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Sample records for decay heat analysis

  1. 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

  2. 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

  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. 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.

  5. 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

  6. 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

  7. 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

  8. 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)

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

    Directory of Open Access Journals (Sweden)

    Stankunas Gediminas

    2017-01-01

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

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. Status of the Japanese decay heat standard

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1992-01-01

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

  5. 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. 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)

  7. 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)

  8. 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.

  9. 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...

  10. 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)

  11. 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)

  12. 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...

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

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1983-01-01

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

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    developed based on the simulation results from two analysis codes, MARS and MAAP. The early initiation and high mass flow rate from the passive safety injection system could guarantee the convergent containment pressure. In addition, in the loss of coolant accident with the failures of safety injection systems, the passive safety injection system and the containment filtered venting system can be used as a new accident management method. (authors)

  17. 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.)

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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)

  10. Consistency among integral measurements of aggregate decay heat power

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. 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)

  20. 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

  1. 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%

  2. 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)

  3. 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

  4. 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

  5. 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)

  6. Analysis of Heat Transfer

    International Nuclear Information System (INIS)

    2003-08-01

    This book deals with analysis of heat transfer which includes nonlinear analysis examples, radiation heat transfer, analysis of heat transfer in ANSYS, verification of analysis result, analysis of heat transfer of transition with automatic time stepping and open control, analysis of heat transfer using arrangement of ANSYS, resistance of thermal contact, coupled field analysis such as of thermal-structural interaction, cases of coupled field analysis, and phase change.

  7. 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

  8. 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)

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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

  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. 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.)

  19. 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)

  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. 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)

  2. 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.

  3. Decay power evaluation for licensing analysis

    International Nuclear Information System (INIS)

    Tran, H.; Schrock, V.E.

    1987-01-01

    The ANSI/ANS 5.1-1979 Standard on Decay Power in shutdown reactors has been available as the basis for accident analysis for the past 7 yr. The US Nuclear Regulatory Commission has made a commitment to use this standard in new licensing approaches and has approved a licensing model for boiling water. More sweeping changes in the licensing rules are currently under review that will involve the use of best-estimate models and a statistical evaluation of the uncertainty (95% confidence level) in the key results. The structure of the decay power standard is well suited for such applications because it provides a statistically meaningful uncertainty in the decay power from fission products. The normalized decay power is a function specific to each point in the reactor volume due to the fact that the fuel composition develops a spatial dependence as burnup proceeds and decay power depends on the mix of fissioning nuclides. For reactor safety calculations it is desirable to employ a single temporal decay power function for the whole core inasmuch as many variations of accident parameters are required. This is the usual approach in large system thermal-hydraulics codes. Such a single representative or generic curve for a specified total operating power history can be acceptable but at the expense of some increase in the uncertainty. In this paper, the author present a method of evaluating the additional uncertainty in the decay power associated with use of a generic curve

  4. 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)

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. Global analysis of muon decay measurements

    International Nuclear Information System (INIS)

    Gagliardi, C.A.; Tribble, R.E.; Williams, N.J.

    2005-01-01

    We have performed a global analysis of muon decay measurements to establish model-independent limits on the space-time structure of the muon decay matrix element. We find limits on the scalar, vector, and tensor coupling of right- and left-handed muons to right- and left-handed electrons. The limits on those terms that involve the decay of right-handed muons to left-handed electrons are more restrictive than in previous global analyses, while the limits on the other nonstandard model interactions are comparable. The value of the Michel parameter η found in the global analysis is -0.0036±0.0069, slightly more precise than the value found in a more restrictive analysis of a recent measurement. This has implications for the Fermi coupling constant G F

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. Heat Capacity Analysis Report

    International Nuclear Information System (INIS)

    Findikakis, A.

    2004-01-01

    The purpose of this report is to provide heat capacity values for the host and surrounding rock layers for the waste repository at Yucca Mountain. The heat capacity representations provided by this analysis are used in unsaturated zone (UZ) flow, transport, and coupled processes numerical modeling activities, and in thermal analyses as part of the design of the repository to support the license application. Among the reports that use the heat capacity values estimated in this report are the ''Multiscale Thermohydrologic Model'' report, the ''Drift Degradation Analysis'' report, the ''Ventilation Model and Analysis Report, the Igneous Intrusion Impacts on Waste Packages and Waste Forms'' report, the ''Dike/Drift Interactions report, the Drift-Scale Coupled Processes (DST and TH Seepage) Models'' report, and the ''In-Drift Natural Convection and Condensation'' report. The specific objective of this study is to determine the rock-grain and rock-mass heat capacities for the geologic stratigraphy identified in the ''Mineralogic Model (MM3.0) Report'' (BSC 2004 [DIRS 170031], Table 1-1). This report provides estimates of the heat capacity for all stratigraphic layers except the Paleozoic, for which the mineralogic abundance data required to estimate the heat capacity are not available. The temperature range of interest in this analysis is 25 C to 325 C. This interval is broken into three separate temperature sub-intervals: 25 C to 95 C, 95 C to 114 C, and 114 C to 325 C, which correspond to the preboiling, trans-boiling, and postboiling regimes. Heat capacity is defined as the amount of energy required to raise the temperature of a unit mass of material by one degree (Nimick and Connolly 1991 [DIRS 100690], p. 5). The rock-grain heat capacity is defined as the heat capacity of the rock solids (minerals), and does not include the effect of water that exists in the rock pores. By comparison, the rock-mass heat capacity considers the heat capacity of both solids and pore

  19. 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.)

