system in thermoelasticity

    KAUST Repository

    Said-Houari, Belkacem

    2011-12-30

    We investigate the decay property of a Timoshenko system of thermoelasticity in the whole space for both Fourier and Cattaneo laws of heat conduction. We point out that although the paradox of infinite propagation speed inherent in the Fourier law is removed by changing to the Cattaneo law, the latter always leads to a solution with the decay property of the regularity-loss type. The main tool used to prove our results is the energy method in the Fourier space together with some integral estimates. We derive L 2 decay estimates of solutions and observe that for the Fourier law the decay structure of solutions is of the regularity-loss type if the wave speeds of the first and the second equations in the system are different. For the Cattaneo law, decay property of the regularity-loss type occurs no matter what the wave speeds are. In addition, by restricting the initial data to U 0∈H s(R)∩L 1,γ(R) with a suitably large s and γ ∈ [0,1], we can derive faster decay estimates with the decay rate improvement by a factor of t -γ/2. © 2011 John Wiley & Sons, Ltd.

  9. Numerical scalings of the decay lengths in the scrape-off layer

    DEFF Research Database (Denmark)

    Militello, F.; Naulin, V; Nielsen, Anders Henry

    2013-01-01

    Numerical simulations of L-mode turbulence in the scrape-off layer (SOL) are used to construct power scaling laws for the characteristic decay lengths of the temperature, density and heat flux at the outer mid-plane. Most of the results obtained are in qualitative agreement with the experimental...... observations despite the known limitation of the model. Quantitative agreement is also obtained for some exponents. In particular, an almost linear inverse dependence of the heat flux decay length with the plasma current is recovered. The relative simplicity of the theoretical model used allows one to gain...

  10. Temperature-gated thermal rectifier for active heat flow control.

    Science.gov (United States)

    Zhu, Jia; Hippalgaonkar, Kedar; Shen, Sheng; Wang, Kevin; Abate, Yohannes; Lee, Sangwook; Wu, Junqiao; Yin, Xiaobo; Majumdar, Arun; Zhang, Xiang

    2014-08-13

    Active heat flow control is essential for broad applications of heating, cooling, and energy conversion. Like electronic devices developed for the control of electric power, it is very desirable to develop advanced all-thermal solid-state devices that actively control heat flow without consuming other forms of energy. Here we demonstrate temperature-gated thermal rectification using vanadium dioxide beams in which the environmental temperature actively modulates asymmetric heat flow. In this three terminal device, there are two switchable states, which can be regulated by global heating. In the "Rectifier" state, we observe up to 28% thermal rectification. In the "Resistor" state, the thermal rectification is significantly suppressed (Rectifier state. This temperature-gated rectifier can have substantial implications ranging from autonomous thermal management of heating and cooling systems to efficient thermal energy conversion and storage.

  11. Study of an active wall solar heating system

    International Nuclear Information System (INIS)

    Kassem, Talal

    2006-01-01

    An active wall solar heating system was built and tested. In the same time a compatible computer program has been according to set the recommended dimensions for the solar collectors where (F-Chart) method used to set the ratio of monthly solar sharing average for the examined heating system. Some parameters, such as collectors' areas, its tilt angle and near earth reflecting were experimentally investigated, affecting the executed active solar heating system performance. The study explain the ability of using this system which is simple, Low coast and high performance in heating residential and public buildings and heating water with ratio of yearly solar sharing achieves the needed saving of using this system.(Author)

  12. Blanket activation and afterheat for the Compact Reversed-Field Pinch Reactor

    International Nuclear Information System (INIS)

    Davidson, J.W.; Battat, M.E.

    1985-01-01

    A detailed assessment has been made of the activation and afterheat for a Compact Reversed-Field Pinch Reactor (CRFPR) blanket using a two-dimensional model that included the limiter, the vacuum ducts, and the manifolds and headers for cooling the limiter and the first and second walls. Region-averaged, multigroup fluxes and prompt gamma-ray/neutron heating rates were calculated using the two-dimensional, discrete-ordinates code TRISM. Activation and depletion calculations were performed with the code FORIG using one-group cross sections generated with the TRISM region-averaged fluxes. Afterheat calculations were performed for regions near the plasma, i.e., the limiter, first wall, etc. assuming a 10-day irradiation. Decay heats were computed for decay periods up to 100 minutes. For the activation calculations, the irradiation period was taken to be one year and blanket activity inventories were computed for decay times to 4 x 10 5 years. These activities were also calculated as the toxicity-weighted biological hazard potential (BHP). 15 refs

  13. 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

  14. Active heat exchange system development for latent heat thermal energy storage

    Science.gov (United States)

    Alario, J.; Kosson, R.; Haslett, R.

    1980-01-01

    Various active heat exchange concepts were identified from among three generic categories: scrapers, agitators/vibrators and slurries. The more practical ones were given a more detailed technical evaluation and an economic comparison with a passive tube-shell design for a reference application (300 MW sub t storage for 6 hours). Two concepts were selected for hardware development: (1) a direct contact heat exchanger in which molten salt droplets are injected into a cooler counterflowing stream of liquid metal carrier fluid, and (2) a rotating drum scraper in which molten salt is sprayed onto the circumference of a rotating drum, which contains the fluid salt is sprayed onto the circumference of a rotating drum, which contains the fluid heat sink in an internal annulus near the surface. A fixed scraper blade removes the solidified salt from the surface which was nickel plated to decrease adhesion forces. In addition to improving performance by providing a nearly constant transfer rate during discharge, these active heat exchanger concepts were estimated to cost at least 25% less than the passive tube-shell design.

  15. 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

  16. 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

  17. Experimental evaluation of sodium to air heat exchanger performance

    International Nuclear Information System (INIS)

    Vinod, V.; Pathak, S.P.; Paunikar, V.D.; Suresh Kumar, V.A.; Noushad, I.B.; Rajan, K.K.

    2013-01-01

    Highlights: ► Sodium to air heat exchangers are used to remove the decay heat produced in fast breeder reactor after shutdown. ► Finned tube sodium to air heat exchanger with sodium on tube side was tested for its heat transfer performance. ► A one dimensional computer code was validated by the experimental data obtained. ► Non uniform sodium and air flow distribution was present in the heat exchanger. - Abstract: Sodium to air heat exchangers (AHXs) is used in Prototype Fast Breeder Reactor (PFBR) circuits to reject the decay heat produced by the radioactive decay of the fission products after reactor shutdown, to the atmospheric air. The heat removal through sodium to air heat exchanger maintains the temperature of reactor components in the pool within safe limits in case of non availability of normal heat transport path. The performance of sodium to air heat exchanger is very critical to ensure high reliability of the decay heat removal systems in sodium cooled fast breeder reactors. Hence experimental evaluation of the adequacy of the heat transfer capability gives confidence to the designers. A finned tube cross flow sodium to air heat exchanger of 2 MW heat transfer capacity with sodium on tube side and air on shell side was tested in the Steam Generator Test Facility at Indira Gandhi Center for Atomic Research, India. Heat transfer experiments were carried out with forced circulation of sodium and air, which confirmed the adequacy of heat removal capacity of the heat exchanger. The testing showed that 2.34 MW of heat power is transferred from sodium to air at nominal flow and temperature conditions. A one dimensional computer code developed for design and analysis of the sodium to air heat exchanger was validated by the experimental data obtained. An equivalent Nusselt number, Nu eq is derived by approximating that the resistance of heat transfer from sodium to air is contributed only by the film resistance of air. The variation of Nu eq with respect

  18. Decay property of Timoshenko system in thermoelasticity

    KAUST Repository

    Said-Houari, Belkacem; Kasimov, Aslan R.

    2011-01-01

    We investigate the decay property of a Timoshenko system of thermoelasticity in the whole space for both Fourier and Cattaneo laws of heat conduction. We point out that although the paradox of infinite propagation speed inherent in the Fourier law

  19. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  20. Estimation of heat transfer and heat source in a molten pool

    Energy Technology Data Exchange (ETDEWEB)

    Yun, J.I.; Suh, K.Y.; Kang, C.S. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

    2001-07-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

  1. Estimation of heat transfer and heat source in a molten pool

    International Nuclear Information System (INIS)

    Yun, J.I.; Suh, K.Y.; Kang, C.S.

    2001-01-01

    Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

  2. Research programme 'Active Solar Energy Use - Solar Heating and Heat Storage'. Activities and projects 2003

    International Nuclear Information System (INIS)

    Hadorn, J.-C.; Renaud, P.

    2003-01-01

    In this report by the research, development and demonstration (RD+D) programme coordinators the objectives, activities and main results in the area of solar heating and heat storage in Switzerland are presented for 2003. In a stagnating market environment the strategy of the Swiss Federal Office of Energy mainly consists in improving the quality and durability of solar collectors and materials, optimizing combisystems for space heating and domestic hot water preparation, searching for storage systems with a higher energy storage density than in the case of sensible heat storage in water, developing coloured solar collectors for more architectonic freedom, and finalizing a seasonal heat storage project for 100 dwellings to demonstrate the feasibility of solar fractions larger than 50% in apartment houses. Support was granted to the Swiss Testing Facility SPF in Rapperswil as in previous years; SPF was the first European testing institute to perform solar collector labeling according to the new rules of the 'Solar Keymark', introduced in cooperation with the European Committee for Standardization CEN. Several 2003 projects were conducted within the framework of the Solar Heating and Cooling Programme of the International Energy Agency IEA. Computerized simulation tools were improved. With the aim of jointly producing high-temperature heat and electric power a solar installation including a concentrating collector and a thermodynamic machine based on a Rankine cycle is still being developed. Seasonal underground heat storage was studied in detail by means of a validated computer simulation programme. Design guidelines were obtained for such a storage used in the summer time for cooling and in the winter time for space heating via a heat pump: depending on the ratio 'summer cooling / winter heating', cooling requires a cooling machine, or direct cooling without such a machine is possible. The report ends up with the list of all supported RD+D projects

  3. Selective activation of human heat shock gene transcription by nitrosourea antitumor drugs mediated by isocyanate-induced damage and activation of heat shock transcription factor.

    Science.gov (United States)

    Kroes, R A; Abravaya, K; Seidenfeld, J; Morimoto, R I

    1991-01-01

    Treatment of cultured human tumor cells with the chloroethylnitrosourea antitumor drug 1,3-bis(2-chloroethyl)-1-nitrosourea (BCNU) selectively induces transcription and protein synthesis of a subset of the human heat shock or stress-induced genes (HSP90 and HSP70) with little effect on other stress genes or on expression of the c-fos, c-myc, or beta-actin genes. The active component of BCNU and related compounds appears to be the isocyanate moiety that causes carbamoylation of proteins and nucleic acids. Transcriptional activation of the human HSP70 gene by BCNU is dependent on the heat shock element and correlates with the level of heat shock transcription factor and its binding to the heat shock element in vivo. Unlike activation by heat or heavy metals, BCNU-mediated activation is strongly dependent upon new protein synthesis. This suggests that BCNU-induced, isocyanate-mediated damage to newly synthesized protein(s) may be responsible for activation of the heat shock transcription factor and increased transcription of the HSP90 and HSP70 genes. Images PMID:2052560

  4. Thyroid activity in heat adaptation

    International Nuclear Information System (INIS)

    Joshi, B.C.; Varshney, V.P.; Sanwal, P.C.

    1980-01-01

    The effect of acute and chronic (22 day round-the-clock) exposure to microenvironmental heat stress (37 deg C DBT) on thyroid activity was studied in Hariana x Holstein Frisian, Hariana x Brown Swiss and Hariana x Jersey non-cycling F 1 crossbred heifers. Vis-a-vis their no-heat norms, the percentage uptake of tri-iodothyronine- 125 I by resin registered a steep fall (about 45 to 60 percent) on acute heat exposures reaching a minimum value in about 2 hrs. The levels started recouping by the 2nd day, plateuing out on the 5th day onwards at slightly subnormal level with up and down fluctuations throughout the three week duration of exposure to heat. There were no significant differences in the pattern or magnitude of response amongst breeds, though in case of Holstein Frisian and Brown Swiss cross-breds the levels of T 3 tended, at times, to overshoot the no-stress norm. (author)

  5. 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.

  6. Diagnostics of Coronal Heating in Solar Active Regions

    Science.gov (United States)

    Fludra, Andrzej; Hornsey, Christopher; Nakariakov, Valery

    2015-04-01

    We aim to develop a diagnostic method for the coronal heating mechanism in active region loops. Observational constraints on coronal heating models have been sought using measurements in the X-ray and EUV wavelengths. Statistical analysis, using EUV emission from many active regions, was done by Fludra and Ireland (2008) who studied power-law relationships between active region integrated magnetic flux and emission line intensities. A subsequent study by Fludra and Warren (2010) for the first time compared fully resolved images in an EUV spectral line of OV 63.0 nm with the photospheric magnetic field, leading to the identification of a dominant, ubiquitous variable component of the transition region EUV emission and a discovery of a steady basal heating, and deriving the dependence of the basal heating rate on the photospheric magnetic flux density. In this study, we compare models of single coronal loops with EUV observations. We assess to what degree observations of individual coronal loops made in the EUV range are capable of providing constraints on the heating mechanism. We model the coronal magnetic field in an active region using an NLFF extrapolation code applied to a photospheric vector magnetogram from SDO/HMI and select several loops that match an SDO/AIA 171 image of the same active region. We then model the plasma in these loops using a 1D hydrostatic code capable of applying an arbitrary heating rate as a function of magnetic field strength along the loop. From the plasma parameters derived from this model, we calculate the EUV emission along the loop in AIA 171 and 335 bands, and in pure spectral lines of Fe IX 17.1 nm and Fe XVI 33.5 nm. We use different spatial distributions of the heating function: concentrated near the loop top, uniform and concentrated near the footpoints, and investigate their effect on the modelled EUV intensities. We find a diagnostics based on the dependence of the total loop intensity on the shape of the heating function

  7. Thermal decay in underfloor air distribution (UFAD) systems: Fundamentals and influence on system performance

    International Nuclear Information System (INIS)

    Lee, Kwang Ho; Schiavon, Stefano; Bauman, Fred; Webster, Tom

    2012-01-01

    Graphical abstract: Surface heat transfer breakdown for an underfloor air distribution (UFAD) system supply plenum. Highlights: ► Thermal decay of a UFAD system is considerable (annual median = 3.7 K). ► Thermal decay is driven by heat transfer through both the concrete slab and the raised floor. ► Thermal decay may lead to higher airflow rates and increased fan and chiller energy consumption. -- Abstract: Underfloor air distribution (UFAD) is a mechanical ventilation strategy in which the conditioned air is primarily delivered to the zone from a pressurized plenum through floor mounted diffusers. Compared to conventional overhead (OH) mixing systems, UFAD has several potential advantages, such as improved thermal comfort and indoor air quality (IAQ), layout flexibility, reduced life cycle costs and improved energy efficiency in suitable climates. In ducted OH systems designers have reasonably accurate control of the diffuser supply temperature, while in UFAD this temperature is difficult to predict due to the heat gain of the conditioned air in the supply plenum. The increase in temperature between the air entering the plenum and air leaving through a diffuser is known as thermal decay. In this study, the detailed whole-building energy simulation program, EnergyPlus, was used to explain the fundamentals of thermal decay, to investigate its influence on energy consumption and to study the parameters that affect thermal decay. It turns out that the temperature rise is considerable (annual median = 3.7 K, with 50% of the values between 2.4 and 4.7 K based on annual simulations). Compared to an idealized simulated UFAD case with no thermal decay, elevated diffuser air temperatures can lead to higher supply airflow rate and increased fan and chiller energy consumption. The thermal decay in summer is higher than in winter and it also depends on the climate. The ground floor with a slab on grade has less temperature rise compared to middle and top floors. An

  8. 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)

  9. The field measurements of the activity-weighted size distributions of radon decay products

    International Nuclear Information System (INIS)

    Wasiolek, P.; Montassier, N.; Hopke, P.K.

    1991-01-01

    Because of the importance of particle size in the calculations of dose deposited in the respiratory tract by radon progeny, the determination of the size distribution of radon decay products in indoor air had increased in interest in recent years. A system for the measurement of the activity-weighted size distribution of radon decay products at environmental levels has been developed. The system (ASC-GSA) utilizes a combination of six multiple wire screens (Graded Screen Array) sampler detector units operated in parallel. The cut off points of the samplers and the data deconvolution procedure allow to obtain the activity fractions of radon progeny in the size range 0.5-500 mn. The computer control of sampling, alpha counting and data storage permits the operation of the system on the semi-continuous basis. The primary application of the ASC-GSA system has been to collect data of activity-weighted size distributions of radon progeny in real house environments. The results of field measurements in several houses with elevated radon levels are presented

  10. Activation of TZM and stainless steel divertor materials in the NET fusion machine

    International Nuclear Information System (INIS)

    Cepraga, D.G.; Menapace, E.; Cambi, G.; Ciattaglia, S.; Petrizzi, L.; Cavallone, G.; Costa, M.; Broccoli, U.

    1994-01-01

    This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code. ((orig.))

  11. Activation of TZM and stainless steel divertor materials in the NET fusion machine

    Energy Technology Data Exchange (ETDEWEB)

    Cepraga, D G [ENEA, INN-FIS, 8 Viale Ercolani, 40138, Bologna (Italy); Menapace, E [ENEA, INN-FIS, 8 Viale Ercolani, 40138, Bologna (Italy); Cambi, G [Bologna University, Physics Department, 33 Via Irnerio, 40126, Bologna (Italy); Ciattaglia, S [ENEA, NUC-FUS, 27 Via E. Fermi, 00044, Frascati (Italy); Petrizzi, L [ENEA, NUC-FUS, 27 Via E. Fermi, 00044, Frascati (Italy); Cavallone, G [NIER S.r.l., 16 Via S. Stefano, 40125, Bologna (Italy); Costa, M [NIER S.r.l., 16 Via S. Stefano, 40125, Bologna (Italy); Broccoli, U [ENEA, NUC-RIN, 4 Via Martiri del Sole, 40100, Bologna (Italy)

    1994-09-01

    This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code. ((orig.))

  12. On the development of a grid-enhanced single-phase convective heat transfer correlation

    International Nuclear Information System (INIS)

    Miller, D.J.; Cheung, F.B.; Bajorek, S.M.

    2011-01-01

    A new single-phase convective heat transfer augmentation correlation has been developed using single phase steam cooling experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. Experimental data obtained from the RBHT single phase steam cooling tests have been evaluated and new findings identified. Previous rod bundle tests showed the importance of spacer grid on the local heat transfer, and that the augmentation in heat transfer downstream of a grid decays exponentially. The RBHT data also shows that the Reynolds number affects the rate at which this augmentation decays. The new correlation includes the strong dependence of heat transfer on both the Reynolds number and the grid blockage ratio. While the effects of both parameters were clearly evident in the RBHT experimental data, existing correlations do not account for the Reynolds number effect. The developed correlation incorporates Reynolds number in the decay curve of heat transfer. The newly developed correlation adequately accounts for the dependence of the heat transfer augmentation decay rate on the local flow Reynolds number. (author)

  13. Conceptual design of an active magnetic regenerative heat circulator based on self-heat recuperation technology

    International Nuclear Information System (INIS)

    Kotani, Yui; Kansha, Yasuki; Tsutsumi, Atsushi

    2013-01-01

    A conceptual design of an active magnetic regenerative (AMR) heat circulator for self-heat recuperation to realize energy savings in thermal processes is proposed. The process fluid heat is recuperated by the magnetocaloric effect of ferromagnetic material through the AMR heat circulation cycle. In an AMR heat circulator, all the process fluid heat is circulated and no make-up heat is added to raise the process fluid to its set temperature. A one-dimensional mathematical model of the AMR heat circulator was constructed to understand its behavior and verify its energy-saving potential. From the constructed one-dimensional mathematical model, it is seen that AMR heat circulator has potential to drastically reduce the total energy consumption in a thermal process. The temperature–entropy diagram shows that in order to gain the maximum energy saving, optimization of the parameters such as the flow rate and geometry of the ferromagnetic working material beds is needed. - Highlights: • Self-heat recuperative active magnetic regenerative heat circulator is introduced. • One-dimensional model is constructed to verify its energy-saving potential. • Total energy consumption in thermal process is drastically reduced. • Further energy can be saved by reducing the overlapping of thermodynamic cycles

  14. 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)

  15. Nuclear decay data measurements at the INEL ISOL facility

    International Nuclear Information System (INIS)

    Greenwood, R.C.; Helmer, R.G.; Putnam, M.H.; Struttmann, D.A.; Watts, K.D.

    1991-01-01

    In recent years, the use of the mass separation technique coupled on-line to a source of fission product nuclides has provided a wealth of new information on the nuclear decay properties of such nuclides. In addition to their relevance in basic studies of nuclear properties of neutron-rich nuclei, the fission product nuclides as a group, because of their intimate link with energy production in fission reactors, occupy a unique position in the field of applied nuclear decay data. Further, in addition to their critical role in nuclear reactor technology (decay heat source term, environmental concerns, etc.), such data have important applications in astrophysical calculations involving the rapid neutron capture process (r-process) of elemental synthesis in stellar environments. The scope of the nuclear decay data measurements being undertaken using the Idaho National Engineering Laboratory's (INEL) isotope separation on-line (ISOL) facility is focused on a systematic study of the gross nuclear decay properties of short-lived fission product isotopes, i.e., ground-state half-lives, beta-decay energies and beta-decay feeding (or beta-strength) distributions. In this paper, the authors discuss the results of new measurements of beta-decay energies and feeding distributions

  16. ELECTRON ENERGY DECAY IN HELIUM AFTERGLOW PLASMAS AT CRYOGENIC TEMPERATURES

    Energy Technology Data Exchange (ETDEWEB)

    Goldan, P. D.; Cahn, J. H.; Goldstein, L.

    1963-10-15

    Studies of decaying afterglow plasmas in helium were ined near 4 deg K by immersion in a liquid helium bath. By means of a Maser Radiometer System, the electron temperature was followed below 200 deg K. Guided microwave propagation and wave interaction techniques premit determination of election number density and collision frequencies for momentum transfer. Electron temperature decay rates of the order of 150 mu sec/p(mm Hg alpha 4.2 deg K) were found. Since thermal relaxation by elastic collisions should be some two orders of magnitude faster than this, the electrons appear to be in quasiequilibrium with a slowly decaying internal heating source. Correlation of the expected decay rates of singlet metastable helium atoms with the electron temperature decay gives good agreement with the present experiment. (auth)

  17. The γ activity from 11Li beta decay

    International Nuclear Information System (INIS)

    Detraz, C.; Guillemaud, D.; Langevin, M.; Naulin, F.; Epherre, M.; Klapisch, R.; Saint-Simon, M. de; Thibault, C.; Touchard, F.

    1980-01-01

    The energies and absolute intensities of the γ-rays from the β-decay of 11 Li are measured. There is no sizable β branch to the 11 Be ground state. Only (5.2+-1.4)% of the β-decay strength does not lead to β-delayed particle emission. New β-delayed neutron branches to excited states of 10 Be are observed and the total delayed neutron emission probability is deduced

  18. NEANDC specialists meeting on yields and decay data of fission product nuclides

    International Nuclear Information System (INIS)

    Chrien, R.E.; Burrows, T.W.

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information

  19. a Search for Nucleon Decay with Multiple Muon Decays

    Science.gov (United States)

    Phillips, Thomas James

    A search was made for nucleon decays which result in multiple delayed muon decays using the HPW (Harvard -Purdue-Wisconsin) water Cerenkov detector. The HPW detector consists of 680 metric tons of purified water instrumented with 704 five-inch photomultiplier tubes. The phototubes are situated on a volume array with a lattice spacing of approximately one meter, and the inside walls of the detector are lined with mirrors. This combination of mirrors and a volume array of phototubes gives the HPW detector a low trigger energy threshold and a high muon decay detection efficiency. The detector is surrounded by wire chambers to provide an active shield, and is located at a depth of 1500 meters-of-water-equivalent in the Silver King Mine in Park City, Utah. The entire HPW data set, consisting of 17.2 million events collec- ted during 282 live days between May 1983 and October 1984, was analyzed. No contained events with multiple muon decays were found in a 180 ton fiducial volume. This is consistent with the background rate from neutrino interactions, which is expected to be 0.7 (+OR-) 0.2 events. The calculated lower lifetime limit for the decay mode p (--->) (mu)('+)(mu)('+)(mu)('-) is: (tau)/B.R. = 1 x 10('31) years (90% C.L.). Limits are calculated for ten other proton decay modes and five bound neutron decay modes, most of which are around 4 x 10('30) years (90% C.L.). No previous studies have reported results from direct searches for eight of these modes.

  20. 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)

  1. 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

  2. Canonical Poly(A Polymerase Activity Promotes the Decay of a Wide Variety of Mammalian Nuclear RNAs.

    Directory of Open Access Journals (Sweden)

    Stefan M Bresson

    2015-10-01

    Full Text Available The human nuclear poly(A-binding protein PABPN1 has been implicated in the decay of nuclear noncoding RNAs (ncRNAs. In addition, PABPN1 promotes hyperadenylation by stimulating poly(A-polymerases (PAPα/γ, but this activity has not previously been linked to the decay of endogenous transcripts. Moreover, the mechanisms underlying target specificity have remained elusive. Here, we inactivated PAP-dependent hyperadenylation in cells by two independent mechanisms and used an RNA-seq approach to identify endogenous targets. We observed the upregulation of various ncRNAs, including snoRNA host genes, primary miRNA transcripts, and promoter upstream antisense RNAs, confirming that hyperadenylation is broadly required for the degradation of PABPN1-targets. In addition, we found that mRNAs with retained introns are susceptible to PABPN1 and PAPα/γ-mediated decay (PPD. Transcripts are targeted for degradation due to inefficient export, which is a consequence of reduced intron number or incomplete splicing. Additional investigation showed that a genetically-encoded poly(A tail is sufficient to drive decay, suggesting that degradation occurs independently of the canonical cleavage and polyadenylation reaction. Surprisingly, treatment with transcription inhibitors uncouples polyadenylation from decay, leading to runaway hyperadenylation of nuclear decay targets. We conclude that PPD is an important mammalian nuclear RNA decay pathway for the removal of poorly spliced and nuclear-retained transcripts.

  3. The non-resonant decay of the fast magnetosonic wave during ICRH of a tokamak plasma

    International Nuclear Information System (INIS)

    Avinash, K.; Core, W.G.; Hellsten, T.; Farrell, C.M.

    1988-01-01

    The non-resonant decay of the fast magnetosonic wave into an Ion Bernstein wave and a quasi-mode is investigated from the point of view of assessing the importance of this process for the observed direct heating of the edge plasma during ion-cyclotron resonance heating (ICRH). Starting from the Maxwell-Vlasov equations, expressions for the threshold electric field and the growth rates of the decay process are obtained. For JET like parameters, the thresholds for the decay are easily exceeded and the growth time for typical fast wave electric field strengths is of the order of a microsecond. The parametric dependence of the threshold on magnetic field, temperature, the density of the various ion species, and electron-ion collisions is studied. Finally the relevance of this process to the heating of plasma edge during ICRH is discussed. (author)

  4. NEANDC specialists meeting on yields and decay data of fission product nuclides

    Energy Technology Data Exchange (ETDEWEB)

    Chrien, R.E.; Burrows, T.W. (eds.)

    1983-01-01

    Separate abstracts were prepared for the 29 papers presented. Workshop reports on decay heat, fission yields, beta- and gamma-ray spectroscopy, and delayed neutrons are included. An appendix contains a survey of the most recent compilations and evaluations containing fission product yield, fission product decay data, and delayed neutron yield information. (WHK)

  5. Estimation of shutdown heat generation rates in GHARR-1 due to ...

    African Journals Online (AJOL)

    Fission products decay power and residual fission power generated after shutdown of Ghana Research Reactor-1 (GHARR-1) by reactivity insertion accident were estimated by solution of the decay and residual heat equations. A Matlab program code was developed to simulate the heat generation rates by fission product ...

  6. The use of dielectric heating in particulate bed dryout experiments

    International Nuclear Information System (INIS)

    Stevens, G.F.; Willshire, S.J.

    1984-09-01

    Decay-heated, liquid-saturated debris beds arise in hypothetical severe accidents with LMFBR and PWR, and a large international effort is currently engaged in experimental studies of the cooling limitations of such beds. Dryout is one of the important cooling limitations. Dielectric heating offers a means of closely simulating decay heating in beds of irregular particles, and is under development at AEE Winfrith for application to experimental studies of dryout. This report describes progress to date. (author)

  7. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  8. On active shieldings in (ββ)0ν 76Ge decay experiments

    International Nuclear Information System (INIS)

    Garcia, E.; Morales, A.; Morales, J.; Nunez-Lagos, R.; Ortiz de Solorzano, A.; Puimedon, J.; Saenz, C.; Salinas, A.; Sarsa, M.L.; Villar, J.A.

    1992-01-01

    The sensitivity of an ultra low background Ge detector for the (ββ) 0ν decay of 76 Ge is estimated in two different experimental set-ups. The main difference between them is the inclusion or not of an active NaI shielding. We find that sensitivity of the Ge detector is not improved by this active shielding either for the O + -->O + or the O + -->2 + (ββ) 0ν transitions. Our results provide a valuable information for future 76 Ge enriched experiments. (orig.)

  9. 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

  10. A decaying factor accounts for contained activity in neuronal networks with no need of hierarchical or modular organization

    International Nuclear Information System (INIS)

    Amancio, Diego R; Oliveira Jr, Osvaldo N; Costa, Luciano da F

    2012-01-01

    The mechanisms responsible for containing activity in systems represented by networks are crucial in various phenomena, for example, in diseases such as epilepsy that affect the neuronal networks and for information dissemination in social networks. The first models to account for contained activity included triggering and inhibition processes, but they cannot be applied to social networks where inhibition is clearly absent. A recent model showed that contained activity can be achieved with no need of inhibition processes provided that the network is subdivided into modules (communities). In this paper, we introduce a new concept inspired in the Hebbian theory, through which containment of activity is achieved by incorporating a dynamics based on a decaying activity in a random walk mechanism preferential to the node activity. Upon selecting the decay coefficient within a proper range, we observed sustained activity in all the networks tested, namely, random, Barabási–Albert and geographical networks. The generality of this finding was confirmed by showing that modularity is no longer needed if the dynamics based on the integrate-and-fire dynamics incorporated the decay factor. Taken together, these results provide a proof of principle that persistent, restrained network activation might occur in the absence of any particular topological structure. This may be the reason why neuronal activity does not spread out to the entire neuronal network, even when no special topological organization exists. (paper)

  11. Thermographic studies of outer target heat fluxes on KSTAR

    Directory of Open Access Journals (Sweden)

    H.H. Lee

    2017-08-01

    Full Text Available A new infra-red (IR thermography system with high spatial resolution has been installed on KSTAR and is now mainly applied to measure the outer divertor heat load profile. The first measurement results of the outer divertor heat load profiles between ELMs have been applied to characterize the inter-ELMs outer divertor heat loads in KSTAR H-mode plasmas. In particular, the power decay length (λq of the divertor heat load profile has been determined by fitting the profile to a convolution of an exponential decay and a Gaussian function. The analysis on the power decay length shows a good agreement with the recent multi-machine λq scaling, which predicts λq of the inter-ELMs divertor heat load to be ∼1 mm under the standard H-mode scenario in ITER. The divertor IR thermography system has also successfully measured the strike point splitting of the outer divertor heat flux during the application of resonant magnetic perturbation (RMP fields. In addition, it has provided a clear evidence that the strike point splitting pattern depends on the RMP fields configuration.

  12. Temperature and heat flux scaling laws for isoviscous, infinite Prandtl number mixed heating convection.

    Science.gov (United States)

    Vilella, Kenny; Deschamps, Frederic

    2018-04-01

    Thermal evolution of terrestrial planets is controlled by heat transfer through their silicate mantles. A suitable framework for modelling this heat transport is a system including bottom heating (from the core) and internal heating, e.g., generated by secular cooling or by the decay of radioactive isotopes. The mechanism of heat transfer depends on the physical properties of the system. In systems where convection is able to operate, two different regimes are possible depending on the relative amount of bottom and internal heating. For moderate internal heating rates, the system is composed of active hot upwellings and cold downwellings. For large internal heating rates, the bottom heat flux becomes negative and the system is only composed of active cold downwellings. Here, we build theoretical scaling laws for both convective regimes following the approach of Vilella & Kaminski (2017), which links the surface heat flux and the temperature jump across both the top and bottom thermal boundary layer (TBL) to the Rayleigh number and the dimensionless internal heating rate. Theoretical predictions are then verified against numerical simulations performed in 2D and 3D-Cartesian geometry, and covering a large range of the parameter space. Our theoretical scaling laws are more successful in predicting the thermal structure of systems with large internal heating rates than that of systems with no or moderate internal heating. The differences between moderate and large internal heating rates are interpreted as differences in the mechanisms generating thermal instabilities. We identified three mechanisms: conductive growth of the TBL, instability impacting, and TBL erosion, the last two being present only for moderate internal heating rates, in which hot plumes are generated at the bottom of the system and are able to reach the surface. Finally, we apply our scaling laws to the evolution of the early Earth, proposing a new model for the cooling of the primordial magma ocean

  13. Correction of dynamic time-activity curves for gamma-camera dead time, radiotracer delivery, and radioactive decay: special considerations with ultrashort-lived radioisotopes

    International Nuclear Information System (INIS)

    Kuruc, A.; Zimmerman, R.E.; Treves, S.

    1985-01-01

    Time-vs.-activity curves obtained by using ultrashort-lived radioisotopes often need to be corrected for the effects of gamma-camera dead time and physical decay. Count loss due to gamma-camera dead time can be monitored by using an electronic oscillator incorporated into the gamma camera. Two algorithms that use this information to correct time-activity curves are discussed. It is also shown that the effect of physical decay on a time-activity curve is dependent on the time course of delivery of the radioisotope to the organ of interest. A mathematical technique that corrects physical decay is described

  14. Fungistatic activity of heat-treated flaxseed determined by response surface methodology.

    Science.gov (United States)

    Xu, Y; Hall, C; Wolf-Hall, C

    2008-08-01

    The objective of this study was to evaluate the effect of heat treatment on the fungistatic activity of flaxseed (Linum usitatissimum) in potato dextrose agar (PDA) medium and a fresh noodle system. The radial growth of Penicilliumn chrysogenum, Aspergillus flavus, and a Penicillium sp. isolated from moldy noodles, as well as the mold count of fresh noodle enriched with heat treated flaxseed, were used to assess antifungal activity. A central composite design in the response surface methodology was used to predict the effect of heating temperature and time on antifungal activity of flaxseed flour (FF). Statistical analysis determined that the linear terms of both variables (that is, heating temperature and time) and the quadratic terms of the heating temperature had significant (P<0.05) effects on the radial growth of all 3 test fungi and the mold count log-cycle reduction of fresh noodle. The interactions between the temperature and time were significant for all dependent variables (P<0.05). Significant reductions in antifungal activities were found when FF was subjected to high temperatures, regardless of heating time. In contrast, prolonging the heating time did not substantially affect the antifungal activities of FF at low temperature. However, 60% of the antifungal activity was retained after FF was heated at 100 degrees C for 15 min, which suggests a potential use of FF as an antifungal additive in food products subjected to low to mild heat treatments.

  15. 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

  16. ORIGEN-S Decay Data Library and Half-Life Uncertainties

    International Nuclear Information System (INIS)

    Hermann, O.W.

    1998-01-01

    The results of an extensive update of the decay data of the ORIGEN-S library are presented in this report. The updated decay data were provided for both the ORIGEN-S and ORIGEN2 libraries in the same project. A complete edit of the decay data plus the available half-life uncertainties are included in Appendix A. A detailed description of the types of data contained in the library, the format of the library, and the data sources are also presented. Approximately 24% of the library nuclides are stable, 66% were updated from ENDF/B-VI, about 8% were updated from ENSDF, and the remaining 2% were not updated. Appendix B presents a listing of percentage changes in decay heat from the old to the updated library for all nuclides containing a difference exceeding 1% in any parameter

  17. Study of non-domestic applications for active solar heating

    Energy Technology Data Exchange (ETDEWEB)

    Stammers, J.R.

    1997-11-01

    The UK Department of Trade and Industry (through ETSU) commissioned this study as part of its active solar programme. It was carried out from October 1996 to June 1997. The objective was to assess the potential for the use of active solar heating in non-domestic applications. The study was carried out by searching the literature, carrying out case studies and interviewing members of the solar industry and experts in other fields. There are currently about 45-50 active solar non-domestic schemes in operation in the UK, mostly for heating tap water in buildings of different types. The biggest potential for future non-domestic sales also lies in solar water heating for buildings. Most of the opportunities seem to be in the following building types: ablutions blocks in caravan and holiday camps, sheltered flats and hostels, nursing homes, office buildings, hotels and guest houses, and schools occupied during the summer. There are some other building types which might present niche markets for solar water heating. The market for active solar systems in space heating and cooling appears to be negligible. There is one other market for active solar heating in the non-domestic building sector. This is for warming water used to maintain stand-by generators at a temperature which allows them to kick in without delay in the event of a mains power failure. The main market is in buildings housing computers which control the provision of vital services, e.g. electricity, water and gas. (author)

  18. Decay of organic free radicals in γ-ray irradiated pepper during thermal treatment as detected by electron spin resonance spectroscopy

    International Nuclear Information System (INIS)

    Ukai, Mitsuko; Shimoyama, Yuhei

    2004-01-01

    Using electron spin resonance (ESR) spectroscopy we analysed the thermal decay process of radicals in γ-Irradiated pepper Upon irradiation, the satellite signals were newly induced and appeared at the symmetric positions of the organic free radical, i.e., the g=2.0 signal. Heat treatment decreased the satellite signals exponentially. The ESR signal of the pepper heated for more than 10 min was essentially the same as that before irradiation. To evaluate the radical decay by heat-treatment, we formulated a time-dependent master equation. We could evaluate the time constant of the radical decay based upon the general solution of the equation together with the nonlinear least-squares method

  19. Activation calculation and environmental safety analysis for fusion experimental breeder (FEB)

    Energy Technology Data Exchange (ETDEWEB)

    Kaiming, Feng [Southwest Inst. of Physics, Leshan, SC (China)

    1996-04-01

    An activation calculation code FDKR and decay chain data library AFDCDLIB are used to calculate the radioactivity, decay heat, dose rate and biological hazard potential (BHP) form activation products, actinides and fission products in a Fusion Experiment Breeder (FEB). The code and library are introduced briefly, and calculation results and decay curves of related hazards after one year operation with 150 MW fusion power are given. The total radioactivity inventory, decay heat and BHP are 5.74 x 10{sup 20} Bq, 8.34 MW and 4.08 x 10{sup 8} km{sup 3} of air, respectively, at shutdown. Results obtained show that the first wall of FEB can meet the nuclear waste disposal criteria for the NRC 10 CFR61 Class C after a few weeks from shutdown. The inventory of important actinides for the fuel reprocessing, such as {sup 232}U and {sup 237}Np were also calculated. It was shown that their concentrations do not excess the limit value of environmental safety required. (9 refs., 4 figs., 9 tabs.).

  20. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  1. Activity of two strobilurin fungicides against three species of decay fungi in agar plate tests

    Science.gov (United States)

    Juliet D. Tang; Tina Ciaramitaro; Maria Tomaso-Peterson; Susan V. Diehl

    2017-01-01

    The objective of this study was to examine the toxicity of strobilurin fungicides against wood decay fungi in order to assess their potential to act as a co-biocide for copper-based wood protection. Two strobilurin fungicides, Heritage (50% azoxystrobin active ingredient) and Insignia (20% pyraclostrobin active ingredients), and copper sulfate pentahydrate were tested...

  2. 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)

  3. 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)

  4. Radioactive decay data tables: A handbook of decay data for application to radiation dosimetry and radiological assessments

    Science.gov (United States)

    Kocher, D. C.; Smith, J. S.

    Decay data are presented for approximately 500 radionuclides including those occurring naturally in the environment, those of potential importance in routine or accidental releases from the nuclear fuel cycle, those of current interest in nuclear medicine and fusion reactor technology, and some of those of interest to Committee 2 of the International Commission on Radiological Protection for the estimation of annual limits on intake via inhalation and ingestion for occupationally exposed individuals. Physical processes involved in radioactive decay which produce the different types of radiation observed, methods used to prepare the decay data sets for each radionuclide in the format of the computerized evaluated nuclear structure data file, the tables of radioactive decay data, and the computer code MEDLIST used to produce the tables are described. Applications of the data to problems of interest in radiation dosimetry and radiological assessments are considered as well as the calculations of the activity of a daughter radionuclide relative to the activity of its parent in a radioactive decay chain.

  5. 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

  6. A gamma heating calculation methodology for research reactor application

    International Nuclear Information System (INIS)

    Lee, Y.K.; David, J.C.; Carcreff, H.

    2001-01-01

    Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

  7. Rare psi decays

    International Nuclear Information System (INIS)

    Partridge, R.

    1986-01-01

    Slightly more than ten years have passed since the psi was discovered, yet the study of psi decays continues to be an active and fruitful area of research. One reason for such longevity is that each successive experiment has increased their sensitivity over previous experiments either by improving detection efficiency or by increasing statistics. This has allowed the observation and, in some cases, detailed studies of rare psi decays. Branching ratios of ≅10-/sup 4/ are now routinely studied, while certain decay channels are beginning to show interesting effects at the 10-/sup 5/ level. Future experiments at the Beijing Electron Positron Collider (BEPC) have the potential for increasing sensitivities by one or two orders of magnitude, thus enabling many interesting studies impossible with current data samples. The author first examines the extent to which psi decays can be used to study electroweak phenomena. The remainder of this work is devoted to the more traditional task of using the psi to study quarks, gluons, and the properties of the strong interaction. Of particular interest is the study of radioactive psi decays, where a number of new particles have been discovered. Recent results regarding two of these particles, the θ(1700) and iota(1450), are discussed, as well as a study of the quark content of the eta and eta' using decays of the psi to vector-pseudoscalar final states

  8. Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

  9. Response of eddy activities to localized diabatic heating in Held-Suarez simulations

    Science.gov (United States)

    Lin, Yanluan; Zhang, Jishi; Li, Xingrui; Deng, Yi

    2018-01-01

    Widespread air pollutions, such as black carbon over East Asia in recent years, could induce a localized diabatic heating, and thus lead to localized static stability and meridional temperature gradient (MTG) changes. Although effect of static stability and MTG on eddies has been addressed by the linear baroclinic instability theory, impacts of a localized heating on mid-latitude eddy activities have not been well explored and quantified. Via a series of idealized global Held-Suarez simulations with different magnitudes of localized heating at different altitudes and latitudes, responses of mid-latitude eddy activity and circulation to these temperature perturbations are systematically investigated. Climatologically, the localized heating in the lower atmosphere induces a wave-like response of eddy activity near the mid-latitude jet stream. Over the heating region, eddy activity tends to be weakening due to the increased static stability. However, there are cyclonic anomalies over the upstream and downstream of the heating region. The zonal mean eddy activity weakens along the baroclinic zone due to reduced MTG and increased static stability. Furthermore, the response of eddy activity increased as the heating magnitude is increased and moved to higher altitudes. The influence of the heating decreases as the heating is prescribed further away from the climatological mid-latitude jet. This implies that the localized heating is most effective over the region with the maximum baroclinicity. Besides, enhanced storm track downstream of the localized heating area found here suggests that increased aerosols over East Asia might strengthen the North Pacific storm track.

  10. Quantitative Structure Activity Relationship of Cinnamaldehyde Compounds against Wood-Decaying Fungi

    Directory of Open Access Journals (Sweden)

    Dongmei Yang

    2016-11-01

    Full Text Available Cinnamaldehyde, of the genius Cinnamomum, is a major constituent of the bark of the cinnamon tree and possesses broad-spectrum antimicrobial activity. In this study, we used best multiple linear regression (BMLR to develop quantitative structure activity relationship (QSAR models for cinnamaldehyde derivatives against wood-decaying fungi Trametes versicolor and Gloeophyllun trabeum. Based on the two optimal QSAR models, we then designed and synthesized two novel cinnamaldehyde compounds. The QSAR models exhibited good correlation coefficients: R2Tv = 0.910 for Trametes versicolor and R2Gt = 0.926 for Gloeophyllun trabeum. Small errors between the experimental and calculated values of two designed compounds indicated that these two QSAR models have strong predictability and stability.