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. Elementary heat transfer analysis

    CERN Document Server

    Whitaker, Stephen; Hartnett, James P

    1976-01-01

    Elementary Heat Transfer Analysis provides information pertinent to the fundamental aspects of the nature of transient heat conduction. This book presents a thorough understanding of the thermal energy equation and its application to boundary layer flows and confined and unconfined turbulent flows. Organized into nine chapters, this book begins with an overview of the use of heat transfer coefficients in formulating the flux condition at phase interface. This text then explains the specification as well as application of flux boundary conditions. Other chapters consider a derivation of the tra

  5. 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.

  6. 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

  7. 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.)

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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

  14. Analysis of new physics in B decays

    Energy Technology Data Exchange (ETDEWEB)

    Feger, Robert

    2010-03-26

    The inclusive semileptonic decay B{yields}X{sub c}l{nu} yields the best extracted value of vertical stroke V{sub cb} vertical stroke with an uncertainty of about 2%, which is due to its precise theoretical description by the heavy-quark expansion (HQE) and the large amount of data collected by the B factories Babar and Belle. Besides the precision determination of vertical stroke V{sub cb} vertical stroke and the HQE parameter it also enables us to test for new physics effects. Normal decays with a small standard model contribution are investigated on the quest for new physics, like flavour-changing neutral currents, which are suppressed in the standard model by the GIM mechanism. But as right-handed weak currents are absent in the standard model their appearance would be a smoking gun signal for new physics in the decay considered here. We perform such an analysis, which is known form the leptonic sector as ''Michel parameter analysis''. Contributions from possible new physics effects are derived with effective field theory methods, yielding an enhanced b {yields} c current, with not only a right-handed vector contribution, but also scalar and tensor couplings. We repeat the computations of the decay with the enhanced current up to O(1/m{sub b}{sup 2}) in the HQE and up to O({alpha}{sub s}) in the perturbative corrections. A moment analysis of the moments of the lepton energy spectrum and the hadronic invariant mass has become a reliable tool in both the theoretical evaluation and the experimental determination. This combines in a HQE fit, which has been improved frequently, yielding a precise determination of the HQE parameters and especially the best extracted value for vertical stroke V{sub cb} vertical stroke. The HQE fit may as well serve as a test for new physics effects in these moments. The theoretical analysis of the moments reveals a low sensitivity of the moments on the non-vector currents. Thus we perform the fit with only a

  15. 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.

  16. Decay assessment through thermographic analysis in architectural and archaeological heritage

    Science.gov (United States)

    Gomez-Heras, Miguel; Martinez-Perez, Laura; Fort, Rafael; Alvarez de Buergo, Monica

    2010-05-01

    Any exposed stone-built structure is subject to thermal variations due to daily, seasonal and secular environmental temperature changes. Surface temperature is a function of air temperature (due to convective heat transfer) and of infrared radiation received through insolation. While convective heat transfer homogenizes surface temperature, stone response to insolation is much more complex and the temporal and spatial temperature differences across structures are enhanced. Surface temperature in stone-built structures will be affected by orientation, sunlight inclination and the complex patterns of light and shadows generated by the often intricate morphology of historical artefacts and structures. Surface temperature will also be affected by different material properties, such as albedo, thermal conductivity, transparency and absorbance to infrared radiation of minerals and rocks. Moisture and the occurrence of salts will also be a factor affecting surface temperatures. Surface temperatures may as well be affected by physical disruptions of rocks due to differences in thermal inertia generated by cracks and other discontinuities. Thermography is a non-invasive, non-destructive technique that measures temperature variations on the surface of a material. With this technique, surface temperature rates of change and their spatial variations can be analysed. This analysis may be used not only to evaluate the incidence of thermal decay as a factor that generates or enhances stone decay, but also to detect and evaluate other factors that affect the state of conservation of architectural and archaeological heritage, as for example moisture, salts or mechanical disruptions.