  11. The impact of the postharvest environment on the viability and virulence of decay fungi.

    Science.gov (United States)

    Liu, Jia; Sui, Yuan; Wisniewski, Michael; Xie, Zhigang; Liu, Yiqing; You, Yuming; Zhang, Xiaojing; Sun, Zhiqiang; Li, Wenhua; Li, Yan; Wang, Qi

    2018-07-03

    Postharvest decay of fruits, vegetables, and grains by fungal pathogens causes significant economic losses. Infected produce presents a potential health risk since some decay fungi produce mycotoxins that are hazardous to human health. Infections are the result of the interplay between host resistance and pathogen virulence. Both of these processes, however, are significantly impacted by environmental factors, such as temperature, UV, oxidative stress, and water activity. In the present review, the impact of various physical postharvest treatments (e.g., heat and UV) on the viability and virulence of postharvest pathogens is reviewed and discussed. Oxidative injury, protein impairment, and cell wall degradation have all been proposed as the mechanisms by which these abiotic stresses reduce fungal viability and pathogenicity. The response of decay fungi to pH and the ability of pathogens to modulate the pH of the host environment also affect pathogenicity. The effects of the manipulation of the postharvest environment by ethylene, natural edible coatings, and controlled atmosphere storage on fungal viability are also discussed. Lastly, avenues of future research are proposed.

  12. Photoproduction data for heating calculations

    International Nuclear Information System (INIS)

    Van der Marck, Steven C.; Koning, Arjan J.; Rochman, Dimitri

    2008-01-01

    For irradiations in a materials test reactor, the prediction of the amount of gamma heating in the reactor is important. Only a good predictive calculation will lead to an irradiation in which the specified temperatures are reached. The photons produced by fission product decay are often missing in spectrum calculations for a reactor, but the contribution of the photons can be computed effectively using engineering correlations for the amount of fission product decay and the ensuing photon spectrum. The prompt photons are usually calculated by a spectrum code based on the underlying nuclear data libraries. For most of the important nuclides, the nuclear data libraries contain data for the photon productions rates. However, there are still many nuclides for which the photon production data are missing, and some of these nuclides contribute to gamma heating. In this paper it is estimated what the contributions to heating are from photon production on nuclides such as 236 U, 238 Pu, 135 I, 135 Xe, 147 Pm, 148 Pm, 148m Pm, and 149 Sm. Also, simple arguments are given to judge the effect from photon production on all other (lumped) fission products, and from 28 Al decay. For all these calculations the High Flux Reactor is used as an example. (authors)

  13. 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.

  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. 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.

  16. 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.

  17. Do protons decay

    International Nuclear Information System (INIS)

    Litchfield, P.J.

    1984-09-01

    The experimental status of proton decay is reviewed after the Leipzig International conference, July 1984. A brief comparative description of the currently active experiments is given. From the overall samples of contained events it can be concluded that the experiments are working well and broadly agree with each other. The candidates for proton decay from each experiment are examined. Although several experiments report candidates at a higher rate than expected from background calculations, the validity of these calculations is still open to doubt. (author)

  18. Effects of Heat Acclimation on Photosynthesis, Antioxidant Enzyme Activities, and Gene Expression in Orchardgrass under Heat Stress

    Directory of Open Access Journals (Sweden)

    Xin Xin Zhao

    2014-09-01

    Full Text Available The present study was designed to examine the effects of heat acclimation on enzymatic activity, transcription levels, the photosynthesis processes associated with thermostability in orchardgrass (Dactylis glomerata L..The stomatal conductance (Gs, net photosynthetic rate (Pn, and transpiration rates (Tr of both heat-acclimated (HA and non-acclimated (NA plants were drastically reduced during heat treatment [using a 5-day heat stress treatment (38/30 °C ‒ day/night followed by a 3-day recovery under control conditions (25/20 °C ‒ day/night, in order to consolidate the second cycle was permitted]. Water use efficiency increased more steeply in the HA (4.9 times versus the NA (1.8 times plants, and the intercellular CO2 concentration decreased gently in NA (10.9% and HA (25.3% plants after 20 d of treatments compared to 0 days’. Furthermore, heat-acclimated plants were able to maintain significant activity levels of superoxide disumutase (SOD, catalase (CAT, guaiacol peroxidase (POD, and transcription levels of genes encoding these enzymes; in addition, HA plants displayed lower malondialdehyde content and lower electrolyte leakage than NA plants. These results suggest that maintenance of activity and transcription levels of antioxidant enzymes as well as photosynthesis are associated with variable thermostability in HA and NA plants. This likely occurs through cellular membrane stabilization and improvements in water use efficiency in the photosynthetic process during heat stress. The association between antioxidant enzyme activity and gene expression, both of which may vary with genetic variation in heat tolerance, is important to further understand the molecular mechanisms that contribute to heat tolerance.

  19. 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

  20. Diurnal variation in heat production related to some physical activities in laying hens.

    Science.gov (United States)

    Li, Y Z; Ito, T; Yamamoto, S

    1991-09-01

    1. Heat production, standing and eating activities, and hourly food intake of 4 laying hens were observed simultaneously and the effects of activity and food intake on heat production were studied. 2. Average heat production during the dark period (20.00 to 06.00 h) was 18.9 kJ/kgW0.75 h which was 33% lower than that during the light period. About 76% of the light-dark difference in the rate of heat production was probably associated with activity and posture. 3. Standing time, which included a range of behavioural activities, occupied 90% of the light period and the increased rate of heat production associated with standing was estimated to be about 18% of daily heat production. 4. Eating time occupied 40% of the light period; the heat production associated with eating activity represented about 5% of daily heat production or 3% of ME intake. 5. Because the regression of heat production on time spent eating agreed with the regression of heat production on hourly food intake, it is suggested that the energy expenditure associated with ad libitum feeding can be estimated for hens from the regression of heat production on hourly food intake.

  1. 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)

  2. R and D of low activated Fe-Mn-Cr high strength non-magnetic steel, (I). Screening test for constituent optimization and fundamental characterization test

    Energy Technology Data Exchange (ETDEWEB)

    Ishiyama, Shintaro; Fukaya, Kiyoshi; Eto, Motokuni; Kikuchi, Mituru [Japan Atomic Energy Research Inst., Tokyo (Japan); Sato, Ikuo; Kusuhashi, Mikio; Hatakeyama, Tuyoshi [Japan Steel Works Ltd., Tokyo (Japan); Takahashi, Heishitiro [Center for Advanced Research of Energy Technology, Hokkaido Univ., Sapporo (Japan)

    2000-02-01

    It is very important to develop low activated/non-magnetic materials as large scale structural materials for fusion reactors. In the structural design of JT-60SU, low activated/non-magnetic materials with high specific strength and low decay heat characterizations are required. In the present paper, a new low activated/non-magnetic material (15.5 Mn-16Cr-0.2N-0.3Si-0.2C (wt%)) based on the conventional high manganese steel with lower Ni, CO, C, N and Mn contents for the purpose of lower activation and decay heat was developed and the mass production conditions of the material were optimized. Fundamental material characterization tests of the new material developed in present study were carried out, and the following conclusions are derived ; (1) Lower activation characterizations with the new materials in the order of 1/10 of that of SUS316L steel, (2) Higher strength of the material in the order of 2{approx}3 of SUS316L steel and (3) Lower decay heat with higher thermal conductivity with comparison of SUS316L steel. (author)

  3. Experimental investigation of an active magnetic regenerative heat circulator applied to self-heat recuperation technology

    International Nuclear Information System (INIS)

    Kotani, Yui; Kansha, Yasuki; Ishizuka, Masanori; Tsutsumi, Atsushi

    2014-01-01

    An experimental investigation into an active magnetic regenerative (AMR) heat circulator based on self-heat recuperation technology, was conducted to evaluate its energy saving potential in heat circulation. In an AMR heat circulator, magnetocaloric effect is applied to recuperate the heat exergy of the process fluid. The recuperated heat can be reused to heat the feed process fluid and realize self-heat recuperation. In this paper, AMR heat circulator has newly been constructed to determine the amount of heat circulated when applied to self-heat recuperation and the energy consumption of the heat circulator. Gadolinium and water was used as the magnetocaloric working material and the process fluid, respectively. The heat circulated amount was determined by measuring the temperature of the process fluid and gadolinium. The net work input for heat circulation was obtained from the magnetizing and demagnetizing forces and the distance travelled by the magnetocaloric bed. The results were compared with the minimum work input needed for heat circulation derived from exergy loss during heat exchange. It was seen that the experimentally obtained value was close to the minimum work input needed for heat circulation. - Highlights: • AMR heat circulator has newly been constructed for experimental evaluation. • Heat circulation in the vicinity of Curie temperature was observed. • Energy consumption of an AMR heat circulator has been measured. • Energy saving for processes near Curie temperature of working material was seen

  4. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  5. Development of Monte Carlo decay gamma-ray transport calculation system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

    2001-06-01

    In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

  6. Decay properties of linear thermoelastic plates: Cattaneo versus Fourier law

    KAUST Repository

    Said-Houari, Belkacem

    2013-01-01

    In this article, we investigate the decay properties of the linear thermoelastic plate equations in the whole space for both Fourier and Cattaneo's laws of heat conduction. We point out that while the paradox of infinite propagation speed inherent

  7. Accurate and simple measurement method of complex decay schemes radionuclide activity

    International Nuclear Information System (INIS)

    Legrand, J.; Clement, C.; Bac, C.

    1975-01-01

    A simple method for the measurement of the activity is described. It consists of using a well-type sodium iodide crystal whose efficiency mith monoenergetic photon rays has been computed or measured. For each radionuclide with a complex decay scheme a total efficiency is computed; it is shown that the efficiency is very high, near 100%. The associated incertainty is low, in spite of the important uncertainties on the different parameters used in the computation. The method has been applied to the measurement of the 152 Eu primary reference [fr

  8. 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

  9. Influence of hot water dip and gamma irradiation on postharvest fungal decay of Galia melons

    International Nuclear Information System (INIS)

    Barkai-Golan, R.; Padova, R.; Ross, I.; Lapidot, M.; Copel, A.; Davidson, H.

    1993-01-01

    Dipping Galia melons in hot water at 52 deg C for 5 min or at 55 deg for 2 min resulted in 12-15% decay (caused by Alternaria alternata, Fusarium spp. and Trichothecium roseum) during prolonged storage (12 d at 6 deg plus 3 d at 18 deg ) compared with 75% decay in untreated fruit or 60% decay in cold-water-dipped fruit. Irradiation at 0.5 or 1 kGy had no significant effect on decay development. However, combination of heat treatment with a 0.5 kGy dose prevented fungal growth, resulting in 5% decay during storage. Combinations of heating with 1 kGy irradiation gave no improvement in anti-fungal effect over treatment with 0.5 kGy and sometimes resulted in a decreased suppressive effect. Reducing the duration of dipping at 55 deg from 2 to 0.5 min, applied alone or in combination with irradiation, considerably reduced the anti-fungal effect of the treatment. The effective combined treatment resulted in 12-15% of slight peel damage, but all the fruits were regarded as marketable. No differences in fruit firmness were recorded among the treatments

  10. Decay property of regularity-loss type of solutions in elastic solids with voids

    KAUST Repository

    Said-Houari, Belkacem; Messaoudi, Salim A.

    2013-01-01

    In this article, we consider two porous systems of nonclassical thermoelasticity in the whole real line. We discuss the long-time behaviour of the solutions in the presence of a strong damping acting, together with the heat effect, on the elastic equation and establish several decay results. Those decay results are shown to be very slow and of regularity-loss type. Some improvements of the decay rates have also been given, provided that the initial data belong to some weighted spaces. © 2013 Copyright Taylor and Francis Group, LLC.

  11. Decay property of regularity-loss type of solutions in elastic solids with voids

    KAUST Repository

    Said-Houari, Belkacem

    2013-12-01

    In this article, we consider two porous systems of nonclassical thermoelasticity in the whole real line. We discuss the long-time behaviour of the solutions in the presence of a strong damping acting, together with the heat effect, on the elastic equation and establish several decay results. Those decay results are shown to be very slow and of regularity-loss type. Some improvements of the decay rates have also been given, provided that the initial data belong to some weighted spaces. © 2013 Copyright Taylor and Francis Group, LLC.

  12. 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.

  13. 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

  14. Decay properties of linear thermoelastic plates: Cattaneo versus Fourier law

    KAUST Repository

    Said-Houari, Belkacem

    2013-02-01

    In this article, we investigate the decay properties of the linear thermoelastic plate equations in the whole space for both Fourier and Cattaneo\\'s laws of heat conduction. We point out that while the paradox of infinite propagation speed inherent in Fourier\\'s law is removed by changing to the Cattaneo law, the latter always leads to a loss of regularity of the solution. The main tool used to prove our results is the energy method in the Fourier space together with some integral estimates. We prove the decay estimates for initial data U0 ∈ Hs(ℝ) ∩ L1(ℝ). In addition, by restricting the initial data to U0 ∈ Hs(ℝ) ∩ L1,γ(ℝ) and γ ∈ [0, 1], we can derive faster decay estimates with the decay rate improvement by a factor of t-γ/2. © 2013 Copyright Taylor and Francis Group, LLC.

  15. 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

  16. 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

  17. Nuclear and activation characteristics of materials in 14.1-MeV and 2.5-MeV neutron field

    International Nuclear Information System (INIS)

    Seki, Yasushi; Takeyasu, Yuuichi.

    1988-11-01

    The nuclear and activation characteristics of various materials and elements of interest in terms of fusion reactor design are calculated and the results are graphically shown. The elements and materials are placed in a simple geometry modelling a blanket and shield of a fusion reactor. The neutrons with 14.1-MeV and 2.5-MeV energy are generated from the region represented as D-T and D-D plasma, respectively. The following activation characteristics after neutron irradiation are shown for each material and element; 1. Time evolution of induced activity, 2. Time evolution of decay heat, 3. Delayed gamma-ray dose distribution, 4. Decay heat distribution. In addition to the above activation characteristics, nuclear characteristics during the neutron irradiation, e.g. neutron energy spectra, neutron and gamma-ray flux distribution, nuclear heating distributions, and neutron and gamma-ray dose rate are also shown. (author)

  18. Heat transfer in a one-dimensional mixed convection loop

    International Nuclear Information System (INIS)

    Kim, Min Joon; Lee, Yong Bum; Kim, Yong Kyun; Kim, Jong Man; Nam, Ho Yun

    1999-01-01

    Effects of non-uniform heating in the core and additional forced circulation during decay heat removal operation are studied with a simplified mixed convection loop. The heat transfer coefficient is calculated analytically and measured experimentally. The analytic solution obtained from a one-dimensional heat equation is found to agree well with the experimental results. The effects of the non-uniform heating and the forced circulation are discussed

  19. Simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Larissa Cunha; Su, Jian, E-mail: larissa@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenhraria Nuclear; Cotta, Renato Machado, E-mail: cotta@mecanica.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (POLI/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica

    2015-07-01

    Single phase natural circulation circuits composed of two convective heat exchangers and connecting tubes are important for the passive heat removal from spent fuel pools (SFP). To keep the structural integrity of the stored spent fuel assemblies, continuously cooling has to be provided in order to avoid increase at the pool temperature and subsequent uncovering of the fuel and enhanced reaction between water and metal releasing hydrogen. Decay heat can achieve considerably high amounts of energy e.g. in the AP1000, considering the emergency fuel assemblies, the maximum heat decay will reach 13 MW in the 15th day (Westinghouse Electric Company, 2010). A highly efficient alternative to do so is by means of natural circulation, which is cost-effective compared to active cooling systems and is inherently safer since presents less associated devices and no external work is required. Many researchers have investigated safety and stability aspects of natural circulation loops (NCL). However, there is a lack of literature concerning the improvement of NCL through a standard unified methodology, especially for natural circulation circuits with two heat exchangers. In the present study, a simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchanges is presented. Relevant dimensionless key groups were proposed to for the design and safety analysis of a scaled NCL for the cooling of spent fuel storage pool with convective cooling and heating. (author)

  20. Radioactive decay data tables: a handbook of decay data for application to radiation dosimetry and radiological assessments

    International Nuclear Information System (INIS)

    Kocher, D.C.

    1982-01-01

    Reviews compendium containing recommended decay data for approx. 500 radionuclides of interest in nuclear medicine and fusion reactor technology or of potential importance in routine or accidental releases from the nuclear fuel cycle. Primary source of the decay data presented in this handbook is the Evaluated Nuclear Structure Data File (ENSDF), developed and maintained by the US Nuclear Data Network. Topics covered include various radioactive decay processes; evaluation process and standards of ENSDF; tables and computer code MEDLIST used to produce ENSDF tables; radiation dosimetry and radiological assessments; parent-daughter activity ratios wherever the adopted decay data may contain significant uncertainties or errors due to the lack of appropriate experimental data

  1. Heating analysis of cobalt adjusters in reactor core

    International Nuclear Information System (INIS)

    Mei Qiliang; Li Kang; Fu Yaru

    2011-01-01

    In order to produce 60 Co source for industry and medicine applications in CANDU-6 reactor, the stainless steel adjusters were replaced with the cobalt adjusters. The cobalt rod will generate the heat when it is irradiated by neutron and γ ray. In addition, 59 Co will be activated and become 60 Co, the ray released due to 60 Co decay will be absorbed by adjusters, and then the adjusters will also generate the heat. So the heating rate of adjusters to be changed during normal operation must be studied, which will be provided as the input data for analyzing the temperature field of cobalt adjusters and the relative heat load of moderator. MCNP code was used to simulate whole core geometric configuration in detail, including reactor fuel, control rod, adjuster, coolant and moderator, and to analyze the heating rate of the stainless steel adjusters and the cobalt adjusters. The maximum heating rate of different cobalt adjuster based on above results will be provided for the steady thermal hydraulic and accident analysis, and make sure that the reactor is safe on the thermal hydraulic. (authors)

  2. Effects of heat on the biological activity of wild Cordyceps sinensis

    Directory of Open Access Journals (Sweden)

    Pengkai Wu

    2015-01-01

    Conclusions: These results suggested that heat treatment does not adversely affect SOD or DNase activity, polysaccharide content, or cordycepin dissolution. Thus, heat treatment might be a safe processing method to extend the storage time of wild C. sinensis without compromising biological activity.

  3. 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

  4. 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

  5. Evidence against solar influence on nuclear decay constants

    Directory of Open Access Journals (Sweden)

    S. Pommé

    2016-10-01

    Full Text Available The hypothesis that proximity to the Sun causes variation of decay constants at permille level has been tested and disproved. Repeated activity measurements of mono-radionuclide sources were performed over periods from 200 days up to four decades at 14 laboratories across the globe. Residuals from the exponential nuclear decay curves were inspected for annual oscillations. Systematic deviations from a purely exponential decay curve differ from one data set to another and are attributable to instabilities in the instrumentation and measurement conditions. The most stable activity measurements of alpha, beta-minus, electron capture, and beta-plus decaying sources set an upper limit of 0.0006% to 0.008% to the amplitude of annual oscillations in the decay rate. Oscillations in phase with Earth's orbital distance to the Sun could not be observed within a 10−6 to 10−5 range of precision. There are also no apparent modulations over periods of weeks or months. Consequently, there is no indication of a natural impediment against sub-permille accuracy in half-life determinations, renormalisation of activity to a distant reference date, application of nuclear dating for archaeology, geo- and cosmochronology, nor in establishing the SI unit becquerel and seeking international equivalence of activity standards.

  6. Melatonin treatment attenuates postharvest decay and maintains nutritional quality of strawberry fruits (Fragaria×anannasa cv. Selva) by enhancing GABA shunt activity.

    Science.gov (United States)

    Aghdam, Morteza Soleimani; Fard, Javad Rezapour

    2017-04-15

    Fresh strawberry fruits as perishable commodities have a short postharvest life and are prone to postharvest fungal decay. In this study, the impact of 0, 1, 10, 100 and 1000μmol/L melatonin on attenuating fungal decay and maintaining nutritional quality of strawberry fruits was investigated during storage at 4°C for 12days. Melatonin treatment at 100μmol/L triggered H 2 O 2 accumulation, which result from higher superoxide dismutase (SOD) activity, associated with lower catalase (CAT) and ascorbate peroxidase (APX) activities, leading to fruits with lower decay. Higher H 2 O 2 accumulation was concurrent with higher phenylalanine ammonia lyase (PAL) enzyme activity leading to higher total phenols and anthocyanins accumulation along with higher DPPH scavenging capacity. Also, strawberry fruits treated with melatonin exhibited higher γ-aminobutyric acid transaminase (GABA-T) enzyme activity which ensured sufficient ATP supplying leading to higher unsaturated/saturated fatty acids (unSFA/SFA) ratio. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. Heat-flow properties of systems with alternate masses or alternate on-site potentials

    Science.gov (United States)

    Pereira, Emmanuel; Santana, Leonardo M.; Ávila, Ricardo

    2011-07-01

    We address a central issue of phononics: the search of properties or mechanisms to manage the heat flow in reliable materials. We analytically study standard and simple systems modeling the heat flow in solids, namely, the harmonic, self-consistent harmonic and also anharmonic chains of oscillators, and we show an interesting insulating effect: While in the homogeneous models the heat flow decays as the inverse of the particle mass, in the chain with alternate masses it decays as the inverse of the square of the mass difference, that is, it decays essentially as the mass ratio (between the smaller and the larger one) for a large mass difference. A similar effect holds if we alternate on-site potentials instead of particle masses. The existence of such behavior in these different systems, including anharmonic models, indicates that it is a ubiquitous phenomenon with applications in the heat flow control.