  17. 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

    conditions, need to be powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research - aimed at achieving enhanced passivity of the DHR system for the GFR - are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study - TRACE and CATHARE - are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a detailed re-analysis of earlier PSI (EIR, at the time) gas-loop experiments conducted in the 1970's. Conclusions and recommendations on the models to be used for transient analysis are derived. In general, it has been shown that the agreement, between experiments and the correlations for heat transfer and friction used in TRACE and CATHARE, is quite satisfactory. The thus validated codes are then used in the two detailed, DHR improvement studies carried out. The first improvement of the reference DHR strategy is the heavy gas injection. Assuming a DHR blower failure after a LOCA, the helium pressure in the guard containment is not high enough to evacuate the decay heat by natural convection. To improve the natural convection, the effects of injecting different heavy gases (N 2 , CO 2 , Ar and a N 2 /He mixture) into the primary circuit were analyzed, in order to address the possibility of dealing with DHR-blower failure while accepting an intermediate back-up pressure in the guard

  18. 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

    powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research aimed at achieving enhanced passivity of the DHR system for the GFR are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study – TRACE and CATHARE – are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a detailed re-analysis of earlier PSI (EIR, at the time) gas-loop experiments conducted in the 1970s. Conclusions and recommendations on the models to be used for transient analysis are derived. In general, it has been shown that the agreement, between experiments and the correlations for heat transfer and friction used in TRACE and CATHARE, is quite satisfactory. The thus validated codes are then used in the two detailed, DHR improvement studies carried out. The first improvement of the reference DHR strategy is the heavy gas injection. Assuming a DHR blower failure after a LOCA, the helium pressure in the guard containment is not high enough to evacuate the decay heat by natural convection. To improve the natural convection, the effects of injecting different heavy gases (N 2 , CO 2 , Ar and a N 2 /He mixture) into the primary circuit were analyzed, in order to address the possibility of dealing with DHR-blower failure while accepting an intermediate back-up pressure in the guard containment

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

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

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  9. 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

  10. 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)

  11. 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)

  12. 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)

  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. 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

  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. 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

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

    International Nuclear Information System (INIS)

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

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

  20. 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...

  1. 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)

  2. 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)

  3. 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)

  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. 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)

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. Analysis for rare decay modes (E787)

    International Nuclear Information System (INIS)

    Muramatsu, Norihito

    2002-01-01

    BNL-AGS-E787 experiment is designed to study stopped K + decays. Rare decay mode K + → π + νν-bar is attractive because an absolute value of a CKM matrix element, |V td |, can be extracted. E787 collected K + → π + νν-bar trigger data, which corresponds to 3.2 x 10 12 K + exposures, in 1995-1997. Offline cuts to suppress backgrounds are developed with 'Bifurcation Method', and the background level is estimated to be 0.083±0.019 events inside the signal region. One candidate event is observed after applying the prepared cuts. This event survives with 10 times tighter cuts, which retain 33% of the acceptance. This observation results in BR(K + → π + νν-bar) = 1.52 -1.26 +3.48 x 10 -10 and 0.0024 td | + → π + π 0 νν-bar could also give information on |V td |. The 1995 data, which corresponds to 1.3 x 10 8 K + exposures, are analyzed, and zero events are observed in the signal region. A background level is estimated to be 0.068±0.021 events. The upper limit of the branching ratio is calculated to be 4.3 x 10 -5 at 90% confidence level. (author)

  13. 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

  14. 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

  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. 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)

  17. 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

  18. Amplitude analysis of B- -> D+pi(-)pi(-) decays

    NARCIS (Netherlands)

    Aaij, R.; Adeva, B.; Adinolfi, M.; Ajaltouni, Z.; Akar, S.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Cartelle, P. Alvarez; Alves, A. A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Andreassi, G.; Andreotti, M.; Andrews, J. E.; Appleby, R. B.; Gutierrez, O. Aquines; Archilli, F.; d'Argent, P.; Romeu, J. Arnau; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Babuschkin, I.; Bachmann, S.; Back, J. J.; Badalov, A.; Baesso, C.; Baker, S.; Baldini, W.; Barlow, R. J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Batsukh, B.; Battista, V.; Beaucourt, L.; Beddow, J.; Bedeschi, F.; Onderwater, C. J. G.; Pellegrino, A.; Tolk, S.

    2016-01-01

    The Dalitz plot analysis technique is used to study the resonant substructures of B- -> D+pi(-)pi(-) decays in a data sample corresponding to 3.0 fb(-1) of pp collision data recorded by the LHCb experiment during 2011 and 2012. A model-independent analysis of the angular moments demonstrates the

  19. 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

  20. '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

  1. General classification and analysis of neutron β-decay experiments

    International Nuclear Information System (INIS)

    Gudkov, V.; Greene, G.L.; Calarco, J.R.