  8. Formation and decay of nuclei heated with high-energy antiprotons

    CERN Document Server

    Lott, B; Eades, J.; Egidy, T.v.; Figuera, P.; Fuchs, H.; Galin, J.; Gulda, K.; Goldenbaum, F.; Hilscher, D.; Jahnke, U.; Jastrzebski, J.; Kurcewicz, W.; Morjean, M.; Pausch, G.; Péghaire, A.; Pienkowski, L.; Polster, D.; Proschitzki, S.; Quednau, B.; Rossner, H.; Schmid, S.; Schmid, W.; Ziem, P.

    1999-01-01

    The decay of nuclei excited via the annihilation of 1.2 GeV antiprotons has been investigated. Thanks to the ability to determine the excitation energy, E sup * , for all events, largely irrespective of their mass partitions, the probabilities of the different channels at play could be estimated as a function of E sup *. The data show the prevalence of fission and evaporation up to E sup * = 4 MeV/nucleon, with no hint of a transition towards multifragmentation.

  9. Active Disturbance Rejection Control of a Heat Integrated Distillation Column

    DEFF Research Database (Denmark)

    Al-Kalbani, Fahad; Zhang, Jie; Bisgaard, Thomas

    2016-01-01

    pressure. However, the control of some HiDC processesis generally difficult due to the strong control loop interaction, high purity of the components and undesired disturbances. Active disturbance rejection control (ADRC) is used in this paperto control a simulated HiDC for separating benzene-toluene......Heat integrated distillation column (HiDC) is the most energy efficient distillation approach making efficient utilization of internal heat integration through heat pump. The rectifying section acts as a heat source with high pressure, while the stripping section operates as a heat sink with low...

  10. Relevance of tidal heating on large TNOs

    Science.gov (United States)

    Saxena, Prabal; Renaud, Joe P.; Henning, Wade G.; Jutzi, Martin; Hurford, Terry

    2018-03-01

    We examine the relevance of tidal heating for large Trans-Neptunian Objects, with a focus on its potential to melt and maintain layers of subsurface liquid water. Depending on their past orbital evolution, tidal heating may be an important part of the heat budget for a number of discovered and hypothetical TNO systems and may enable formation of, and increased access to, subsurface liquid water. Tidal heating induced by the process of despinning is found to be particularly able to compete with heating due to radionuclide decay in a number of different scenarios. In cases where radiogenic heating alone may establish subsurface conditions for liquid water, we focus on the extent by which tidal activity lifts the depth of such conditions closer to the surface. While it is common for strong tidal heating and long lived tides to be mutually exclusive, we find this is not always the case, and highlight when these two traits occur together. We find cases where TNO systems experience tidal heating that is a significant proportion of, or greater than radiogenic heating for periods ranging from100‧s of millions to a billion years. For subsurface oceans that contain a small antifreeze component, tidal heating due to very high initial spin states may enable liquid water to be preserved right up to the present day. Of particular interest is the Eris-Dysnomia system, which in those cases may exhibit extant cryovolcanism.

  11. Key role of asymmetric interactions in low-dimensional heat transport

    International Nuclear Information System (INIS)

    Chen, Shunda; Zhang, Yong; Wang, Jiao; Zhao, Hong

    2016-01-01

    We study the heat current autocorrelation function (HCAF) in one-dimensional, momentum-conserving lattices. In particular, we explore if there is any link between the decaying characteristics of the HCAF and asymmetric interparticle interactions. The Lennard-Jones model is investigated intensively in view of its significance to applications. It is found that, in the time range accessible to numerical simulations, the HCAF decays faster than power-law manners, and in some cases it decays even exponentially. Following the Green–Kubo formula, fast decay of the HCAF implies convergence of the heat conductivity, which is also corroborated by simulations. In addition, with a comparison to the Fermi–Pasta–Ulam-β model of symmetric interactions, the HCAF of the Fermi–Pasta–Ulam-α–β model of asymmetric interactions is also investigated. The results of all these studies lead to that, in certain ranges of parameters, fast decaying of the HCAF can be observed and correlated to the asymmetry degree of interactions. (paper: classical statistical mechanics, equilibrium and non-equilibrium)

  12. Calculation of heat fluxes induced by radio frequency heating on the actively cooled protections of ion cyclotron resonant heating (ICRH) and lower hybrid (LH) antennas in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ritz, G., E-mail: Guillaume.ritz@gmail.com [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Corre, Y., E-mail: Yann.corre@cea.fr [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Rault, M.; Missirlian, M. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint Paul-lez-Durance (France); Martinez, A.; Ekedahl, A.; Colas, L.; Guilhem, D.; Salami, M.; Loarer, T. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France)

    2013-10-15

    Highlights: ► The heat flux generated by radiofrequency (RF) heating was calculated using Tore Supra's heating antennas. ► The highest heat flux value, generated by ions accelerated in RF-rectified sheath potentials, was 5 MW/m{sup 2}. ► The heat flux on the limiters of antennas was in the same order of magnitude as that on the toroidal pumping limiter. -- Abstract: Lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) are recognized as important auxiliary heating and current drive methods for present and next step fusion devices. However, these radio frequency (RF) systems generate a heat flux up to several MW/m{sup 2} on the RF antennas during plasma operation. This paper focuses on the determination of the heat flux deposited on the lateral protections of the RF antennas in Tore Supra. The heat flux was calculated by finite element method (FEM) using a model of the lateral protection. The FEM calculation was based on surface temperature measurements using infrared cameras monitoring the RF antennas. The heat flux related to the acceleration of electrons in front of the LHCD grills (LHCD active) and to the acceleration of ions in RF-rectified sheath potentials (ICRH active) were calculated. Complementary results on the heat flux related to fast ions (ICRH active with a relatively low magnetic field) are also reported in this paper.

  13. Actively heated high-resolution fiber-optic-distributed temperature sensing to quantify streambed flow dynamics in zones of strong groundwater upwelling

    Science.gov (United States)

    Briggs, Martin A.; Buckley, Sean F.; Bagtzoglou, Amvrossios C.; Werkema, Dale D.; Lane, John W.

    2016-01-01

    Zones of strong groundwater upwelling to streams enhance thermal stability and moderate thermal extremes, which is particularly important to aquatic ecosystems in a warming climate. Passive thermal tracer methods used to quantify vertical upwelling rates rely on downward conduction of surface temperature signals. However, moderate to high groundwater flux rates (>−1.5 m d−1) restrict downward propagation of diurnal temperature signals, and therefore the applicability of several passive thermal methods. Active streambed heating from within high-resolution fiber-optic temperature sensors (A-HRTS) has the potential to define multidimensional fluid-flux patterns below the extinction depth of surface thermal signals, allowing better quantification and separation of local and regional groundwater discharge. To demonstrate this concept, nine A-HRTS were emplaced vertically into the streambed in a grid with ∼0.40 m lateral spacing at a stream with strong upward vertical flux in Mashpee, Massachusetts, USA. Long-term (8–9 h) heating events were performed to confirm the dominance of vertical flow to the 0.6 m depth, well below the extinction of ambient diurnal signals. To quantify vertical flux, short-term heating events (28 min) were performed at each A-HRTS, and heat-pulse decay over vertical profiles was numerically modeled in radial two dimension (2-D) using SUTRA. Modeled flux values are similar to those obtained with seepage meters, Darcy methods, and analytical modeling of shallow diurnal signals. We also observed repeatable differential heating patterns along the length of vertically oriented sensors that may indicate sediment layering and hyporheic exchange superimposed on regional groundwater discharge.

  14. Impact of dark matter decays and annihilations on structure formation

    NARCIS (Netherlands)

    Mapelli, M.; Ripamonti, E.

    2007-01-01

    Abstract: We derived the evolution of the energy deposition in the intergalactic medium (IGM) by different decaying (or annihilating) dark matter (DM) candidates. Heavy annihilating DM particles (with mass larger than a few GeV) have no influence on reionization and heating, even if we assume that

  15. Active ion temperature measurement with heating neutral beam

    International Nuclear Information System (INIS)

    Miura, Yukitoshi; Matsuda, Toshiaki; Yamamoto, Shin

    1987-03-01

    When the heating neutral-beam (hydrogen beam) is injected into a deuterium plasma, the density of neutral particles is increased locally. By using this increased neutral particles, the local ion temperature is measured by the active charge-exchange method. The analyzer is the E//B type mass-separated neutral particle energy analyzer and the measured position is about one third outside of the plasma radius. The deuterium energy spectrum is Maxwellian, and the temperature is increased from 350 eV to 900 eV during heating. Since the local hydrogen to deuterium density concentration and the density of the heating neutral-beam as well as the ion temperature can be obtained good S/N ratio, the usefulness of this method during neutral-beam heating is confirmed by this experiment. (author)

  16. Harvesting the decay energy of 26Al to drive lightning discharge in protoplanetary discs

    Science.gov (United States)

    Johansen, Anders; Okuzumi, Satoshi

    2018-01-01

    Chondrules in primitive meteorites likely formed by recrystallisation of dust aggregates that were flash-heated to nearly complete melting. Chondrules may represent the building blocks of rocky planetesimals and protoplanets in the inner regions of protoplanetary discs, but the source of ubiquitous thermal processing of their dust aggregate precursors remains elusive. Here we demonstrate that escape of positrons released in the decay of the short-lived radionuclide 26Al leads to a large-scale charging of dense pebble structures, resulting in neutralisation by lightning discharge and flash-heating of dust and pebbles. This charging mechanism is similar to a nuclear battery where a radioactive source charges a capacitor. We show that the nuclear battery effect operates in circumplanetesimal pebble discs. The extremely high pebble densities in such discs are consistent with conditions during chondrule heating inferred from the high abundance of sodium within chondrules. The sedimented mid-plane layer of the protoplanetary disc may also be prone to charging by the emission of positrons, if the mass density of small dust there is at least an order of magnitude above the gas density. Our results imply that the decay energy of 26Al can be harvested to drive intense lightning activity in protoplanetary discs. The total energy stored in positron emission is comparable to the energy needed to melt all solids in the protoplanetary disc. The efficiency of transferring the positron energy to the electric field nevertheless depends on the relatively unknown distribution and scale-dependence of pebble density gradients in circumplanetesimal pebble discs and in the protoplanetary disc mid-plane layer.

  17. Effect of Heat Treatment on the Surface Properties of Activated Carbons

    Directory of Open Access Journals (Sweden)

    Meriem Belhachemi

    2011-01-01

    Full Text Available This work reports the effect of heat treatment on the porosity and surface chemistry of two series of activated carbons prepared from a local agricultural biomass material, date pits, by physical activation with carbon dioxide and steam. Both series samples were oxidized with nitric acid and subsequently heat treated under N2 at 973 K in order to study the effect of these treatments in porosity and surface functional groups of activated carbons. When the activated carbons were heat treated after oxidation the surface area and the pore volume increase for both activated carbons prepared by CO2 and steam activations. However the amount of surface oxygen complexes decreases, the samples keep the most stable oxygen surface groups evolved as CO by temperature-programmed desorption experiments at high temperature. The results show that date pits can be used as precursors to produce activated carbons with a well developed porosity and tailored oxygen surface groups.

  18. HAGRID/ VANDLE spectroscopy of Rb decays

    Science.gov (United States)

    King, Thomas; Grzywacz, Robert; Taylor, Steven; Paulauskas, Stanley; Smith, Karl; Vandle Collaboration

    2017-09-01

    Many neutron-rich isotopes that contribute in both decay heat production and r-process nucleosynthesis have substantial beta-delayed neutron branching ratios. Beta-delayed neutron emission is a relatively complicated mechanism which can leave the daughter in an gamma-emitting excited state. A comprehensive understanding of their energy output and decay strength, S_beta, therefore requires the detection of both neutrons and gamma rays in coincidence. A series of measurements of delayed neutron precursors were performed at the On-Line Test Facility (OLTF) at the Oak Ridge National Laboratories using chemically selective ion sources and an enhanced VANDLE array. The main goal of this experiment was to revisit the decays of IAEA-marked priority precursors, including bromine, rubidium, cesium, and iodine, that are required to model the global properties in the fission of 238U.The unique data set, with neutron and gamma ray coincidences, benefited from the addition of a high-efficiency gamma-ray array, consisting of 16 LaBr3 crystals (HAGRiD), and a set of large volume NaI detectors to the VANDLE array. Characterization of and preliminary results from the new gamma-ray array for the decays of 94Rb and 97Rb will be presented. National Nuclear Security Administration under the Stewardship Science Academic Alliances program through DOE Award No. DE-NA0002132 and the Office of Nuclear Physics, U.S. Department of Energy under Award No. DE-FG02-96ER40983.

  19. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.

    1979-10-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  20. 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)

  1. Neutron decay, semileptonic hyperon decay and the Cabibbo model

    International Nuclear Information System (INIS)

    Siebert, H.W.

    1989-01-01

    The decay rates and formfactor ratios of neutron decay and semileptonic hyperon decays are compared in the framework of the Cabibbo model. The results indicate SU(3) symmetry breaking. The Kobayashi-Maskawa matrix element V us determined from these decays is in good agreement with the value determined from K→πeν decays, and with unitarity of the KM-matrix. (orig.)

  2. Active solar heating industry development study

    International Nuclear Information System (INIS)

    1995-01-01

    Despite the fact that solar water heating systems are technologically viable and commercially available, this Energy Technology Support Unit report shows that there is no established market in the United Kingdom. The Solar Trade Association (STA) has undertaken an Active Solar Heating Industry Development Study which is reported here. The data is derived from a questionnaire survey completed by companies, organizations and individuals operating within the industry. Information was also gathered from utility companies, and STAs elsewhere in Europe. Barriers which need to be overcome include lack of public awareness, especially in the construction industry, lack of capital investment and other financial disincentives, little or no government support, and lack of organization and quality monitoring and assurance within the industry itself. (UK)

  3. 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.

  4. The 49K beta decay

    International Nuclear Information System (INIS)

    Hansen, P.G.; Huck, A.; Klotz, G.; Knipper, A.; Miehe, C.; Walter, G.; Jonson, B.; Mattsson, S.; Ravn, H.L.; Kratz, K.L.

    1981-01-01

    The decay of 49 K has been studied through neutron and gamma spectroscopy techniques. The 49 K activity was formed by 600 MeV proton fragmentation reactions in a uranium carbide target. The observed β-strength, in addition to the general behaviour expected from the gross theory of β-decay, displays two resonances centered at about 6.5 MeV and 9.5 MeV in 49 Ca. This structure is discussed in simple shell-model terms. (orig.)

  5. 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.

  6. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  7. Modeling and analysis of alternative concept of ITER vacuum vessel primary heat transfer system

    International Nuclear Information System (INIS)

    Carbajo, Juan; Yoder, Graydon; Dell'Orco, G.; Curd, Warren; Kim, Seokho

    2010-01-01

    A RELAP5-3D model of the ITER (Latin for 'the way') vacuum vessel (VV) primary heat transfer system has been developed to evaluate a proposed design change that relocates the heat exchangers (HXs) from the exterior of the tokamak building to the interior. This alternative design protects the HXs from external hazards such as wind, tornado, and aircraft crash. The proposed design integrates the VV HXs into a VV pressure suppression system (VVPSS) tank that contains water to condense vapour in case of a leak into the plasma chamber. The proposal is to also use this water as the ultimate sink when removing decay heat from the VV system. The RELAP5-3D model has been run under normal operating and abnormal (decay heat) conditions. Results indicate that this alternative design is feasible, with no effects on the VVPSS tank under normal operation and with tank temperature and pressure increasing under decay heat conditions resulting in a requirement to remove steam generated if the VVPSS tank low pressure must be maintained.

  8. Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

    International Nuclear Information System (INIS)

    Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.

    1977-01-01

    The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

  9. Global existence and decay of solutions of the Cauchy problem in thermoelasticity with second sound

    KAUST Repository

    Kasimov, Aslan R.; Racke, Reinhard; Said-Houari, Belkacem

    2013-01-01

    We consider the one-dimensional Cauchy problem in non-linear thermoelasticity with second sound, where the heat conduction is modelled by Cattaneo's law. After presenting decay estimates for solutions to the linearized problem, including refined estimates for data in weighted Lebesgue-spaces, we prove a global existence theorem for small data together with improved decay estimates, in particular for derivatives of the solutions. © 2013 Taylor & Francis.

  10. Weak decays of doubly heavy baryons. Multi-body decay channels

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Yu-Ji; Wang, Wei; Xing, Ye; Xu, Ji [Shanghai Jiao Tong University, INPAC, Shanghai Key Laboratory for Particle Physics and Cosmology, MOE Key Laboratory for Particle Physics, Astrophysics and Cosmology, School of Physics and Astronomy, Shanghai (China)

    2018-01-15

    The newly-discovered Ξ{sub cc}{sup ++} decays into the Λ{sub c}{sup +}K{sup -}π{sup +}π{sup +}, but the experimental data has indicated that this decay is not saturated by any two-body intermediate state. In this work, we analyze the multi-body weak decays of doubly heavy baryons Ξ{sub cc}, Ω{sub cc}, Ξ{sub bc}, Ω{sub bc}, Ξ{sub bb} and Ω{sub bb}, in particular the three-body nonleptonic decays and four-body semileptonic decays. We classify various decay modes according to the quark-level transitions and present an estimate of the typical branching fractions for a few golden decay channels. Decay amplitudes are then parametrized in terms of a few SU(3) irreducible amplitudes. With these amplitudes, we find a number of relations for decay widths, which can be examined in future. (orig.)

  11. Active latent heat storage with a screw heat exchanger - experimental results for heat transfer and concept for high pressure steam

    Science.gov (United States)

    Zipf, Verena; Willert, Daniel; Neuhäuser, Anton

    2016-05-01

    An innovative active latent heat storage concept was invented and developed at Fraunhofer ISE. It uses a screw heat exchanger (SHE) for the phase change during the transport of a phase change material (PCM) from a cold to a hot tank or vice versa. This separates heat transfer and storage tank in comparison to existing concepts. A test rig has been built in order to investigate the heat transfer coefficients of the SHE during melting and crystallization of the PCM. The knowledge of these characteristics is crucial in order to assess the performance of the latent heat storage in a thermal system. The test rig contains a double shafted SHE, which is heated or cooled with thermal oil. The overall heat transfer coefficient U and the convective heat transfer coefficient on the PCM side hPCM both for charging and discharging have been calculated based on the measured data. For charging, the overall heat transfer coefficient in the tested SHE was Uch = 308 W/m2K and for discharging Udis = 210 W/m2K. Based on the values for hPCM the overall heat transfer coefficients for a larger SHE with steam as heat transfer fluid and an optimized geometry were calculated with Uch = 320 W/m2K for charging and Udis = 243 W/m2K for discharging. For pressures as high as p = 100 bar, an SHE concept has been developed, which uses an organic fluid inside the flight of the SHE as working media. With this concept, the SHE can also be deployed for very high pressure, e.g. as storage in solar thermal power plants.

  12. 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

    active fire area is along the frontal fireline and all the smoke is blown to leeward direction of the nuclear power plant site. The time-dependent change of particle matter density is utilized to evaluate a cumulative amount of particle matter at the air filter of the decay heat removal system. The total amount of particle matter at the air filter is around several hundred grams per m"2 which is well below the operational limit of the air filter of 15 kg/m"2. (author)

  13. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  14. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  15. Validation of the TASS/SMR-S Code for the PRHRS Condensation Heat Transfer Model

    International Nuclear Information System (INIS)

    Jun, In Sub; Yang, Soo Hyoung; Chung, Young Jong; Lee, Won Jae

    2011-01-01

    When some accidents or events are occurred in the SMART, the secondary system is used to remove the core decay heat for the long time such as a feedwater system. But if the feedwater system can't remove the residual core heat because of its malfunction, the core decay heat is removed using the Passive Residual Heat Removal System (PRHRS). The PRHRS is passive type safety system adopted to enhance the safety of the SMART. It can fundamentally eliminate the uncertainty of operator action. TASS/SMR-S (Transient And Setpoint Simulation/ System-integrated Modular Reactor-Safety) code has various heat transfer models reflecting the design features of the SMART. One of the heat transfer models is the PRHRS condensation heat transfer model. The role of this model is to calculate the heat transfer coefficient in the heat exchanger (H/X) tube side using the relevant heat transfer correlations for all of the heat transfer modes. In this paper, the validation of the condensation heat transfer model was carried out using the POSTECH H/X heat transfer test

  16. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10-4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  17. 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)

  18. Weak decays

    International Nuclear Information System (INIS)

    Wojcicki, S.

    1978-11-01

    Lectures are given on weak decays from a phenomenological point of view, emphasizing new results and ideas and the relation of recent results to the new standard theoretical model. The general framework within which the weak decay is viewed and relevant fundamental questions, weak decays of noncharmed hadrons, decays of muons and the tau, and the decays of charmed particles are covered. Limitation is made to the discussion of those topics that either have received recent experimental attention or are relevant to the new physics. (JFP) 178 references

  19. Radioactive-nuclide decay data in science and technology

    International Nuclear Information System (INIS)

    Reich, C.W.; Helmer, R.G.

    1975-01-01

    The scope of ENDF/B has recently been expanded to include radioactive-nuclide decay data. In this paper, the content and organization of the decay data which are included in ENDF/B are presented and discussed. The application of decay data in a wide variety of nuclear-related activities is illustrated by a number of examples. Two items pointed up by the ENDF/B decay-data compilation effort are treated: the identification of deficiencies in the data; and the importance of a radioactive-nuclide metrology effort oriented toward supplying these needs in a systematic fashion. 3 figures, 2 tables

  20. Radioactive-nuclide decay data in science and technology

    International Nuclear Information System (INIS)

    Reich, C.W.; Helmer, R.G.

    1975-01-01

    The scope of ENDF/B has recently been expanded to include radioactive-nuclide decay data. In this paper, the content and organization of the decay data which are included in ENDF/B are presented and discussed. The application of decay data in a wide variety of nuclear-related activities is illustrated by a number of examples. Two items pointed up by the ENDF/B decay-data compilation effort are treated: the identification of deficiencies in the data; and the importance of a radioactive-nuclide metrology effort oriented toward supplying these needs in a systematic fashion. (3 figures, 1 table)

  1. Global existence and decay of solutions of the Cauchy problem in thermoelasticity with second sound

    KAUST Repository

    Kasimov, Aslan R.