    2006-01-01

    A general analysis of the sensitivities of neutron β-decay experiments to manifestations of possible interaction beyond the standard model is carried out. In a consistent fashion, we take into account all known radiative and recoil corrections arising in the standard model. This provides a description of angular correlations in neutron decay in terms of one parameter, which is accurate to the level of ∼10 -5 . Based on this general expression, we present an analysis of the sensitivities to new physics for selected neutron decay experiments. We emphasize that the usual parametrization of experiments in terms of the tree-level coefficients a,A, and B is inadequate when the experimental sensitivities are at the same or higher level relative to the size of the corrections to the tree-level description

  2. 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

  3. 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.

  4. 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.)

  5. PQCD analysis of inclusive semileptonic decays of B mesons

    International Nuclear Information System (INIS)

    Li, H.; Yu, H.

    1996-01-01

    We develop the perturbative QCD formalism for inclusive semileptonic B meson decays, which includes Sudakov suppression from the resummation of large radiative corrections near the high end of charged lepton energy. Transverse degrees of freedom of partons are introduced to facilitate the factorization of B meson decays. Ambiguities appearing in the quark-level analysis are then avoided. A universal distribution function, arising from the nonperturbative Fermi motion of the b quark, is constructed according to the heavy quark effective field theory based operator product expansion, through which the mean and the width of the distribution function are related to hadronic matrix elements of local operators. Charged lepton spectra of the B→X ul ν decay are presented. We find 50% suppression near the end point of the spectrum. The overall suppression on the total decay rate is 8% for the free quark model, and is less than 7% for the use of smooth distribution functions. With our predictions, it is then possible to extract the Cabibbo-Kobayashi-Maskawa matrix element parallel V ub parallel from experimental data. We also discuss possible implications of our analysis when confronted with the rather small observed semileptonic branching ratio in B meson decays. copyright 1996 The American Physical Society

  6. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. 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)

  9. 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)

  10. 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

  11. Introduction to charm decay analysis in fixed target experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bediaga, Ignacio; Goebel, Carla

    1996-01-01

    We present an introduction to data analysis in Experimental High Energy Physics, and some concepts and useful tools are discussed. To illustrate, we use the data of E-791, a fixed target experiment recently realized at Fermilab. In particular, we analyse decay modes of D{sup +} meson with three charged particles in the final state. (author). 8 refs., 22 figs., 1 tab.

  12. Analysis of the B+ →D+ K decay

    Indian Academy of Sciences (India)

    2016-12-06

    Dec 6, 2016 ... decay. In perturbative QCD, Xiao et al have consid- ered complete twist-3 contributions, and in the QCDF, state-of-the art analysis was performed according to. QCDF by four-parameter scenarios. The branching ratios obtained from these scenarios, considering large arbitrary numbers (the same work has ...

  13. Introduction to charm decay analysis in fixed target experiments

    International Nuclear Information System (INIS)

    Bediaga, Ignacio; Goebel, Carla.

    1996-01-01

    We present an introduction to data analysis in Experimental High Energy Physics, and some concepts and useful tools are discussed. To illustrate, we use the data of E-791, a fixed target experiment recently realized at Fermilab. In particular, we analyse decay modes of D + meson with three charged particles in the final state. (author). 8 refs., 22 figs., 1 tab

  14. 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)

  15. 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.)

  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. Heat transfer capability analysis of heat pipe for space reactor

    International Nuclear Information System (INIS)

    Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang

    2015-01-01

    To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. Thermal hydraulic analysis of the encapsulated nuclear heat source

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

    2001-07-01

    An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

  4. On the decay analysis of electroproduced higher-spin hadrons

    Energy Technology Data Exchange (ETDEWEB)

    Actor, A [Heidelberg Univ. (F.R. Germany). Inst. fuer Theoretische Physik

    1975-01-01

    We study exclusive electroproduction of two hadrons where one or both of the hadrons is a higher-spin resonance whose decay is analysed. Interesting electroproduction experiments of this type are discussed. A complete formal apparatus is given for the decay analysis of spin-J particles produced in this way with polarized or unpolarized electron beams and targets. The cases J=1, 3/2, 2 and 5/2 are worked out in detail. A conventional helicity frame analysis in terms of s-channel helicity amplitudes is given. Also we rearrange the formalism for use in the Gottfried-Jackson frame with everything given in terms of t-channel helicity amplitudes. The t-channels formalism makes it possible to completely separate the contributions from longitudinal and transverse virtual photon t-channel helicity states when only the laboratory azimuthal angle between the lepton and hadron planes is variable.

  5. 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)

  6. 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)

  7. 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

  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. 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.

  10. 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.

  11. 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.