    2013-06-04

    We consider the one-dimensional Cauchy problem in non-linear thermoelasticity with second sound, where the heat conduction is modelled by Cattaneo\\'s law. After presenting decay estimates for solutions to the linearized problem, including refined estimates for data in weighted Lebesgue-spaces, we prove a global existence theorem for small data together with improved decay estimates, in particular for derivatives of the solutions. © 2013 Taylor & Francis.

  2. Local energy decay for linear wave equations with variable coefficients

    Science.gov (United States)

    Ikehata, Ryo

    2005-06-01

    A uniform local energy decay result is derived to the linear wave equation with spatial variable coefficients. We deal with this equation in an exterior domain with a star-shaped complement. Our advantage is that we do not assume any compactness of the support on the initial data, and its proof is quite simple. This generalizes a previous famous result due to Morawetz [The decay of solutions of the exterior initial-boundary value problem for the wave equation, Comm. Pure Appl. Math. 14 (1961) 561-568]. In order to prove local energy decay, we mainly apply two types of ideas due to Ikehata-Matsuyama [L2-behaviour of solutions to the linear heat and wave equations in exterior domains, Sci. Math. Japon. 55 (2002) 33-42] and Todorova-Yordanov [Critical exponent for a nonlinear wave equation with damping, J. Differential Equations 174 (2001) 464-489].

  3. Influence of Magnetic Field Decay on Electron Capture in Magnetars ...

    Indian Academy of Sciences (India)

    Abstract. The de-excited energy of electron capture (EC) induced by magnetic field decay may be a new source for heating magnetar crust, so we do a quantitative calculation on EC process near the outer crust and analyse their influence on persistent X-ray radiation of magnetars, adopt- ing the experimental data or the ...

  4. Space shuttle/food system study. Volume 2, Appendix A: Active heating system-screening analysis. Appendix B: Reconstituted food heating techniques analysis

    Science.gov (United States)

    1974-01-01

    Technical data are presented which were used to evaluate active heating methods to be incorporated into the space shuttle food system design, and also to evaluate the relative merits and penalties associated with various approaches to the heating of rehydrated food during space flight. Equipment heating candidates were subject to a preliminary screening performed by a selection rationale process which considered the following parameters; (1) gravitational effect; (2) safety; (3) operability; (4) system compatibility; (5) serviceability; (6) crew acceptability; (7) crew time; (8) development risk; and (9) operating cost. A hot air oven, electrically heated food tray, and microwave oven were selected for further consideration and analysis. Passive, semi-active, and active food preparation approaches were also studied in an effort to determine the optimum method for heating rehydrated food. Potential complexity, cost, vehicle impact penalties, and palatability were considered in the analysis. A summary of the study results is provided along with cost estimates for each of the potential sytems

  5. The Impacts of Heating Strategy on Soil Moisture Estimation Using Actively Heated Fiber Optics.

    Science.gov (United States)

    Dong, Jianzhi; Agliata, Rosa; Steele-Dunne, Susan; Hoes, Olivier; Bogaard, Thom; Greco, Roberto; van de Giesen, Nick

    2017-09-13

    Several recent studies have highlighted the potential of Actively Heated Fiber Optics (AHFO) for high resolution soil moisture mapping. In AHFO, the soil moisture can be calculated from the cumulative temperature ( T cum ), the maximum temperature ( T max ), or the soil thermal conductivity determined from the cooling phase after heating ( λ ). This study investigates the performance of the T cum , T max and λ methods for different heating strategies, i.e., differences in the duration and input power of the applied heat pulse. The aim is to compare the three approaches and to determine which is best suited to field applications where the power supply is limited. Results show that increasing the input power of the heat pulses makes it easier to differentiate between dry and wet soil conditions, which leads to an improved accuracy. Results suggest that if the power supply is limited, the heating strength is insufficient for the λ method to yield accurate estimates. Generally, the T cum and T max methods have similar accuracy. If the input power is limited, increasing the heat pulse duration can improve the accuracy of the AHFO method for both of these techniques. In particular, extending the heating duration can significantly increase the sensitivity of T cum to soil moisture. Hence, the T cum method is recommended when the input power is limited. Finally, results also show that up to 50% of the cable temperature change during the heat pulse can be attributed to soil background temperature, i.e., soil temperature changed by the net solar radiation. A method is proposed to correct this background temperature change. Without correction, soil moisture information can be completely masked by the background temperature error.

  6. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's Commission at Three Mile Island

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Wilson, W.B.

    1979-10-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis.

  7. Active control of near-field radiative heat transfer between graphene-covered metamaterials

    Science.gov (United States)

    Zhao, Qimei; Zhou, Ting; Wang, Tongbiao; Liu, Wenxing; Liu, Jiangtao; Yu, Tianbao; Liao, Qinghua; Liu, Nianhua

    2017-04-01

    In this study, the near-field radiative heat transfer between graphene-covered metamaterials is investigated. The electric surface plasmons (SPs) supported by metamaterials can be coupled with the SPs supported by graphene. The near-field heat transfer between the graphene-covered metamaterials is significantly larger than that between metamaterials because of the strong coupling in our studied frequency range. The relationship between heat flux and chemical potential is studied for different vacuum gaps. Given that the chemical potential of graphene can be tuned by the external electric field, heat transfer can be actively controlled by modulating the chemical potential. The heat flux for certain vacuum gaps can reach a maximum value when the chemical potential is at a particular value. The results of this study are beneficial for actively controlling energy transfer.

  8. Active control of near-field radiative heat transfer between graphene-covered metamaterials

    International Nuclear Information System (INIS)

    Zhao, Qimei; Zhou, Ting; Wang, Tongbiao; Liu, Wenxing; Liu, Jiangtao; Yu, Tianbao; Liao, Qinghua; Liu, Nianhua

    2017-01-01

    In this study, the near-field radiative heat transfer between graphene-covered metamaterials is investigated. The electric surface plasmons (SPs) supported by metamaterials can be coupled with the SPs supported by graphene. The near-field heat transfer between the graphene-covered metamaterials is significantly larger than that between metamaterials because of the strong coupling in our studied frequency range. The relationship between heat flux and chemical potential is studied for different vacuum gaps. Given that the chemical potential of graphene can be tuned by the external electric field, heat transfer can be actively controlled by modulating the chemical potential. The heat flux for certain vacuum gaps can reach a maximum value when the chemical potential is at a particular value. The results of this study are beneficial for actively controlling energy transfer. (paper)

  9. Investigation of second grade fluid through temperature dependent thermal conductivity and non-Fourier heat flux

    Science.gov (United States)

    Hayat, T.; Ahmad, Salman; Khan, M. Ijaz; Alsaedi, A.; Waqas, M.

    2018-06-01

    Here we investigated stagnation point flow of second grade fluid over a stretchable cylinder. Heat transfer is characterized by non-Fourier law of heat flux and thermal stratification. Temperature dependent thermal conductivity and activation energy are also accounted. Transformations procedure is applying to transform the governing PDE's into ODE's. Obtained system of ODE's are solved analytically by HAM. Influence of flow variables on velocity, temperature, concentration, skin friction and Sherwood number are analyzed. Obtained outcome shows that velocity enhanced through curvature parameter, viscoelastic parameter and velocities ratio variable. Temperature decays for larger Prandtl number, thermal stratification, thermal relaxation and curvature parameter. Sherwood number and concentration field show opposite behavior for higher estimation of activation energy, reaction rate, curvature parameter and Schmidt number.

  10. Heating entrepreneur activity in 2003

    International Nuclear Information System (INIS)

    Nikkola, A.; Solmio, H.

    2004-01-01

    According to TTS Institute information, at the end of 2003 there were heating entrepreneurs responsible for fuel management and heat production in at least 212 heating plants in Finland. The number of operative plants increased by 36 from the previous year. At the end of 2003, the total boiler capacity for solid fuel in the plants managed by the heating entrepreneurs exceeded 100 megawatts. The average boiler capacity of the plants was 0.5 megawatts. Heating entrepreneur-ship was most common in west Finland, where 40 percent of the plants are located. There were some 94 heating plants managed by cooperatives or limited companies. Single entrepre neurs or entrepreneur networks consisting of several entrepreneurs were responsible for heat production in 117 plants. Heating entrepreneurs used approximately 290,000 loose cubic metres of forest chips, which is about seven percent of the volume used for heating and power plant energy production in 2003. In addition, the heating entrepreneurs used a total of 40,000 loose cubic metres of other wood fuel and an estimated 20,000 loose cubic metres of sod and milled peat. Municipalities are still the most important customer group for heating entrepreneurs. However, thenumber of private customers is growing. Industrial company, other private company or properly was the main customer already for every fourth plant established during 2003. (orig.)

  11. Heat activation and stability of amylases from Bacillus species | Ajayi ...

    African Journals Online (AJOL)

    Leitch and Collier sporulating Bacillus medium was used to isolate some strains of Bacillus species from soil, wastewater and food sources in Ibadan, Oyo State, Nigeria, by heat activation method. Heat treatment at 80oC allowed the growth of sporulating Bacillus species, in the culture sample source without other bacteria ...

  12. Analysis of natural convection in volumetrically-heated melt pools

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.

    1996-12-01

    Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation

  13. Repairing Student Misconceptions in Heat Transfer Using Inquiry-Based Activities

    Science.gov (United States)

    Prince, Michael; Vigeant, Margot; Nottis, Katharyn

    2016-01-01

    Eight inquiry-based activities, described here in sufficient detail for faculty to adopt in their own courses, were designed to teach students fundamental concepts in heat transfer. The concept areas chosen were (1) factors affecting the rate vs. amount of heat transfer, (2) temperature vs. perceptions of hot and cold, (3) temperature vs. energy…

  14. 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.

  15. How a retrotransposon exploits the plant's heat stress response for its activation.

    Directory of Open Access Journals (Sweden)

    Vladimir V Cavrak

    2014-01-01

    Full Text Available Retrotransposons are major components of plant and animal genomes. They amplify by reverse transcription and reintegration into the host genome but their activity is usually epigenetically silenced. In plants, genomic copies of retrotransposons are typically associated with repressive chromatin modifications installed and maintained by RNA-directed DNA methylation. To escape this tight control, retrotransposons employ various strategies to avoid epigenetic silencing. Here we describe the mechanism developed by ONSEN, an LTR-copia type retrotransposon in Arabidopsis thaliana. ONSEN has acquired a heat-responsive element recognized by plant-derived heat stress defense factors, resulting in transcription and production of full length extrachromosomal DNA under elevated temperatures. Further, the ONSEN promoter is free of CG and CHG sites, and the reduction of DNA methylation at the CHH sites is not sufficient to activate the element. Since dividing cells have a more pronounced heat response, the extrachromosomal ONSEN DNA, capable of reintegrating into the genome, accumulates preferentially in the meristematic tissue of the shoot. The recruitment of a major plant heat shock transcription factor in periods of heat stress exploits the plant's heat stress response to achieve the transposon's activation, making it impossible for the host to respond appropriately to stress without losing control over the invader.

  16. HEATING OF FLARE LOOPS WITH OBSERVATIONALLY CONSTRAINED HEATING FUNCTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Qiu Jiong; Liu Wenjuan; Longcope, Dana W. [Department of Physics, Montana State University, Bozeman, MT 59717-3840 (United States)

    2012-06-20

    We analyze high-cadence high-resolution observations of a C3.2 flare obtained by AIA/SDO on 2010 August 1. The flare is a long-duration event with soft X-ray and EUV radiation lasting for over 4 hr. Analysis suggests that magnetic reconnection and formation of new loops continue for more than 2 hr. Furthermore, the UV 1600 Angstrom-Sign observations show that each of the individual pixels at the feet of flare loops is brightened instantaneously with a timescale of a few minutes, and decays over a much longer timescale of more than 30 minutes. We use these spatially resolved UV light curves during the rise phase to construct empirical heating functions for individual flare loops, and model heating of coronal plasmas in these loops. The total coronal radiation of these flare loops are compared with soft X-ray and EUV radiation fluxes measured by GOES and AIA. This study presents a method to observationally infer heating functions in numerous flare loops that are formed and heated sequentially by reconnection throughout the flare, and provides a very useful constraint to coronal heating models.

  17. Classification of decays involving variable decay chains with convolutional architectures

    CERN Multimedia

    CERN. Geneva

    2018-01-01

    Vidyo contribution We present a technique to perform classification of decays that exhibit decay chains involving a variable number of particles, which include a broad class of $B$ meson decays sensitive to new physics. The utility of such decays as a probe of the Standard Model is dependent upon accurate determination of the decay rate, which is challenged by the combinatorial background arising in high-multiplicity decay modes. In our model, each particle in the decay event is represented as a fixed-dimensional vector of feature attributes, forming an $n \\times k$ representation of the event, where $n$ is the number of particles in the event and $k$ is the dimensionality of the feature vector. A convolutional architecture is used to capture dependencies between the embedded particle representations and perform the final classification. The proposed model performs outperforms standard machine learning approaches based on Monte Carlo studies across a range of variable final-state decays with the Belle II det...

  18. Role of activators and vacancies in the gamma-scintillation decay in CsI-Na

    International Nuclear Information System (INIS)

    Panova, A.N.; Grinev, B.V.; Sojfer, L.M.; Shakhova, K.V.; Kosinov, N.N.; Mitichkin, A.I.; Korsunova, S.P.; Lavrent'ev, F.F.

    2004-01-01

    We studied the influence of activator concentration (CNaI) and plastic deformation on the change in the contribution of the slow component to the decay of gamma-scintillation in CsI-Na crystals. The influence of CNaI on the change in the form of the luminescence excitation spectrum in the region of the absorption of activator centers (AC) and centers of vacancy nature (VNC) is investigated. The effect of CNaI on the change in the intercenter decay time of the mentioned centers is studied too. It is shown that AC and VNC participate in the photoluminescence and gamma-scintillations of CsI-Na crystals. In gamma-scintillations AC are responsible for the component τ i 370 ns, whereas the components τ 1 ' = 460 ns and τ 2 ∼ 2 μs are bound up with VNC. The decrease of τ γ from 770 to 560 ns with the growth of C from 2·10 -3 to 3·10 -2 mol. % NaI, and after plastic deformation of the crystals (ε = 5 %) along the axis from 570 to 470 ns is caused by the decrease in the number of VNC. Mechanisms of gamma-scintillations of CsI-Na crystals AC and VNC, as well as the decrease in the number of VNC are discussed. (authors)

  19. Heat and Mass Transfer at Hot Surface Ignition of Coal Particle

    OpenAIRE

    Glushkov Dmitrii O.; Kosintsev Andrey. G.; Shlegel Nikita E.; Vershinina Ksenia Yu.

    2015-01-01

    This paper describes the experimental investigations of the characteristics of heat and mass transfer during the conductive heating of a coal particle. We have established the boundary conditions of combustion initiation, and the conditions of thermal decomposition and solid fuel particles decay, characterized by the temperature of a heat source, and the duration of the respective stages.

  20. 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

  1. OH vibrational activation and decay dynamics of CH4-OH entrance channel complexes

    International Nuclear Information System (INIS)

    Wheeler, Martyn D.; Tsiouris, Maria; Lester, Marsha I.; Lendvay, Gyoergy

    2000-01-01

    Infrared spectroscopy has been utilized to examine the structure and vibrational decay dynamics of CH 4 -OH complexes that have been stabilized in the entrance channel to the CH 4 +OH hydrogen abstraction reaction. Rotationally resolved infrared spectra of the CH 4 -OH complexes have been obtained in the OH fundamental and overtone regions using an IR-UV (infrared-ultraviolet) double-resonance technique. Pure OH stretching bands have been identified at 3563.45(5) and 6961.98(4) cm-1 (origins), along with combination bands involving the simultaneous excitation of OH stretching and intermolecular bending motions. The infrared spectra exhibit extensive homogeneous broadening arising from the rapid decay of vibrationally activated CH 4 -OH complexes due to vibrational relaxation and/or reaction. Lifetimes of 38(5) and 25(3) ps for CH 4 -OH prepared with one and two quanta of OH excitation, respectively, have been extracted from the infrared spectra. The nascent distribution of the OH products from vibrational predissociation has been evaluated by ultraviolet probe laser-induced fluorescence measurements. The dominant inelastic decay channel involves the transfer of one quantum of OH stretch to the pentad of CH 4 vibrational states with energies near 3000 cm-1. The experimental findings are compared with full collision studies of vibrationally excited OH with CH 4 . In addition, ab initio electronic structure calculations have been carried out to elucidate the minimum energy configuration of the CH 4 -OH complex. The calculations predict a C 3v geometry with the hydrogen of OH pointing toward one of four equivalent faces of the CH 4 tetrahedron, consistent with the analysis of the experimental infrared spectra. (c) 2000 American Institute of Physics

  2. TMI-2 decay power: LASL fission-product and actinide decay power calculations for the President's commission on the accident at Three Mile Island

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.

    1980-03-01

    Fission-product and actinide decay heating, gas content, curies, and detailed contributions of the most important nuclide contributors were supplied in a series of letters following requests from the Presidential Commission on the Accident at Three Mile Island. In addition, similar data assuming different irradiation (power) histories were requested for purposes of comparison. This report consolidates the tabular and graphical data supplied and explains its basis

  3. Decay of the cosmological constant: Equivalence of quantum tunneling and thermal activation in two spacetime dimensions

    International Nuclear Information System (INIS)

    Gomberoff, Andres; Henneaux, Marc; Teitelboim, Claudio

    2005-01-01

    We study the decay of the cosmological constant in two spacetime dimensions through production of pairs. We show that the same nucleation process looks as quantum-mechanical tunneling (instanton) to one Killing observer and as thermal activation (thermalon) to another. Thus, we find another striking example of the deep interplay between gravity, thermodynamics and quantum mechanics which becomes apparent in presence of horizons

  4. Scintillating bolometers: A promising tool for rare decays search

    Energy Technology Data Exchange (ETDEWEB)

    Pattavina, L., E-mail: luca.pattavina@mib.infn.it

    2013-12-21

    The idea of using a scintillating bolometer was first suggested for solar neutrino experiments in 1989. After many years of developments, now we are able to exploit this experimental technique, based on the calorimetric approach with cryogenic particle detectors, to investigate rare events such as Neutrinoless Double Beta Decay and interaction of Dark Matter candidates. The possibility to have high resolution detectors in which a very large part of the natural background can be discriminated with respect to the weak expected signal is very appealing. The goal to distinguish the different types of interactions in the detector can be achieved by means of scintillating bolometer. The simultaneous read-out of the heat and scintillation signals made with two independent bolometers enable this precious feature leading to possible background free experiment. In the frame of the LUCIFER project we report on how exploiting this technique to investigate Double Beta Decay for different isotope candidates. Moreover we demonstrate how scintillating bolometers are suited for investigating other rare events such as α decays of long living isotopes of lead and bismuth.

  5. Proton decay theory

    International Nuclear Information System (INIS)

    Marciano, W.J.

    1983-01-01

    Topics include minimal SU(5) predictions, gauge boson mediated proton decay, uncertainties in tau/sub p/, Higgs scalar effects, proton decay via Higgs scalars, supersymmetric SU(5), dimension 5 operators and proton decay, and Higgs scalars and proton decay

  6. Beta-decay rate and beta-delayed neutron emission probability of improved gross theory

    Science.gov (United States)

    Koura, Hiroyuki

    2014-09-01

    A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for unmeasured nuclei are adopted from the KTUY nuclear mass formula, which is based on the spherical-basis method. Considering the properties of the integrated Fermi function, we can roughly categorized energy region of excited-state of a daughter nucleus into three regions: a highly-excited energy region, which fully affect a delayed neutron probability, a middle energy region, which is estimated to contribute the decay heat, and a region neighboring the ground-state, which determines the beta-decay rate. Some results will be given in the presentation. A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for

  7. Safety aspects of activation products in a compact Tokamak Fusion Power Plant

    International Nuclear Information System (INIS)

    Willenberg, H.J.; Bickford, W.E.

    1978-10-01

    Neutron activation of materials in a compact tokamak fusion reactor has been investigated. Results of activation product inventory, dose rate, and decay heat calculations in the blanket and injectors are presented for a reactor design with stainless steel structures. Routine transport of activated materials into the plasma and vacuum systems is discussed. Accidental release of radioactive materials as a result of liquid lithium spills is also considered

  8. A study of a small nuclear power plant system for district heating

    International Nuclear Information System (INIS)

    Imamura, Mitsuru; Sato, Kotaro; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

    2008-01-01

    We have studied nuclear power plant for district heating. Already some towns and villages in Hokkaido have requested small reactor for district heating. Using existing technology allows us to shorten development period and to keep a lid on development cost. We decided to develop new reactor based on 'MUTSU' reactor technology. 'MUTSU' had already proved its safety. And 'MUTSU' reactor was boron free reactor. It allows plant system to become more compact and simple. And load following capability by core reactivity become bigger. It means to reduce control rod movement. It leads to dependability enhancement. We calculated burn-up calculation of erbium addition fuel. In the result the core life became about 10 years. In the district heating system, there are not only district heating but also snow melting with warm water. It uses steam condenser's heat, which are only discharged now. This small plant has passive safety system. It is natural cooling of containment vessel. In case of loss of coolant accident, decay heat can remove by natural convection air cooling after 6 hours. Decay heat within 6 hours can remove by evaporative heat transfer of pool on containment vessel. (author)

  9. Analysis of natural convection in volumetrically-heated melt pools

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

    Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation. Refs, figs, tabs.

  10. Mitigation of radon and thoron decay products by filtration

    International Nuclear Information System (INIS)

    Wang Jin; Meisenberg, Oliver; Chen Yongheng; Karg, Erwin; Tschiersch, Jochen

    2011-01-01

    Inhalation of indoor radon ( 222 Rn) and thoron ( 220 Rn) decay products is the most important source of exposure to ionizing radiation for the human respiratory tract. Decreasing ventilation rates due to energy saving reasons in new buildings suggest additional active mitigation techniques to reduce the exposure in homes with high radon and thoron concentrations but poor ventilation. Filtration techniques with HEPA filters and simple surgical mask material have been tested for their potential to reduce the indoor exposure in terms of the total effective dose for mixed radon and thoron indoor atmospheres. The tests were performed inside an experimental room providing stable conditions. Filtration (at filtration rates of 0.2 h -1 and larger) removes attached radon and thoron decay products effectively but indoor aerosol as well. Therefore the concentration of unattached decay products (which have a higher dose coefficient) may increase. The decrease of the attached decay product concentrations could be theoretically described by a slowly decreasing exponential process. For attached radon decay products, it exhibited a faster but weaker removal process compared to attached thoron decay products (- 70% for attached radon decay products and - 80% for attached thoron decay products at a filtration rate of 0.5 h -1 with an HEPA filter). The concentration of unattached thoron decay products increased distinctly during the filtration process (+ 300%) while that of unattached radon decay products rose only slightly though at a much higher level (+ 17%). In the theoretical description these observed differences could be attributed to the different half-lives of the nuclides. Considering both effects, reduced attached and increased unattached decay product concentrations, filtration could significantly decrease the total effective dose from thoron whereas the overall effect on radon dose is small. A permanent filtration is recommended because of the slow decrease of the

  11. Effect of heat treatment on anodic activation of aluminium by trace element indium

    Energy Technology Data Exchange (ETDEWEB)

    Graver, Brit [Department of Materials Science and Engineering, Norwegian University of Science and Technology, N-7491 Trondheim (Norway); Helvoort, Antonius T.J. van [Department of Physics, Norwegian University of Science and Technology, N-7491 Trondheim (Norway); Nisancioglu, Kemal, E-mail: kemal.nisancioglu@material.ntnu.n [Department of Materials Science and Engineering, Norwegian University of Science and Technology, N-7491 Trondheim (Norway)

    2010-11-15

    Research highlights: {yields} Indium segregation activates AlIn alloy surface anodically in chloride solution. {yields} Enrichment of In on Al surface can occur thermally by heat treatment at 300 {sup o}C. {yields} Increasing temperature homogenises indium in aluminium reducing anodic activation. {yields} Indium can activate AlIn surface by segregating through dealloying of aluminium. {yields} Anodic activation is caused by AlIn amalgam formation at aluminium surface. - Abstract: The presence of trace elements in Group IIIA-VA is known to activate aluminium anodically in chloride environment. The purpose of this paper is to investigate the surface segregation of trace element In by heat treatment and resulting surface activation. Model binary AlIn alloys, containing 20 and 1000 ppm by weight of In, were characterized after heat treatment at various temperatures by use of glow discharge optical emission spectroscopy, electron microscopy and electrochemical polarization. Heat treatment for 1 h at 300 {sup o}C gave significant segregation of discrete In particles (thermal segregation), which activated the surface. Indium in solid solution with aluminium, obtained by 1 h heat treatment at 600 {sup o}C, also activated by surface segregation of In on alloy containing 1000 ppm In, resulting from the selective dissolution of the aluminium component during anodic oxidation (anodic segregation). The effect of anodic segregation was reduced by decreasing indium concentration in solid solution; it had negligible effect at the 20 ppm level. The segregated particles were thought to form a liquid phase alloy with aluminium during anodic polarization, which in turn, together with the chloride in the solution destabilized the oxide.

  12. β-decay properties in the Cs decay chain

    Science.gov (United States)

    Benzoni, G.; Lică, R.; Borge, M. J. G.; Fraile, L. M.; IDS Collaboration

    2018-02-01

    The study of the decay of neutron-rich Cs isotopes has two main objectives: on one side β decay is a perfect tool to access the low-spin structures in the daughter Ba nuclei, where the evolution of octupole deformed shapes can be followed, while, on the other hand, the study of the gross properties of these decays, in terms of decay rates and branching to delayed-neutron emission, are fundamental inputs for the modelling of the r-process in the Rare-Earth Elements peak. Results obtained at CERN-ISOLDE are discussed within this framework and compared to existing data and predictions from state-of-the-art nuclear models.

  13. Effect of Activation Temperature and Heating Duration on Physical Characteristics of Activated Carbon Prepared from Agriculture Waste

    Directory of Open Access Journals (Sweden)

    Tham Yee Jun

    2010-01-01

    Full Text Available This study was conducted to determine the physical characteristics of activated carbon prepared from durian shell in varied heating durations from 10 min to 30 min and activation temperatures of 400C and 500C. Durian shells have been characterized in term of ultimate and proximate analysis, chemical composition and thermal behaviour with a view to be used as activated carbon precursor. Durian shell activated carbon was prepared by impregnating 10g of sample in 10% (v/v concentration of phosphoric acid for 24 h, followed by carbonization at 400C and 500C with different heating durations under nitrogen atmosphere. The results showed that various treatment conditions affect the percentage of yield, BET surface area, micropore volume, and average pore diameter. The highest surface area (SBET 1024 m2/g was obtained at 500C and 20 min of heating duration with 63% of yield and 0.21 cm3/g micropore volume.

  14. Electrochemical activation, voltage decay and hysteresis of Li-rich layered cathode probed by various cobalt content

    KAUST Repository

    Wu, Yingqiang

    2018-02-01

    The high capacity of Li-rich layered cathode materials have attracted great attention for the greater energy density lithium ion (Li-ion) batteries, but the understanding of knowledge associated with electrochemical behaviours are still needed to improve their performances further. In this study, different amount of Co content is designed in Li-rich layered compounds (0.5Li2MnO3·0.5LiMn0.5-xNi0.5-xCo2xO2, 0 ≤ x ≤ 0.2), and the stepwise electrochemical activation process is applied to explore the features. We discover that the substitution of Co3+ ions can accelerate the electrochemical activation of Li2MnO3 component, and the Co-doped compound delivers much higher capacities even they suffer an apparent voltage decay comparing to the Co-free one. Besides, a fast metal ions migration exists (e.g., from the metastable tetrahedral site to the lower energy cubic site) in initial dozens of cycles (e.g., 30 cycles at 0.1C); thereafter, they likely return to the original octahedral site, as demonstrated in the voltage decay and hysteresis analysis.

  15. Safety characteristics of small heat producing reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1987-10-01

    The primary objectives of protection in nuclear power plants are the possibility to shut the reactor down in case of emergency and keep it subcritical in the long run, the existence of a heat sink for post-decay heat removal in order to avoid overheating, let alone core meltdown, and the containment of radioactivity within the barriers designed for this purpose, thus preventing significant activity release. In principle, these objectives can be met in various ways, namely by active, passive or inherent technical safeguards systems. In practice, a mixture of these approaches is employed in almost all cases. What matters in the end is the assessment of the overall concept, not of some outstanding feature. Inherent characteristics are easier to achieve in small reactors. However, also in this case, inherent safety does not mean absolute safety. If inherent safety characteristics were all encompassing, they would have to include self-healing effects. However, inanimate matter is incapable of such self-organization. Consequently, inherent characteristics in nuclear technology by definition should include the increased use of dissipative processes in the thermal part of the plant. (author)

  16. Molecular dynamics calculation of half-lives for thermal decay of Lennard-Jones clusters

    International Nuclear Information System (INIS)

    Smith, R.W.

    1991-01-01

    Molecular dynamics has been used with a Lenard-Jones (6-12) potential in order to study the decay behavior of neutral Argon clusters containing between 12 and 14 atoms. The clusters were heated to temperatures well above their melting points and then tracked in time via molecular dynamics until evaporation of one or more atoms was observed. In each simulation, the mode of evaporation, energy released during evaporation, and cluster lifetime were recorded. Results from roughly 2000 simulation histories were combined in order to compute statistically significant values of cluster half-lives and decay energies. It was found that cluster half-life decreases with increasing energy and that for a given value of excess energy (defined as E=(E tot -E gnd )/n), the 13 atom cluster is more stable against decay than clusters containing either 12 or 14 atoms. The dominant decay mechanism for all clusters was determined to be single atom emission. (orig.)

  17. The Chemistry of Self-Heating Food Products: An Activity for Classroom Engagement

    Science.gov (United States)

    Oliver-Hoyo, Maria T.; Pinto, Gabriel; Llorens-Molina, Juan Antonio

    2009-01-01

    Two commercial self-heating food products have been used to apply chemical concepts such as stoichiometry, enthalpies of reactions and solutions, and heat transfer in a classroom activity. These products are the self-heating beverages sold in Europe and the Meals, Ready to Eat or MREs used primarily by the military in the United States. The main…

  18. Heat production in granitic rocks

    DEFF Research Database (Denmark)

    Artemieva, Irina; Thybo, Hans; Jakobsen, Kiki

    2017-01-01

    Granitic rocks play special role in the dynamics and evolution of the Earth and its thermal regime. First, their compositional variability, reflected in the distribution of concentrations of radiogenic elements, provides constraints on global differentiation processes and large scale planetary...... evolution, where emplacement of granites is considered a particularly important process for the formation of continental crust. Second, heat production by radioactive decay is among the main heat sources in the Earth. Therefore knowledge of heat production in granitic rocks is pivotal for thermal modelling...... of the continental lithosphere, given that most radiogenic elements are concentrated in granitic rocks of the upper continental crust whereas heat production in rocks of the lower crust and lithospheric mantle is negligible. We present and analyze a new global database GRANITE2017 (with about 500 entries...

  19. Heat shock modulates the subcellular localization, stability, and activity of HIPK2

    International Nuclear Information System (INIS)

    Upadhyay, Mamta; Bhadauriya, Pratibha; Ganesh, Subramaniam

    2016-01-01

    The homeodomain-interacting protein kinase-2 (HIPK2) is a highly conserved serine/threonine kinase and is involved in transcriptional regulation. HIPK2 is a highly unstable protein, and is kept at a low level under normal physiological conditions. However, exposure of cells to physiological stress – such as hypoxia, oxidative stress, or UV damage – is known to stabilize HIPK2, leading to the HIPK2-dependent activation of p53 and the cell death pathway. Therefore HIPK2 is also known as a stress kinase and as a stress-activated pro-apoptotic factor. We demonstrate here that exposure of cells to heat shock results in the stabilization of HIPK2 and the stabilization is mediated via K63-linked ubiquitination. Intriguingly, a sub-lethal heat shock (42 °C, 1 h) results in the cytoplasmic localization of HIPK2, while a lethal heat shock (45 °C, 1 h) results in its nuclear localization. Cells exposed to the lethal heat shock showed significantly higher levels of the p53 activity than those exposed to the sub-lethal thermal stress, suggesting that both the level and the nuclear localization are essential for the pro-apoptotic activity of HIPK2 and that the lethal heat shock could retain the HIPK2 in the nucleus to promote the cell death. Taken together our study underscores the importance of HIPK2 in stress mediated cell death, and that the HIPK2 is a generic stress kinase that gets activated by diverse set of physiological stressors.

  20. Heat shock modulates the subcellular localization, stability, and activity of HIPK2

    Energy Technology Data Exchange (ETDEWEB)

    Upadhyay, Mamta; Bhadauriya, Pratibha; Ganesh, Subramaniam, E-mail: sganesh@iitk.ac.in

    2016-04-15

    The homeodomain-interacting protein kinase-2 (HIPK2) is a highly conserved serine/threonine kinase and is involved in transcriptional regulation. HIPK2 is a highly unstable protein, and is kept at a low level under normal physiological conditions. However, exposure of cells to physiological stress – such as hypoxia, oxidative stress, or UV damage – is known to stabilize HIPK2, leading to the HIPK2-dependent activation of p53 and the cell death pathway. Therefore HIPK2 is also known as a stress kinase and as a stress-activated pro-apoptotic factor. We demonstrate here that exposure of cells to heat shock results in the stabilization of HIPK2 and the stabilization is mediated via K63-linked ubiquitination. Intriguingly, a sub-lethal heat shock (42 °C, 1 h) results in the cytoplasmic localization of HIPK2, while a lethal heat shock (45 °C, 1 h) results in its nuclear localization. Cells exposed to the lethal heat shock showed significantly higher levels of the p53 activity than those exposed to the sub-lethal thermal stress, suggesting that both the level and the nuclear localization are essential for the pro-apoptotic activity of HIPK2 and that the lethal heat shock could retain the HIPK2 in the nucleus to promote the cell death. Taken together our study underscores the importance of HIPK2 in stress mediated cell death, and that the HIPK2 is a generic stress kinase that gets activated by diverse set of physiological stressors.

  1. Deposition of radon decay products on the skin of balneotherapy patients in Gastein

    International Nuclear Information System (INIS)

    Tempfer, H.; Schober, A.; Lettner, H.; Hofmann, W.; Foisner, W.; Steger, F.

    2001-01-01

    In Bad Gastein and Bad Hofgastein, radon-rich thermal water is used for the treatment of various diseases, especially for Morbus Bechterew, a degenerative illness of the spinal column. One of the therapeutic applications of the thermal water is the exposition of patients in 450 1 bathtubs (''Best'sche Wanne''). To check the hypothesis that the radon decay products contribute to the therapeutic effect, the adsorption behaviour of radon decay products was examined. A group of patients was exposed to the thermal water for 20 minutes. Immediately after leaving the bathtub, the activities were measured at several locations of the human body (forearms, belly, lower legs) by alpha spectrometry. The activities of the decay products 218 Po, 214 Pb and 214 Bi/ 214 Po were determined from the measured spectra and decay curves. In order to study the time dependence of the decay product adsorption, one test person was exposed for 10, 20, to 30, 40 and 60 minutes; afterwards the decay curves were recorded again over a period of 30 minutes. The results showed that the activities of the decay products on the skin can differ appreciably. On the one hand, the distribution on the body surface seems to be quite inhomogeneous, on the other hand, the deposition varies considerably from person to person. As a result of this, skin activities varied between 1.2 and 4.1 Bq/cm 2 . The exposition and measurement of copper and PVC plates confirmed earlier results, that the deposition of radon decay products in water on artificial surfaces is much less than that on living surfaces. In conclusion, the remarkably high decay product activities on the skin does indeed suggest a positive correlation with the observed therapeutic effects. (orig.) [de

  2. Pre-formation of α-clusters in natural α-decay

    International Nuclear Information System (INIS)

    Basu, M.K.

    1977-01-01

    A recent investigation into the fundamental mechanism of α-decay with an α-cluster as the starting point has clearly revealed that α-formation and α-emission is one and the same process and that tunneling is not necessary for α-decay. This α-decay without tunneling (ADWT) study provides reliable expressions for the α-decay radius and for the absolute α-preformation probability Pα. ADWT values of Pα for even-even nuclei are qualitatively a decreasing function of the masses of nuclei and quantitatively lie in the range 0.6-0.08 in very good agreement with the experimental values. For naturally α-active nuclei, Pα lies in the range 0.17-0.08 and exhibits shell-effect and deformation effect in the form of a distinct shell-closure at N=126 and enhanced values for nuclei off a shell-closure region, respectively. This is also in accord with experimental observations. The observed trend of Pα values is rather puzzling in that heavy nuclei with low Pαs are spontaneously α-active whereas light and intermediate nuclei with high Pαs are not. A further investigation into the role of α-clusters in the intra-nuclear mechanism of α-formation and α-emission marks out, on the one hand, Coulomb repulsion as the cause of α-decay rather than as a barrier to it, and, yields, on the other hand, a necessary and sufficient condition for spontaneous α-activity of a nucleus. This condition clearly explains why low Pαs and not high Pαs favour natural α-activity. (author)

  3. Partial radiogenic heat model for Earth revealed by geoneutrino measurements

    NARCIS (Netherlands)

    Abe, S.; et al., [Unknown; Decowski, M.P.

    2011-01-01

    The Earth has cooled since its formation, yet the decay of radiogenic isotopes, and in particular uranium, thorium and potassium, in the planet’s interior provides a continuing heat source. The current total heat flux from the Earth to space is 44.2±1.0 TW, but the relative contributions from

  4. 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

  5. Effect of heat treatment on brewer's yeast fermentation activity

    OpenAIRE

    Kharandiuk, Tetiana; Kosiv, Ruslana; Palianytsia, Liubov; Berezovska, Natalia

    2015-01-01

    The influence of temperature treatment of brewer's yeast strain Saflager W-34/70 at temperatures of -17, 20, 25, 30, 35, 40 °C on their fermentative activity was studied. It was established that the freezing of yeast leads to a decrease of fermentation activity in directly proportional to the duration way. Fermentative activity of yeast samples can be increased by 20-24% by heat treatment at 35 °C during 15-30 minutes.

  6. Defect production at exciton decay in ionic crystals

    International Nuclear Information System (INIS)

    Lushchik, Ch.B.

    1984-01-01

    On the example of alkali halide crystals experimentally detected phenomenon of structural point defect production in wide-gap nonmetallic solids at low-temperature radiationless decay of self-localizing excitons and recombination of electrons with self-localized holes is considered. Factors promoting radiationless transformation of electron excitations to not small oscillations of many atoms (heat release), but to separate ion large shifts, that determine one of the most important mechanisms of radiation instability of solids, used, in particular, for data recording, are discussed

  7. Three-wave interaction during electron cyclotron resonance heating and current drive

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Jacobsen, Asger Schou; Hansen, Søren Kjer

    2016-01-01

    Non-linear wave-wave interactions in fusion plasmas, such as the parametric decay instability (PDI) of gyrotron radiation, can potentially hamper the use of microwave diagnostics. Here we report on anomalous scattering in the ASDEX Upgrade tokamak during electron cyclotron resonance heating...... experiments. The observations can be linked to parametric decay of the gyrotron radiation at the second harmonic upper hybrid resonance layer....

  8. Ion cyclotron resonance heating

    International Nuclear Information System (INIS)

    Tajima, T.

    1982-01-01

    Ion cyclotron resonance heating of plasmas in tokamak and EBT configurations has been studied using 1-2/2 and 2-1/2 dimensional fully self-consistent electromagnetic particle codes. We have tested two major antenna configurations; we have also compared heating efficiencies for one and two ion species plasmas. We model a tokamak plasma with a uniform poloidal field and 1/R toroidal field on a particular q surface. Ion cyclotron waves are excited on the low field side by antennas parallel either to the poloidal direction or to the toroidal direction with different phase velocities. In 2D, minority ion heating (vsub(perpendicular)) and electron heating (vsub(parallel),vsub(perpendicular)) are observed. The exponential electron heating seems due to the decay instability. The minority heating is consistent with mode conversion of fast Alfven waves and heating by electrostatic ion cyclotron modes. Minority heating is stronger with a poloidal antenna. The strong electron heating is accompanied by toroidal current generation. In 1D, no thermal instability was observed and only strong minority heating resulted. For an EBT plasma we model it by a multiple mirror. We have tested heating efficiency with various minority concentrations, temperatures, mirror ratios, and phase velocities. In this geometry we have beach or inverse beach heating associated with the mode conversion layer perpendicular to the toroidal field. No appreciable electron heating is observed. Heating of ions is linear in time. For both tokamak and EBT slight majority heating above the collisional rate is observed due to the second harmonic heating. (author)

  9. In-trap decay spectroscopy for {beta}{beta} decays

    Energy Technology Data Exchange (ETDEWEB)

    Brunner, Thomas

    2011-01-18

    The presented work describes the implementation of a new technique to measure electron-capture (EC) branching ratios (BRs) of intermediate nuclei in {beta}{beta} decays. This technique has been developed at TRIUMF in Vancouver, Canada. It facilitates one of TRIUMF's Ion Traps for Atomic and Nuclear science (TITAN), the Electron Beam Ion Trap (EBIT) that is used as a spectroscopy Penning trap. Radioactive ions, produced at the radioactive isotope facility ISAC, are injected and stored in the spectroscopy Penning trap while their decays are observed. A key feature of this technique is the use of a strong magnetic field, required for trapping. It radially confines electrons from {beta} decays along the trap axis while X-rays, following an EC, are emitted isotropically. This provides spatial separation of X-ray and {beta} detection with almost no {beta}-induced background at the X-ray detector, allowing weak EC branches to be measured. Furthermore, the combination of several traps allows one to isobarically clean the sample prior to the in-trap decay spectroscopy measurement. This technique has been developed to measure ECBRs of transition nuclei in {beta}{beta} decays. Detailed knowledge of these electron capture branches is crucial for a better understanding of the underlying nuclear physics in {beta}{beta} decays. These branches are typically of the order of 10{sup -5} and therefore difficult to measure. Conventional measurements suffer from isobaric contamination and a dominating {beta} background at theX-ray detector. Additionally, X-rays are attenuated by the material where the radioactive sample is implanted. To overcome these limitations, the technique of in-trap decay spectroscopy has been developed. In this work, the EBIT was connected to the TITAN beam line and has been commissioned. Using the developed beam diagnostics, ions were injected into the Penning trap and systematic studies on injection and storage optimization were performed. Furthermore, Ge

  10. In-trap decay spectroscopy for ββ decays

    International Nuclear Information System (INIS)

    Brunner, Thomas

    2011-01-01

    The presented work describes the implementation of a new technique to measure electron-capture (EC) branching ratios (BRs) of intermediate nuclei in ββ decays. This technique has been developed at TRIUMF in Vancouver, Canada. It facilitates one of TRIUMF's Ion Traps for Atomic and Nuclear science (TITAN), the Electron Beam Ion Trap (EBIT) that is used as a spectroscopy Penning trap. Radioactive ions, produced at the radioactive isotope facility ISAC, are injected and stored in the spectroscopy Penning trap while their decays are observed. A key feature of this technique is the use of a strong magnetic field, required for trapping. It radially confines electrons from β decays along the trap axis while X-rays, following an EC, are emitted isotropically. This provides spatial separation of X-ray and β detection with almost no β-induced background at the X-ray detector, allowing weak EC branches to be measured. Furthermore, the combination of several traps allows one to isobarically clean the sample prior to the in-trap decay spectroscopy measurement. This technique has been developed to measure ECBRs of transition nuclei in ββ decays. Detailed knowledge of these electron capture branches is crucial for a better understanding of the underlying nuclear physics in ββ decays. These branches are typically of the order of 10 -5 and therefore difficult to measure. Conventional measurements suffer from isobaric contamination and a dominating β background at theX-ray detector. Additionally, X-rays are attenuated by the material where the radioactive sample is implanted. To overcome these limitations, the technique of in-trap decay spectroscopy has been developed. In this work, the EBIT was connected to the TITAN beam line and has been commissioned. Using the developed beam diagnostics, ions were injected into the Penning trap and systematic studies on injection and storage optimization were performed. Furthermore, Ge detectors, for the detection of X-rays, were

  11. Decay tank

    International Nuclear Information System (INIS)

    Matsumura, Seiichi; Tagishi, Akinori; Sakata, Yuji; Kontani, Koji; Sudo, Yukio; Kaminaga, Masanori; Kameyama, Iwao; Ando, Koei; Ishiki, Masahiko.