  12. Heat transfer analysis of short helical borehole heat exchangers

    International Nuclear Information System (INIS)

    Zarrella, Angelo; De Carli, Michele

    2013-01-01

    Highlights: ► Vertical ground heat exchanger with a helical shaped pipe is analyzed. ► The model considers the interaction between the ground and the environment. ► The results of the model are in good agreement with the experimental values. ► The weather conditions considerably affect the fluid heat carrier temperature. ► The pitch between the turns does not affect the behaviour of the heat exchanger. -- Abstract: In this paper a numerical model to analyze the thermal behaviour of vertical ground heat exchangers with a helical shaped pipe is presented. This type of configuration can be a suitable alternative to conventional ground heat exchangers, especially when the heating and cooling loads of the building are very low. The model describes the heat transfer problem by means of a network of interconnected thermal resistances and capacitances. Moreover, as the investigated ground heat exchanger is usually installed in shallow depth, the model takes into account the interaction between the ground and the ambient environment which affects the fluid heat carrier temperature into the heat exchanger and, as a consequence, the energy efficiency of the heat pump. After a sensitivity analysis on the mesh parameters, the presented model is compared with experimental data and the simulation results show good agreement with the measurements. Finally, analyses to investigate the influence of the weather conditions, of the axial heat transfer and of the pitch between the turns of the helical pipe for two types of ground are carried out.

  13. 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.

  14. 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

  15. 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

  16. 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)

  17. 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).

  18. Quark-diagram analysis of charmed-baryon decays

    International Nuclear Information System (INIS)

    Kohara, Y.

    1991-01-01

    The Cabibbo-allowed two-body nonleptonic decays of charmed baryons to a SU(3)-octet (or -decuplet) baryon and a pseudoscalar meson are examined on the basis of the quark-diagram scheme. Some relations among the decay amplitudes or rates of various decay modes are derived. The decays of Ξ c + to a decuplet baryon are forbidden

  19. 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)

  20. 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)

  1. 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

  2. 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)

  3. 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

  4. Amplitude analysis of $B^- \\to D^+ \\pi^- \\pi^-$ decays

    CERN Document Server

    Aaij, Roel; Adinolfi, Marco; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Andreassi, Guido; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; d'Argent, Philippe; Arnau Romeu, Joan; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Babuschkin, Igor; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baker, Sophie; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Batsukh, Baasansuren; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Bellee, Violaine; Belloli, Nicoletta; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Betti, Federico; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bezshyiko, Iaroslava; Bifani, Simone; Billoir, Pierre; Bird, Thomas; Birnkraut, Alex; Bitadze, Alexander; Bizzeti, Andrea; Blake, Thomas; Blanc, Frederic; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Boettcher, Thomas; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borgheresi, Alessio; Borghi, Silvia; Borisyak, Maxim; Borsato, Martino; Bossu, Francesco; Boubdir, Meriem; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Buchanan, Emma; Burr, Christopher; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Camboni, Alessandro; Campana, Pierluigi; Campora Perez, Daniel; Campora Perez, Daniel Hugo; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chatzikonstantinidis, Georgios; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chobanova, Veronika; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Costa Sobral, Cayo Mar; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dall'Occo, Elena; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Aguiar Francisco, Oscar; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Serio, Marilisa; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Demmer, Moritz; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Dijkstra, Hans; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dovbnya, Anatoliy; Dreimanis, Karlis; Dufour, Laurent; Dujany, Giulio; Dungs, Kevin; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Déléage, Nicolas; Easo, Sajan; Ebert, Marcus; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Farley, Nathanael; Farry, Stephen; Fay, Robert; Fazzini, Davide; Ferguson, Dianne; Fernandez Albor, Victor; Fernandez Prieto, Antonio; Ferrari, Fabio; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fini, Rosa Anna; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fleuret, Frederic; Fohl, Klaus; Fontana, Marianna; Fontanelli, Flavio; Forshaw, Dean Charles; Forty, Roger; Franco Lima, Vinicius; Frank, Markus; Frei, Christoph; Fu, Jinlin; Furfaro, Emiliano; Färber, Christian; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; Garcia Martin, Luis Miguel; García Pardiñas, Julián; Garra Tico, Jordi; Garrido, Lluis; Garsed, Philip John; Gascon, David; Gaspar, Clara; Gavardi, Laura; Gazzoni, Giulio; Gerick, David; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianì, Sebastiana; Gibson, Valerie; Girard, Olivier Göran; Giubega, Lavinia-Helena; Gizdov, Konstantin; Gligorov, V.V.; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gorelov, Igor Vladimirovich; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Griffith, Peter; Grillo, Lucia; Gruberg Cazon, Barak Raimond; Grünberg, Oliver; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Göbel, Carla; Hadavizadeh, Thomas; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; Hatch, Mark; He, Jibo; Head, Timothy; Heister, Arno; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hushchyn, Mikhail; Hussain, Nazim; Hutchcroft, David; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; Jiang, Feng; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Kariuki, James Mwangi; Karodia, Sarah; Kecke, Matthieu; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khairullin, Egor; Khanji, Basem; Khurewathanakul, Chitsanu; Kirn, Thomas; Klaver, Suzanne; Klimaszewski, Konrad; Koliiev, Serhii; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Kozachuk, Anastasiia; Kozeiha, Mohamad; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krokovny, Pavel; Kruse, Florian; Krzemien, Wojciech; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kuonen, Axel Kevin; Kurek, Krzysztof; Kvaratskheliya, Tengiz; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lanfranchi, Gaia; Langenbruch, Christoph; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Leflat, Alexander; Lefrançois, Jacques; Lefèvre, Regis; Lemaitre, Florian; Lemos Cid, Edgar; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Liu, Xuesong; Loh, David; Longstaff, Iain; Lopes, Jose; Lucchesi, Donatella; Lucio Martinez, Miriam; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Lusiani, Alberto; Lyu, Xiao-Rui; Machefert, Frederic; Maciuc, Florin; Maev, Oleg; Maguire, Kevin; Malde, Sneha; Malinin, Alexander; Maltsev, Timofei; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Marks, Jörg; Martellotti, Giuseppe; Martin, Morgan; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massacrier, Laure Marie; Massafferri, André; Matev, Rosen; Mathad, Abhijit; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; Meadows, Brian; Meier, Frank; Meissner, Marco; Melnychuk, Dmytro; Merk, Marcel; Merli, Andrea; Michielin, Emanuele; Milanes, Diego Alejandro; Minard, Marie-Noelle; Mitzel, Dominik Stefan; Mogini, Andrea; Molina Rodriguez, Josue; Monroy, Ignacio Alberto; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Mulder, Mick; Mussini, Manuel; Müller, Dominik; Müller, Janine; Müller, Katharina; Müller, Vanessa; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nandi, Anita; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen-Mau, Chung; Nieswand, Simon; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Oldeman, Rudolf; Onderwater, Gerco; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Aranzazu; Pais, Preema Rennee; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parker, William; Parkes, Christopher; Passaleva, Giovanni; Pastore, Alessandra; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantinos; Petrolini, Alessandro; Petrov, Aleksandr; Petruzzo, Marco; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pikies, Malgorzata; Pinci, Davide; Pistone, Alessandro; Piucci, Alessio; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Pomery, Gabriela Johanna; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Poslavskii, Stanislav; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rama, Matteo; Ramos Pernas, Miguel; Rangel, Murilo; Raniuk, Iurii; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; dos Reis, Alberto; Remon Alepuz, Clara; Renaudin, Victor; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vicente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Rogozhnikov, Alexey; Roiser, Stefan; Romanovskiy, Vladimir; Romero Vidal, Antonio; Ronayne, John William; Rotondo, Marcello; Rudolph, Matthew Scott; Ruf, Thomas; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sadykhov, Elnur; Sagidova, Naylya; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santimaria, Marco; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schael, Stefan; Schellenberg, Margarete; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmelzer, Timon; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schubert, Konstantin; Schubiger, Maxime; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Sergi, Antonino; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Siddi, Benedetto Gianluca; Silva Coutinho, Rafael; Silva de Oliveira, Luiz Gustavo; Simi, Gabriele; Simone, Saverio; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Eluned; Smith, Iwan Thomas; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Stefko, Pavol; Stefkova, Slavorima; Steinkamp, Olaf; Stemmle, Simon; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Syropoulos, Vasileios; Szczekowski, Marek; Szumlak, Tomasz; T'Jampens, Stephane; Tayduganov, Andrey; Tekampe, Tobias; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tilley, Matthew James; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Toriello, Francis; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Traill, Murdo; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tully, Alison; Tuning, Niels; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valat, Sebastien; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vecchi, Stefania; van Veghel, Maarten; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Venkateswaran, Aravindhan; Vernet, Maxime; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Volkov, Vladimir; Vollhardt, Achim; Voneki, Balazs; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Vázquez Sierra, Carlos; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wang, Jianchun; Ward, David; Wark, Heather Mckenzie; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wicht, Jean; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Williams, Timothy; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wraight, Kenneth; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yin, Hang; Yu, Jiesheng; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zarebski, Kristian Alexander; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhang, Yu; Zhelezov, Alexey; Zheng, Yangheng; Zhokhov, Anatoly; Zhu, Xianglei; Zhukov, Valery; Zucchelli, Stefano

    2016-10-05

    The Dalitz plot analysis technique is used to study the resonant substructures of $B^- \\to D^+ \\pi^- \\pi^-$ decays in a data sample corresponding to 3.0 fb$^-1$ of $pp$ collision data recorded by the LHCb experiment during 2011 and 2012. A model-independent analysis of the angular moments demonstrates the presence of resonances with spins 1, 2 and 3 at high $D^+\\pi^-$ mass. The data are fitted with an amplitude model composed of a quasi-model-independent function to describe the $D^+\\pi^-$ S-wave together with virtual contributions from the $D^*(2007)^{0}$ and $B^{*0}$ states, and components corresponding to the $D^*_2(2460)^{0}$, $D^*_1(2680)^{0}$, $D^*_3(2760)^{0}$ and $D^*_2(3000)^{0}$ resonances. The masses and widths of these resonances are determined together with the branching fractions for their production in $B^- \\to D^+ \\pi^- \\pi^-$ decays. The $D^+\\pi^-$ S-wave has phase motion consistent with that expected due to the presence of the $D^*_0(2400)^{0}$ state. These results constitute the first obser...