    1990-01-01

    The present invention concerns an decay tank for decaying a radioactivity concentration of a fluid containing radioactive material. The inside of an decay tank body is partitioned by partitioning plates to form a flow channel. A porous plate is attached at the portion above the end of the partitioning plate, that is, a portion where the flow is just turned. A part of the porous plate has a slit-like opening on the side close to the partitioning plate, that is, the inner side of the flow at the turning portion thereof. Accordingly, the primary coolants passed through the pool type nuclear reactor and flown into the decay tank are flow caused to uniformly over the entire part of the tank without causing swirling. Since a distribution in a staying time is thus decreased, the effect of decaying 16 N as radioactive nuclides in the primary coolants is increased even in a limited volume of the tank. (I.N.)

  12. Heat pipe heat exchangers in heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Stulc, P; Vasiliev, L L; Kiseljev, V G; Matvejev, Ju N

    1985-01-01

    The results of combined research and development activities of the National Research Institute for Machine Design, Prague, C.S.S.R. and the Institute for Heat and Mass Transfer, Minsk, U.S.S.R. concerning intensification heat pipes used in heat pipe heat exchangers are presented. This sort of research has been occasioned by increased interest in heat power economy trying to utilise waste heat produced by various technological processes. The developed heat pipes are deployed in construction of air-air, gas-air or gas-gas heat recovery exchangers in the field of air-engineering and air-conditioning. (author).

  13. Design to nullify activity movement in heat transport systems

    International Nuclear Information System (INIS)

    Hemmings, R.L.; Barber, D.

    1975-01-01

    This article describes the methods by which designers can reduce the adverse effects of system corrosion and the resultant activation of the corrosion products in heat transport systems. The presentation will cover: a) choice of materials; b) assessment of the need of components; c) control of system chemistry; d) factors considered in sizing HTS purification systems; i) control of activation and fission products; ii) decontamination. (author)

  14. Decay uncovered in nonverbal short-term memory.

    Science.gov (United States)

    Mercer, Tom; McKeown, Denis

    2014-02-01

    Decay theory posits that memory traces gradually fade away over the passage of time unless they are actively rehearsed. Much recent work exploring verbal short-term memory has challenged this theory, but there does appear to be evidence for trace decay in nonverbal auditory short-term memory. Numerous discrimination studies have reported a performance decline as the interval separating two tones is increased, consistent with a decay process. However, most of this tone comparison research can be explained in other ways, without reference to decay, and these alternative accounts were tested in the present study. In Experiment 1, signals were employed toward the end of extended retention intervals to ensure that listeners were alert to the presence and frequency content of the memoranda. In Experiment 2, a mask stimulus was employed in an attempt to distinguish between a highly detailed sensory trace and a longer-lasting short-term memory, and the distinctiveness of the stimuli was varied. Despite these precautions, slow-acting trace decay was observed. It therefore appears that the mere passage of time can lead to forgetting in some forms of short-term memory.

  15. Hadronic decay of late-decaying particles and big-bang nucleosynthesis

    Energy Technology Data Exchange (ETDEWEB)

    Kawasaki, Masahiro [Research Center for the Early Universe, Graduate School of Science, University of Tokyo, Tokyo 113-0033 (Japan)]. E-mail: masahiro_kawasaki@mac.com; Kohri, Kazunori [Department of Earth and Space Science, Osaka University, Osaka 560-0043 (Japan); Moroi, Takeo [Department of Physics, Tohoku University, Sendai 980-8578 (Japan)

    2005-10-06

    We study the big-bang nucleosynthesis (BBN) scenario with late-decaying exotic particles with lifetime longer than {approx}1 s. With a late-decaying particle in the early universe, predictions of the standard BBN scenario can be significantly altered. Therefore, we derive constraints on its primordial abundance. We pay particular attention to hadronic decay modes of such particles. We see that the non-thermal production process of D, {sup 3}He and {sup 6}Li provides a stringent upper bound on the primordial abundance of late-decaying particles with hadronic branching ratio.

  16. 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

  17. d-3He reaction measurements during fast wave minority heating in PLT

    International Nuclear Information System (INIS)

    Chrien, R.E.; Strachan, J.D.

    1983-01-01

    Time- and energy-resolved d- 3 He fusion reactions have been measured to infer the energy of the d + or He ++ minority ions heated near their cyclotron frequency by the magnetosonic fast wave. The average energy of the reacting 3 He ions during 3 He minority heating is in the range of 100 to 400 keV, as deduced from the magnitude of the reaction rate, its decay time, and the energy spread of the proton reaction products. The observed reaction rate and its scaling with wave power and electron density and temperature are in qualitative agreement with a radial reaction rate model using the minority distribution predicted from quasilinear velocity space diffusion. Oscillations in the reaction rate are observed concurrent with sawtooth and m = 2 MHD activity in the plasma

  18. Optical absorption and electron spin resonance in natural, irradiated and heated spodumene

    International Nuclear Information System (INIS)

    Ito, A.S.; Isotani, S.

    1983-09-01

    Heat treatment and X and γ-Rays irradiation of lylac and colorless natural spodumene, LiAlSi 2 O 6 , have been studied. Irradiation produces a color change, from lylac or colorless to green. Irradiated samples heated at 200 0 C turn lylac and bleach at 400 0 C. Optical absorption spectra were decomposed into gaussian line shape bands and it is observed that green and lylac centers are simultaneously created by irradiation. These centers are independent from each other. Optical absorption, EPR and X-Ray fluorescence results show the spectra of impurities, responsible by the presence of the created centers and indicated Mn participation in the process of centers creation and destruction. Decay Kinetics of green and lylac centers have been studied and it is observed that simple Kinetic models do not apply to these cases. An empirical fit allowed the calculation of activation energies of the lylac and green centers decays. Discussions about the present results led us to propose a new model where the green and lylac centers are due to Mn 3+ ions in two different Al 3+ sites, whose absorption are intensified by the interaction with an electron trapped in a neighbour oxygen. (Author) [pt

  19. Development of Kabila rocket: A radioisotope heated thermionic plasma rocket engine

    Directory of Open Access Journals (Sweden)

    Kalomba Mboyi

    2015-04-01

    Full Text Available A new type of plasma rocket engine, the Kabila rocket, using a radioisotope heated thermionic heating chamber instead of a conventional combustion chamber or catalyst bed is introduced and it achieves specific impulses similar to the ones of conventional solid and bipropellant rockets. Curium-244 is chosen as a radioisotope heat source and a thermal reductive layer is also used to obtain precise thermionic emissions. The self-sufficiency principle is applied by simultaneously heating up the emitting material with the radioisotope decay heat and by powering the different valves of the plasma rocket engine with the same radioisotope decay heat using a radioisotope thermoelectric generator. This rocket engine is then benchmarked against a 1 N hydrazine thruster configuration operated on one of the Pleiades-HR-1 constellation spacecraft. A maximal specific impulse and power saving of respectively 529 s and 32% are achieved with helium as propellant. Its advantages are its power saving capability, high specific impulses and simultaneous ease of storage and restart. It can however be extremely voluminous and potentially hazardous. The Kabila rocket is found to bring great benefits to the existing spacecraft and further research should optimize its geometric characteristics and investigate the physical principals of its operation.

  20. Estimation of delayed neutron emission probability by using the gross theory of nuclear β-decay

    International Nuclear Information System (INIS)

    Tachibana, Takahiro

    1999-01-01

    The delayed neutron emission probabilities (P n -values) of fission products are necessary in the study of reactor physics; e.g. in the calculation of total delayed neutron yields and in the summation calculation of decay heat. In this report, the P n -values estimated by the gross theory for some fission products are compared with experiment, and it is found that, on the average, the semi-gross theory somewhat underestimates the experimental P n -values. A modification of the β-decay strength function is briefly discussed to get more reasonable P n -values. (author)

  1. Brain activity and fatigue during prolonged exercise in the heat

    DEFF Research Database (Denmark)

    Nielsen, Bodil; Hyldig, Tino; Bidstrup, F.

    2001-01-01

    We hypothesized that fatigue due to hyperthermia during prolonged exercise in the heat is in part related to alterations in frontal cortical brain activity. The electroencephalographic activity (EEG) of the frontal cortex of the brain was measured in seven cyclists [maximal O2 uptake (VO2max) 4...... min of exercise; P

  2. Rare decays and CP asymmetries in charged B decays

    International Nuclear Information System (INIS)

    Deshpande, N.G.

    1991-01-01

    The theory of loop induced rare decays and the rate asymmetry due to CP violation in charged B Decays in reviewed. After considering b → sγ and b → se + e - decays, the asymmetries for pure penguin process are estimated first. A larger asymmetry can result in those modes where a tree diagram and a penguin diagram interfere, however these estimates are necessarily model dependent. Estimates of Cabbibo suppressed penguins are also considered

  3. Absolute transition intensities in decay of 92Rb

    International Nuclear Information System (INIS)

    Lhersonneau, G.; Rizzi, V.; Barzakh, A.

    2005-01-01

    In a first step studied the evolution of the ion current of 92 Rb and 94 Sr produced by surface ionisation versus temperature. Ratios of peak areas of γ lines emitted in Sr (1428 keV) and Rb (837 keV) decays are shown. The temperature is decreased from working conditions until the ion currents become very weak. The plateau at low temperature can be understood by the fact that Sr is not ionised anymore and there is no more decease possible. Then, the only contribution to the Sr decay peak is from the decay of Rb collected before transport of the activity to the counting station

  4. Heat transfer analysis to investigate the core catcher plate assembly in SFR

    International Nuclear Information System (INIS)

    Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

  5. Status report on preliminary design activities for solar heating and cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    Information presented provides status and progress on the development of solar heating and cooling systems. The major emphasis is placed on program organization, system size definition, site identification, system approaches, heat pump and equipment design, collector procurement, and other preliminary design activities as part of the contract requirements.

  6. Radioactive Decay

    Science.gov (United States)

    Radioactive decay is the emission of energy in the form of ionizing radiation. Example decay chains illustrate how radioactive atoms can go through many transformations as they become stable and no longer radioactive.

  7. Long-lived radicals produced by γ-irradiation or vital activity in plants, animals, cells, and protein solution: their observation and inhomogeneous decay dynamics

    International Nuclear Information System (INIS)

    Miyazaki, Tetsuo; Morikawa, Akiyuki; Kumagai, Jun; Ikehata, Masateru; Koana, Takao; Kikuchi, Shoshi

    2002-01-01

    Long-lived radicals produced by γ-irradiation or vital activity in plants, animals, cells, and protein (albumin) solution were studied by electron spin resonance spectroscopy. Long-lived radicals produced by vital activity exist in biological systems, such as plants, animals, and cells, in the range of 0.1-20 nmol g -1 . Since vital organs keep the radicals at a constant concentration, the radicals are probably related to life conservation. Long-lived radicals are also produced by γ-irradiation of cells or protein solution. The radicals decay after death of living things or after γ-irradiation. We found that the decay dynamics in all biological systems can be expressed by the same kinetic equation of an inhomogeneous reaction

  8. Heat wave experiments on the W7-AS stellarator

    International Nuclear Information System (INIS)

    Hartfuss, H.J.; Erckmann, V.; Gasparino, U.; Giannone, L.; Maassberg, H.; Tutter, M.

    1993-01-01

    Power modulation with well localized ECRH power deposition at both 70 and 140 GHz, has been used to generate temperature perturbations which propagate away from the deposition region. Radiometry of the ECE is used to diagnose the generated temperature perturbation as a function of distance to the deposition zone. The decay of the amplitude and the delay of the wave provide the information to determine the electron thermal diffusivity. This value is then compared with the one derived from a global power balance. It is found that both values agree with the error bars. The technique has also been applied in recent experiments during L-H-mode transitions in W7-AS demonstrating a significant reduction in the effective heat diffusivity in the plasma core during the H-phase. The modulated ECRH causes a modulation of the Shafranov shift. Interference of the prompt shift with the heat wave results in an apparent asymmetry of the decay length of the heat wave with respect to the plasma centre. (orig.)

  9. Natural convection heat transfer in SIGMA experiment

    International Nuclear Information System (INIS)

    Lee, Seung Dong; Lee, Gang Hee; Suh, Kune Yull

    2004-01-01

    A loss-of-coolant accident (LOCA) results in core melt formation and relocation at various locations within the reactor core over a considerable period of time. If there is no effective cooling mechanism, the core debris may heat up and commence natural circulation. The high temperature pool of molten core material will threaten the structural integrity of the reactor vessel. The extent and urgency of this threat depend primarily upon the intensity of the internal heat sources and upon the consequent distribution of the heat fluxes on the vessel walls in contact with the molten core material pools. In such a steady molten pool convection state, the thermal loads against the vessel would be determined by the in-vessel heat transfer distribution involving convective and conductive heat transfer from the decay-heated core material pool to the lower head wall in contact with the core material. In this study, upward and downward heat transfer fraction ratio is focused on

  10. Beta-gamma spectroscopy for double beta decays and Lepton number conservation

    International Nuclear Information System (INIS)

    Ejiri, H.; Takahashi, N.; Shibata, T.; Nagai, Y.; Okada, K.; Kamikubota, N.; Watanabe, T.

    1984-01-01

    In this paper neutrino-less double β decays (Oν ββ) of /sup 76/Ge were studied by means of the newly developed ELEGANTS (Electron gamma-ray neutrino spectrometer). It consists of a 171 cc pure Ge detector surrounded bu a big 4π-NaI detector, and active and inactive filters. Measurement of both the electron signal from the Ge detector and γ-ray signals from the 4π-NaI detector made it possible to select the true double decay events from background events due to the other radio-active isotopes and cosmic rays. The ELEGANTS showed the highest sensitivity for detecting the neutrino-less double β decay. The preliminary data obtained so far give a lower limit of the half life T/sub 1/2/≥2.2 10/sup 22/y for the O/sup +/→O/sup +/ Oν ββ decay and T/sub 1/2/ ≥1.5.10/sup 22/y for the O/sup +/→2/sup +/ Oν ββ decay of /sup 76/Ge

  11. The Monte-Carlo code DECAY to simulate the decay of baryon and meson resonances

    International Nuclear Information System (INIS)

    Haenssgen, K.; Ritter, S.

    1983-01-01

    The code DECAY simulates the decay of unpolarized baryon and meson resonances in the laboratory frame. DECAY treats some resonances among these all baryon resonances of the spin 3/2 + decuplet and all meson resonances of the spin 1 - nonet. A given resonance decays via two or three particle decay steps until all decay products are stable particles. Program summary and code description are given. (author)

  12. Decay of hypernuclei

    International Nuclear Information System (INIS)

    Bando, H.

    1985-01-01

    The pionic and non-mesonic decays of hypernuclei are discussed. In the first part, various decay processes which could be useful to obtain information of hypernuclear structure are discussed. The experimental data concerning the pionic and non-mesonic decays are discussed in the second part. As the experimental data, there are only few lifetime data and some crude data on the non-mesonic to π decay ratio. In the third and the fourth parts, some theoretical analyses are made on the pionic and the nonmesonic decays. DDHF calculation was performed for Λ and N systems by using Skyrme type ΛN and NN effective interactions. A suppression factor of the order of 10 -3 for A nearly equal 100 was obtained. (Aoki, K.)

  13. Radon and its decay product activities in the magmatic area and the adjacent volcano-sedimentary Intrasudetic Basin

    Directory of Open Access Journals (Sweden)

    D. Tchorz

    2007-06-01

    Full Text Available In the magmatic area of Sudetes covering the Karkonosze granite and adjacent volcano-sedimentary Intrasudetic Basin a study of atmospheric radon activity was performed by means of SSNTD Kodak LR-115. The study was completed by gamma spectrometric survey of eU and eTh determined by gamma activity of radon decay products 214Bi and 208Tl respectively. In the case of the western part of the Karkonosze granite area the radon decay products activity in the granitic basement was found to be as high as 343 Bq/kg for 214Bi and 496 Bq/kg for 208Tl respectively. Atmospheric radon content measured by means of Kodak LR115 track detector at the height of 1.5 m was found as high as 70 Bq/m3 in the regions, where no mining activities took place. However in the eastern part of the granitic massif in the proximity of abandoned uranium mine atmospheric radon content was found to be 6000 Bq/m3. In the case of sedimentary basin where sedimentary sequence of Carboniferous rocks has been penetrated by younger gases and fluids of volcanic origin uranium mineralization developed. The region known from its CO2 outburst during coal mining activity is characterized by good ventilation of the uranium enriched geological basement resulting in increased atmospheric radon activity being in average 72 Bq/m3. In the vicinity of coal mine tailing an increase up to 125 Bq/m3 can be observed. Seasonal variations of atmospheric radon content are influenced in agricultural areas by cyclic cultivation works (plough on soils of increased uranium content and in the case of post-industrial brownfields varying rates of radon exhalation from tailings due to different meteorological conditions.

  14. Activity size distributions for long-lived radon decay products in aerosols collected in Barcelona (Spain)

    International Nuclear Information System (INIS)

    Camacho, A.; Valles, I.; Vargas, A.; Gonzalez-Perosanz, M.; Ortega, X.

    2009-01-01

    The activity median aerodynamic diameters (AMADs) of long-lived radon decay product ( 210 Pb, 210 Po) in aerosols collected in the Barcelona area (Northeast Spain) during the period from April 2006 to February 2008 are presented. The 210 Po mean AMAD was 420 nm, while the 210 Pb mean AMAD was 500 nm. The temporal evolution of 210 Pb and 210 Po AMADs shows maxima in autumn and winter and minima in spring and summer. 210 Pb AMAD are being used to estimate the mean-residence time of atmospheric aerosols.

  15. 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)

  16. Maintenance, endogeneous, respiration, lysis, decay and predation

    DEFF Research Database (Denmark)

    loosdrecht, Marc C. M. Van; Henze, Mogens

    1999-01-01

    mechanism is microbiologically correct. The lysis/decay model mechanism is a strongly simplified representation of reality. This paper tries to review the processes grouped under endogenous respiration in activated sludge models. Mechanisms and processes such as maintenance, lysis, internal and external...... decay, predation and death-regeneration are discussed. From recent microbial research it has become evident that cells do not die by themselves. Bacteria are however subject to predation by protozoa. Bacteria store reserve polymers that in absence of external substrate are used for growth...

  17. Snowpack radiative heating: Influence on Tibetan Plateau climate

    OpenAIRE

    Flanner, Mark G; Zender, C. S.

    2005-01-01

     Solar absorption decays exponentially with depth in snowpacks. However, most climate models constrain all snowpack absorption to occur uniformly in the top-most snow layer. We show that 20–45% of solar absorption by deep snowpacks occurs more than 2 cm beneath the surface. Accounting for vertically-resolved solar heating alters steady-state snow mass without changing bulk snow albedo, and ice-albedo feedback amplifies this effect. Vertically-resolved snowpack heating reduces winter snow mass...

  18. Radiation Heat Transfer Effect on Thermal Sizing of Air-Cooling Heat Exchanger of Emergency Cooldown Tank

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joo Hyung; Kim, Young In; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Myoung Jun; Lee, Hee Joon [School of Mechanical Eng., Kookmin University, Seoul (Korea, Republic of)

    2014-10-15

    An attempt has begun to extend the life time of emergency cooldown tank (ECT) by Korea Atomic Energy Research Institute (KAERI) researchers. Moon et al. recently reported a basic concept upon how to keep the ECT in operation beyond 72 hours after an accident occurs without any active corrective actions for the postulated design basis accidents. When the SMART (System-integrated Modular Advanced Reac-Tor) received its Standard Design Approval (SDA) for the first time in the world, hybrid safety systems are applied. However, the passive safety systems of SMART are being enforced in response to the public concern for much safer reactors since the Fukushima accident occurred. The ECT is a major component of a passive residual heat removal system (PRHRS), which is one of the most important systems to enhance the safety of SMART. It is being developed in a SMART safety enhancement project to contain enough cooling water to remove a sensible heat and a decay heat from reactor core for 72 hours since an accident occurs. Moon et al. offered to install another heat exchanger above the ECT and to recirculate an evaporated steam into water, which enables the ECT to be in operation, theoretically, indefinitely. An investigation was made to determine how long and how many tubes were required to meet the purpose of the study. In their calculation, however, a radiation heat transfer effect was neglected. The present study is to consider the radiation heat transfer for the design of air-cooling heat exchanger. Radiation heat transfer is normally ignored in many situations, but this is not the case for the present study. Kim et al. conducted thermal sizing of scaled-down ECT heat exchanger, which will be used to validate experimentally the basic concept of the present study. Their calculation is also examined to see if a radiation heat transfer effect was taken into consideration. The thermal sizing of an air-cooling heat exchanger was conducted including radiation heat transfer

  19. Direct Hysteresis Heating of Catalytically Active Ni–Co Nanoparticles as Steam Reforming Catalyst

    DEFF Research Database (Denmark)

    Mortensen, Peter Mølgaard; Engbæk, Jakob Soland; Vendelbo, Søren Bastholm

    2017-01-01

    We demonstrated a proof-of-concept catalytic steam reforming flow reactor system heated only by supported magnetic nickel–cobalt nanoparticles in an oscillating magnetic field. The heat transfer was facilitated by the hysteresis heating in the nickel–cobalt nanoparticles alone. This produced...... a sufficient power input to equilibrate the reaction at above 780 °C with more than 98% conversion of methane. The high conversion of methane indicated that Co-rich nanoparticles with a high Curie temperature provide sufficient heat to enable the endothermic reaction, with the catalytic activity facilitated...... by the Ni content in the nanoparticles. The magnetic hysteresis losses obtained from temperature-dependent hysteresis measurements were found to correlate well with the heat generation in the system. The direct heating of the catalytic system provides a fast heat transfer and thereby overcomes the heat...