  5. 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

  6. 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)

  7. 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

  8. Cluster decay analysis and related structure effects of fissionable ...

    Indian Academy of Sciences (India)

    2015-08-01

    Aug 1, 2015 ... Collective clusterization approach of dynamical cluster decay model (DCM) has been ... fusion–fission process resulting in the emission of symmetric and/or ... represents the relative separation distance between two fragments or clusters ... decay constant λ or decay half-life T1/2 is defined as λ = (ln 2/T1/2) ...

  9. 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.

  10. A phenomenological analysis of non-resonant charm meson decays

    International Nuclear Information System (INIS)

    Bediaga, I.; Goebel, C.; Mendez-Galain, R.

    1997-07-01

    We analyse the consequences of the usual assumption of a constant function to fit non-resonant decays from experimental Dalitz plot describing charmed meson decays. We first show, using the D + -> K 0 π + π 0 decay channel as an example, how an inadequate extraction of the non-resonant contribution could yield incorrect measurements for the resonant channels. We analyse how the correct study of this decay will provide a test for the validity of factorization in D meson decays. Finally, we show how form factors that can be measured from the D + s -> π - π + π + decay. We emphasize its relevance for the study of the decay τ -> v t 3π and the extraction of the α 1 meson width. (author)

  11. A phenomenological analysis of non-resonant charm meson decays

    Energy Technology Data Exchange (ETDEWEB)

    Bediaga, I.; Goebel, C. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Mendez-Galain, R. [Montevideo Univ. (Uruguay). Facultad de Ingenieria

    1997-07-01

    We analyse the consequences of the usual assumption of a constant function to fit non-resonant decays from experimental Dalitz plot describing charmed meson decays. We first show, using the D{sup +} -> K{sup 0} {pi}{sup +} {pi}{sup 0} decay channel as an example, how an inadequate extraction of the non-resonant contribution could yield incorrect measurements for the resonant channels. We analyse how the correct study of this decay will provide a test for the validity of factorization in D meson decays. Finally, we show how form factors that can be measured from the D{sup +}{sub s} -> {pi}{sup -} {pi}{sup +} {pi}{sup +} decay. We emphasize its relevance for the study of the decay {tau} -> v{sub t} 3{pi} and the extraction of the {alpha}{sub 1} meson width. (author) 26 refs., 6 figs.

  12. 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)

  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. 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

  15. Weak decays of doubly heavy baryons. SU(3) analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Wei; Xing, Zhi-Peng; Xu, Ji [Shanghai Jiao Tong University, INPAC, Shanghai Key Laboratory for Particle Physics and Cosmology, School of Physics and Astronomy, Shanghai (China)

    2017-11-15

    Motivated by the recent LHCb observation of doubly charmed baryon Ξ{sub cc}{sup ++} in the Λ{sub c}{sup +}K{sup -}π{sup +}π{sup +} final state, we analyze the 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} under the flavor SU(3) symmetry. The decay amplitudes for various semileptonic and nonleptonic decays are parametrized in terms of a few SU(3) irreducible amplitudes. We find a number of relations or sum rules between decay widths and CP asymmetries, which can be examined in future measurements at experimental facilities like LHC, Belle II and CEPC. Moreover, once a few decay branching fractions have been measured in the future, some of these relations may provide hints for exploration of new decay modes. (orig.)

  16. Dalitz plot analysis of D→Kππ decays

    International Nuclear Information System (INIS)

    Anjos, J.C.; Appel, J.A.; Bean, A.; Bracker, S.B.; Browder, T.E.; Cremaldi, L.M.; Danner, G.M.; Duboscq, J.; Elliott, J.R.; Escobar, C.O.; Gibney, M.C.; Gordon, A.S.; Hartner, G.F.; Karchin, P.E.; Kumar, B.R.; Losty, M.J.; Luste, G.J.; Mantsch, P.M.; Martin, J.F.; McHugh, S.; Menary, S.R.; Morrison, R.J.; Nash, T.; Ong, P.; Pinfold, J.; Punkar, G.; Purohit, M.V.; Raab, J.R.; Ross, W.R.; Santoro, A.F.S.; Shoup, A.L.; Sidhu, J.S.; Sliwa, K.; Sokoloff, M.D.; Souza, M.H.G.; Streetman, M.E.; Stundzia, A.B.; Volkmuth, W.D.; Witherell, M.S.