  20. Measurement of the Convective Heat-Transfer Coefficient

    Science.gov (United States)

    Conti, Rosaria; Gallitto, Aurelio Agliolo; Fiordilino, Emilio

    2014-01-01

    We propose an experiment for investigating how objects cool down toward the thermal equilibrium with their surroundings. We describe the time dependence of the temperature difference of the cooling objects and the environment with an exponential decay function. By measuring the thermal constant t, we determine the convective heat-transfer…

  1. On bulk viscosity and moduli decay

    International Nuclear Information System (INIS)

    Laine, Mikko

    2010-01-01

    This pedagogically intended lecture, one of four under the header 'Basics of thermal QCD', reviews an interesting relationship, originally pointed out by Boedeker, that exists between the bulk viscosity of Yang-Mills theory (of possible relevance to the hydrodynamics of heavy ion collision experiments) and the decay rate of scalar fields coupled very weakly to a heat bath (appearing in some particle physics inspired cosmological scenarios). This topic serves, furthermore, as a platform on which a number of generic thermal field theory concepts are illustrated. The other three lectures (on the QCD equation of state and the rates of elastic as well as inelastic processes experienced by heavy quarks) are recapitulated in brief encyclopedic form. (author)

  2. Heat conduction within linear thermoelasticity

    CERN Document Server

    Day, William Alan

    1985-01-01

    J-B. J. FOURIER'S immensely influential treatise Theorie Analytique de la Chaleur [21J, and the subsequent developments and refinements of FOURIER's ideas and methods at the hands of many authors, provide a highly successful theory of heat conduction. According to that theory, the growth or decay of the temperature e in a conducting body is governed by the heat equation, that is, by the parabolic partial differential equation Such has been the influence of FOURIER'S theory, which must forever remain the classical theory in that it sets the standard against which all other theories are to be measured, that the mathematical investigation of heat conduction has come to be regarded as being almost identicalt with the study of the heat equation, and the reader will not need to be reminded that intensive analytical study has t But not entirely; witness, for example, those theories which would replace the heat equation by an equation which implies a finite speed of propagation for the temperature. The reader is refe...

  3. CP violation in K decays and rare decays

    International Nuclear Information System (INIS)

    Buchalla, G.

    1996-12-01

    The present status of CP violation in decays of neutral kaons is reviewed. In addition selected rare decays of both K and B mesons are discussed. The emphasis is in particular on observables that can be reliably calculated and thus offer the possibility of clean tests of standard model flavor physics. 105 refs

  4. Preliminary study of insects associated to indoor body decay in Colombia

    Directory of Open Access Journals (Sweden)

    Yardany Ramos-Pastrana

    2014-12-01

    Full Text Available Preliminary study of insects associated to indoor body decay in Colombia. This is the first report studying insects associated to indoor body decay process of a white pig (Sus scrofa (Artiodactyla, Suidae in a controlled indoor environment in an urban area of Florencia city, Amazonia Piedmont, Colombia. For a period of 54 days, 9,220 individuals (immature and adults, distributed in 3 orders, 5 families, 10 genera, and 10 species were collected using entomological nets and tweezers. Five decaying stages are described (fresh, bloated, active decay, advanced decay and remains. During the fresh stage we recorded Cochliomyia macellaria (Fabricius, 1775, Chrysomya albiceps (Wiedemann, 1819, Ophyra aenescens (Wiedemann, 1830, Oxysarcodexia sp., Lepidodexia sp. and Lasiophanes sp.; during the bloating stage C. macellaria, C. albiceps, Lucilia eximia (Wiedemann, 1819, Hemilucillia semidiaphana (Rondani, 1850, Musca domestica Linnaeus, 1758, O. aenescens, Oxysarcodexia sp., Lepidodexia sp., Dermestes maculatus De Geer, 1774 and Lasiphanes sp.; during the active decay C. macellaria, C. albiceps, L. eximia, M. domestica, O. aenescens, Lepidodexia sp. D. maculatus and Lasiophanes sp.; during the advanced decay C. macellaria, C. albiceps, M. domestica, Lepidodexia sp. and Lasiophanes sp.; and during the remains stage C. albiceps, D. maculatus and Lasiophanes sp. The insects were sorted out in 3 ecological categories; necrophagous, predators and parasites and sarco-saprophagous. According to Chao and Jack estimators, total richness was observed on day 20, with 100% of the expected species.

  5. Effects of heat stress on serum insulin, adipokines, AMP-activated protein kinase, and heat shock signal molecules in dairy cows.

    Science.gov (United States)

    Min, Li; Cheng, Jian-bo; Shi, Bao-lu; Yang, Hong-jian; Zheng, Nan; Wang, Jia-qi

    2015-06-01

    Heat stress affects feed intake, milk production, and endocrine status in dairy cows. The temperature-humidity index (THI) is employed as an index to evaluate the degree of heat stress in dairy cows. However, it is difficult to ascertain whether THI is the most appropriate measurement of heat stress in dairy cows. This experiment was conducted to investigate the effects of heat stress on serum insulin, adipokines (leptin and adiponectin), AMP-activated protein kinase (AMPK), and heat shock signal molecules (heat shock transcription factor (HSF) and heat shock proteins (HSP)) in dairy cows and to research biomarkers to be used for better understanding the meaning of THI as a bioclimatic index. To achieve these objectives, two experiments were performed. The first experiment: eighteen lactating Holstein dairy cows were used. The treatments were: heat stress (HS, THI average=81.7, n=9) and cooling (CL, THI average=53.4, n=9). Samples of HS were obtained on August 16, 2013, and samples of CL were collected on April 7, 2014 in natural conditions. The second experiment: HS treatment cows (n=9) from the first experiment were fed for 8 weeks from August 16, 2013 to October 12, 2013. Samples for moderate heat stress, mild heat stress, and no heat stress were obtained, respectively, according to the physical alterations of the THI. Results showed that heat stress significantly increased the serum adiponectin, AMPK, HSF, HSP27, HSP70, and HSP90 (Pdairy cows. When heat stress treatment lasted 8 weeks, a higher expression of HSF and HSP70 was observed under moderate heat stress. Serum HSF and HSP70 are sensitive and accurate in heat stress and they could be potential indicators of animal response to heat stress. We recommend serum HSF and HSP70 as meaningful biomarkers to supplement the THI and evaluate moderate heat stress in dairy cows in the future.

  6. Effect of tunnel cross section on gas temperatures and heat fluxes in case of large heat release rate

    International Nuclear Information System (INIS)

    Fan, Chuan Gang; Li, Ying Zhen; Ingason, Haukur; Lönnermark, Anders

    2016-01-01

    Highlights: • The effect of tunnel cross section together with ventilation velocity was studied. • Ceiling temperature varies clearly with tunnel height, but little with tunnel width. • Downstream temperature decreases with increasing tunnel dimensions. • HRR is an important factor that influences decay rate of excess gas temperature. • An equation considering both tunnel dimensions and HRR was developed. - Abstract: Tests with liquid and solid fuels in model tunnels (1:20) were performed and analysed in order to study the effect of tunnel cross section (width and height) together with ventilation velocity on ceiling gas temperatures and heat fluxes. The model tunnel was 10 m long with varying width (0.3 m, 0.45 m and 0.6 m) and height (0.25 m and 0.4 m). Test results show that the maximum temperature under the ceiling is a weak function of heat release rate (HRR) and ventilation velocity for cases with HRR more than 100 MW at full scale. It clearly varies with the tunnel height and is a weak function of the tunnel width. With a lower tunnel height, the ceiling is closer to the base of continuous flame zone and the temperatures become higher. Overall, the gas temperature beneath the ceiling decreases with the increasing tunnel dimensions, and increases with the increasing longitudinal ventilation velocity. The HRR is also an important factor that influences the decay rate of excess gas temperature, and a dimensionless HRR integrating HRR and other two key parameters, tunnel cross-sectional area and distance between fuel centre and tunnel ceiling, was introduced to account for the effect. An equation for the decay rate of excess gas temperature, considering both the tunnel dimensions and HRR, was developed. Moreover, a larger tunnel cross-sectional area will lead to a smaller heat flux.

  7. Modeling of heat transfer in a horizontal heat-generating layer by an effective diffusivity approach

    International Nuclear Information System (INIS)

    Cheung, F.B.; Shiah, S.W.

    1994-01-01

    The concept of effective diffusivity is employed to model various processes of heat transfer in a volumetrically heated fluid layer subjected to different initial and boundary conditions. The approach, which involves the solution of only heat diffusion equations, is found to give rather accurate predictions of the transient response of an initially stagnant fluid layer to a step input of power as well as the developing and decaying nature of the flow following a step change in the internal Rayleigh number from one state of steady convection to another. The approach is also found to be applicable to various flow regions of a heat-generating fluid layer, and is not limited to the case in which the entire layer is in turbulent motion. The simplicity and accuracy of the method are clearly illustrated in the analysis. Validity of the effective diffusivity approach is demonstrated by comparing the predicted results with corresponding experimental data

  8. Proton decay and light sterile neutrinos

    Science.gov (United States)

    Helo, Juan C.; Hirsch, Martin; Ota, Toshihiko

    2018-06-01

    Within the standard model, non-renormalizable operators at dimension six ( d = 6) violate baryon and lepton number by one unit and thus lead to proton decay. Here, we point out that the proton decay mode with a charged pion and missing energy can be a characteristic signature of d = 6 operators containing a light sterile neutrino, if it is not accompanied by the standard π0 e + final state. We discuss this effect first at the level of effective operators and then provide a concrete model with new physics at the TeV scale, in which the lightness of the active neutrinos and the stability of the proton are related.

  9. Heat storage in Asian elephants during submaximal exercise: behavioral regulation of thermoregulatory constraints on activity in endothermic gigantotherms.

    Science.gov (United States)

    Rowe, M F; Bakken, G S; Ratliff, J J; Langman, V A

    2013-05-15

    Gigantic size presents both opportunities and challenges in thermoregulation. Allometric scaling relationships suggest that gigantic animals have difficulty dissipating metabolic heat. Large body size permits the maintenance of fairly constant core body temperatures in ectothermic animals by means of gigantothermy. Conversely, gigantothermy combined with endothermic metabolic rate and activity likely results in heat production rates that exceed heat loss rates. In tropical environments, it has been suggested that a substantial rate of heat storage might result in a potentially lethal rise in core body temperature in both elephants and endothermic dinosaurs. However, the behavioral choice of nocturnal activity might reduce heat storage. We sought to test the hypothesis that there is a functionally significant relationship between heat storage and locomotion in Asian elephants (Elephas maximus), and model the thermoregulatory constraints on activity in elephants and a similarly sized migratory dinosaur, Edmontosaurus. Pre- and post-exercise (N=37 trials) measurements of core body temperature and skin temperature, using thermography were made in two adult female Asian elephants at the Audubon Zoo in New Orleans, LA, USA. Over ambient air temperatures ranging from 8 to 34.5°C, when elephants exercised in full sun, ~56 to 100% of active metabolic heat production was stored in core body tissues. We estimate that during nocturnal activity, in the absence of solar radiation, between 5 and 64% of metabolic heat production would be stored in core tissues. Potentially lethal rates of heat storage in active elephants and Edmontosaurus could be behaviorally regulated by nocturnal activity.

  10. Impact of dark matter on reionization and heating

    NARCIS (Netherlands)

    Mapelli, M.; Ripamonti, E.

    2007-01-01

    Abstract: We derived the evolution of the energy deposition in the intergalactic medium (IGM) by different decaying (or annihilating) dark matter (DM) candidates. Heavy annihilating DM particles (with mass larger than a few GeV) have no influence on reionization and heating, even if we assume that

  11. Heat production / host rock compatibility; Waermeentwicklung / Gesteinsvertraeglichkeit

    Energy Technology Data Exchange (ETDEWEB)

    Meleshyn, A.; Weyand, T.; Bracke, G.; Kull, H.; Wieczorek, K.

    2016-05-15

    For the final high-level radioactive waste repository potential host rock formations are either rock salt or clays (Kristallin). Heat generating waste (decay heat of the radioactive materials) can be absorbed by the host rock. The effect of temperature increase on the thermal conductivity, the thermal expansion and the mechanical properties of salt, Kristallin, clays and argilliferous geotechnical barriers are described. Further issues of the report are the mineralogical behavior, phase transformations, hydrochemistry, microbial processes, gas formation, thermochemical processes and gas ingress. Recommendations for further research are summarized.

  12. Behavior of radon and its decay products in physical media

    International Nuclear Information System (INIS)

    Chavez B, A.

    1990-06-01

    This study was carried out to know the radon behavior, it shows some of its decay products of short life, the same as the equations that describe the growth of the activity of each decay product in a source that initially is radon. The study threw results that they are applicable in geology, uranium prospecting, as well as in radiological safety. The use of membranes to filter the decay products of radon and the use of these for protection of the detector, it has opened a new line in the study of the radon. (Author)

  13. Design Evaluation of Thermal-hydraulic Test Facility for a Finned-tube Sodium-to-Air Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungmo; Kim, Byeong-Yeon; Ko, Yung Joo; Cho, Youngil; Kim, Jong-Man; Son, Seok-Kwon; Jo, Youngchul; Kang, Byeong Su; Jung, Minhwan; Eoh, Jaehyuk; Lee, Hyeong-Yeon; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This paper introduces the recent progress of overall design phase for the SELFA facility and deals with basic thermal-hydraulic design parameters and its design validation as well. For more reliable design of the safety-grade decay heat removal system (DHRS) in PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), two kinds of sodium-to-air heat exchangers have been employed in the system as an ultimate heat sink. One is a natural draft sodium-to-air heat exchanger (AHX) with helically-coiled sodium tubes, and the other is a forced draft sodium-to-air heat exchanger (FHX) with finned tubes with a straight-type arranged. Since the FHX is normally operated in an active mode with a forced air draft conditions, its performance should be verified for any anticipated operating conditions. To validate the test section design, evaluations of both thermal-hydraulic and mechanical design have been carried out, and some new concepts or devices were newly employed to replicate the prototypic conditions as closely as possible.

  14. Charm Decays at BABAR

    International Nuclear Information System (INIS)

    Charles, M.

    2004-01-01

    The results of several studies of charmed mesons and baryons at BABAR are presented. First, searches for the rare decays D 0 → l + l - are presented and new upper limits on these processes are established. Second, a measurement of the branching fraction of the isospin-violating hadronic decay D* s (2112) + → D s + π 0 relative to the radiative decay D* s (2112) + → D s + γ is made. Third, the decays of D* sJ (2317) + and D sJ (2460) + mesons are studied and ratios of branching fractions are measured. Fourth, Cabibbo-suppressed decays of the Λ c + are examined and their branching fractions measured relative to Cabibbo-allowed modes. Fifth, the Χ c 0 is studied through its decays to Χ - π + and (Omega) - K + ; in addition to measuring the ratio of branching fractions for Χ c 0 produced from the c(bar c) continuum, the uncorrected momentum spectrum is measured, providing clear confirmation of Χ c 0 production in B decays

  15. Heat-Activated Persulfate Oxidation of Chlorinated Solvents in Sandy Soil

    Directory of Open Access Journals (Sweden)

    Jialu Liu

    2014-01-01

    Full Text Available Heat-activated persulfate oxidative treatment of chlorinated organic solvents containing chlorinated ethenes and ethanes in soil was investigated with different persulfate dosages (20 g/L, 40 g/L, and 60 g/L and different temperatures (30°C, 40°C, and 50°C. Chlorinated organic solvents removal was increased as persulfate concentration increase. The persulfate dosage of 20 g/L with the highest OE (oxidant efficiency value was economically suitable for chlorinated organic solvents removal. The increasing temperature contributed to the increasing depletion of chlorinated organic solvents. Chlorinated ethenes were more easily removed than chlorinated ethanes. Moreover, the persulfate depletion followed the pseudo-first-order reaction kinetics (kps=0.0292 [PS]0+0.0008, R2=0.9771. Heat-activated persulfate appeared to be an effective oxidant for treatment of chlorinated hydrocarbons.

  16. Using microwave heating to improve the desorption efficiency of high molecular weight VOC from beaded activated carbon.

    Science.gov (United States)

    Fayaz, Mohammadreza; Shariaty, Pooya; Atkinson, John D; Hashisho, Zaher; Phillips, John H; Anderson, James E; Nichols, Mark

    2015-04-07

    Incomplete regeneration of activated carbon loaded with organic compounds results in heel build-up that reduces the useful life of the adsorbent. In this study, microwave heating was tested as a regeneration method for beaded activated carbon (BAC) loaded with n-dodecane, a high molecular weight volatile organic compound. Energy consumption and desorption efficiency for microwave-heating regeneration were compared with conductive-heating regeneration. The minimum energy needed to completely regenerate the adsorbent (100% desorption efficiency) using microwave regeneration was 6% of that needed with conductive heating regeneration, owing to more rapid heating rates and lower heat loss. Analyses of adsorbent pore size distribution and surface chemistry confirmed that neither heating method altered the physical/chemical properties of the BAC. Additionally, gas chromatography (with flame ionization detector) confirmed that neither regeneration method detectably altered the adsorbate composition during desorption. By demonstrating improvements in energy consumption and desorption efficiency and showing stable adsorbate and adsorbent properties, this paper suggests that microwave heating is an attractive method for activated carbon regeneration particularly when high-affinity VOC adsorbates are present.

  17. Heat transfer calculations on the KNK II emergency cooling system

    International Nuclear Information System (INIS)

    Vossebrecker, H.; Groenefeld, G.

    1976-12-01

    The Licensing Authority had demanded that in case of the change of the KNK thermal core into a fast core the decay heat removal system must be improved by a diverse and spatially separated emergency cooling system. In order to meet this requirement an existing nitrogen system of the facility is extended in such a manner that the decay heat will be removed by a nitrogen flow passing through the gap between reactor vessel and guard vessel. The heat transport from the core to the vessel is accomplished by natural convection flow rates which are generated by density differences between the hot core subassemblies, the reflector subassemblies and other passages between the upper and the lower plenum. The calculations show that the maximum temperatures in the core do not reach the sodium boiling-point. The maximum vessel temperature is 673 deg. C. In this report the function of the emergency cooling system and the methods of calculation are described, the input data and the results are stated and it is shown that the calculated temperatures are conservative [de

  18. A study of a small nuclear power plant system for district heating

    International Nuclear Information System (INIS)

    Imamura, Mitsuru; Sato, Kotaro; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

    2009-01-01

    We have studied nuclear power plant for district heating. Already some towns and villages in Hokkaido have requested small reactor for district heating. Using existing technology allows us to shorten development period and to keep a lid on development cost. We decided to develop new reactor based on 'MUTSU' reactor technology because 'MUTSU' had already proved its safety. And this reactor was boron free reactor. It allows plant system to reduce the chemical control system. And moderator temperature coefficient is deeply negative. It means to improve its operability and leads to dependability enhancement. We calculated burn-up calculation of erbium addition fuel. In the result, the core life became about 10 years. And we adapt the cassette type refueling during outagein in order to maintain nonproliferation. In the district heating system, a double heat exchanger system enables to response to load change in season. To obtain the acceptance of public, this system has a leak prevention system of radioactive materials to public. And road heating system of low grade heat utilization from turbine condenser leads to improve the heat utilization efficiency. We carried out performance evaluation test of district heating pipeline. Then the heat loss of pipeline is estimated at about 0.440degC/km. This result meets general condition, which is about 1degC/km. This small plant has passive safety system. It is natural cooling of containment vessel. In case of loss of coolant accident, decay heat can remove by natural convection air cooling after 6 hours. Decay heat within 6 hours can remove by evaporative heat transfer of pool on containment vessel. (author)

  19. Rare B decays, rare τ decays, and grand unification

    International Nuclear Information System (INIS)

    Sher, M.; Yuan, Y.

    1991-01-01

    In multi-Higgs-boson extensions of the standard model, tree-level flavor-changing neutral currents exist naturally, unless suppressed by some symmetry. For a given rate, the exchanged scalar or pseudoscalar mass is very sensitive to the flavor-changing coupling between the first two generations. Since the Yukawa couplings of the first two generations are unknown and certainly very small, bounds which rely on some assumed value of this flavor-changing coupling are quite dubious. One might expect the size (and reliability) of the Yukawa couplings involving the third generation to be greater. In this paper, we consider processes involving τ's and B's, and determine the bounds on the flavor-changing couplings which involve third-generation fields. The strongest bound in the quark sector comes from B-bar B mixing and in the lepton sector, surprisingly, from μ→eγ. It is then noted that the flavor-changing couplings in the quark sector are related to those in the lepton sector in many grand unified theories, and one can ask whether an analysis of rare τ decays or rare B decays will provide the strongest constraints. We show that rare B decays provide the strongest bounds, and that no useful information can be obtained from rare τ decays. It is also noted that the most promising decay modes are B→Kμτ and B s →μτ, and we urge experimenters to look for rare decay modes of the B in which a τ is in the final state

  20. Column: Factors Affecting Data Decay

    Directory of Open Access Journals (Sweden)

    Kevin Fairbanks

    2012-06-01

    Full Text Available In nuclear physics, the phrase decay rate is used to denote the rate that atoms and other particles spontaneously decompose. Uranium-235 famously decays into a variety of daughter isotopes including Thorium and Neptunium, which themselves decay to others. Decay rates are widely observed and wildly different depending on many factors, both internal and external. U-235 has a half-life of 703,800,000 years, for example, while free neutrons have a half-life of 611 seconds and neutrons in an atomic nucleus are stable.We posit that data in computer systems also experiences some kind of statistical decay process and thus also has a discernible decay rate. Like atomic decay, data decay fluctuates wildly. But unlike atomic decay, data decay rates are the result of so many different interplaying processes that we currently do not understand them well enough to come up with quantifiable numbers. Nevertheless, we believe that it is useful to discuss some of the factors that impact the data decay rate, for these factors frequently determine whether useful data about a subject can be recovered by forensic investigation.(see PDF for full column