    1993-01-01

    Decays of the D 0 meson to K - π + π 0 and bar K 0 π + π - and of the D + to K - π + π + have been analyzed for resonant substructure. We present results on the amplitudes and phases of each decay mode and compare the results with other measurements. We confirm the highly nonresonant nature of the D + to K - π + π + decays. There is general agreement with theoretical models for the branching ratios measured

  17. Transient Analysis of a Magnetic Heat Pump

    Science.gov (United States)

    Schroeder, E. A.

    1985-01-01

    An experimental heat pump that uses a rare earth element as the refrigerant is modeled using NASTRAN. The refrigerant is a ferromagnetic metal whose temperature rises when a magnetic field is applied and falls when the magnetic field is removed. The heat pump is used as a refrigerator to remove heat from a reservoir and discharge it through a heat exchanger. In the NASTRAN model the components modeled are represented by one-dimensional ROD elements. Heat flow in the solids and fluid are analyzed. The problem is mildly nonlinear since the heat capacity of the refrigerant is temperature-dependent. One simulation run consists of a series of transient analyses, each representing one stroke of the heat pump. An auxiliary program was written that uses the results of one NASTRAN analysis to generate data for the next NASTRAN analysis.

  18. 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)

  19. Isospin analysis of CP asymmetries in B decays

    International Nuclear Information System (INIS)

    Gronau, M.; London, D.

    1990-09-01

    There is some theoretical uncertainty in the predictions for CP violating hadronic asymmetries in neutral B decays to CP eigenstates due to the existence of penguin diagrams. Using isospin relatins, we show that it is possible to remove this uncertainty for the decays B d 0 → ππ, up to a 4-fold ambiguity. (orig.)

  20. 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...

  1. Mars Exploration Rover Heat Shield Recontact Analysis

    Science.gov (United States)

    Raiszadeh, Behzad; Desai, Prasun N.; Michelltree, Robert

    2011-01-01

    The twin Mars Exploration Rover missions landed successfully on Mars surface in January of 2004. Both missions used a parachute system to slow the rover s descent rate from supersonic to subsonic speeds. Shortly after parachute deployment, the heat shield, which protected the rover during the hypersonic entry phase of the mission, was jettisoned using push-off springs. Mission designers were concerned about the heat shield recontacting the lander after separation, so a separation analysis was conducted to quantify risks. This analysis was used to choose a proper heat shield ballast mass to ensure successful separation with low probability of recontact. This paper presents the details of such an analysis, its assumptions, and the results. During both landings, the radar was able to lock on to the heat shield, measuring its distance, as it descended away from the lander. This data is presented and is used to validate the heat shield separation/recontact analysis.

  2. Analysis of super-allowed Fermi beta-decay

    International Nuclear Information System (INIS)

    Wilkinson, D.H.

    1994-11-01

    Analysis of tile Jπ = 0 + → 0 + super-allowed Fermi transitions within isospin triplets is limited in the precision of its outcome not by the accuracy of the experimental input data nor by the confidence with which the radiative corrections can be applied but by knowledge of the nuclear mismatch: the subversion of the isospin symmetry along the multiplets by the charge-dependence of the forces, both Coulomb and specifically nuclear. Theoretical estimates of the mismatch differ considerably from author to author, their direct application results in clear violation of tile hypothesis of conservation of the vector current and clear inconsistency with unitarity of the Cabibbo-Kobayashi-Maskawa matrix. This paper pursues and elaborates the previous suggestion that, in these unsatisfactory circumstances, the best procedure is to look to the experimental data themselves to determine and eliminate the mismatch by appropriate extrapolation to Z=O. This is done: (i) without any prior correction for mismatch; (ii) after correction for the full theoretical mismatch; (iii) after correction for case-to-case fluctuations in the theoretical mismatch. These three procedures are individually statistically satisfactory and mutually consistent in their extrapolation to Z = 0 despite the variety of the theoretical mismatches on which, in varying degrees, they are based. The resultant unitarity test for the CKM matrix is IV ud I 2 + IV us I 2 + IV ub I 2 = 1.0003 ± 0.0014. The associated value for the operational vector coupling constant is: G v * / (hc) 3 = (1.15155±0.00064) x 10 -5 GeV -2 . If unitarity of the CKM matrix is alternatively assumed one may conclude, from a similar analysis, that the mean charge of the fermionic fields between which beta-decay takes place is Q-bar = 0.172±0.060 and that, at the 90% confidence level, b F -3 were b F is the relative effective scalar coupling constant. Neutron decay is also discussed, with the provisional recommendations: G A * /(hc) 3

  3. Updated analysis of some two-body charmless B decays

    International Nuclear Information System (INIS)

    Chiang Chengwei; Rosner, Jonathan L.

    2002-01-01

    New data from the BaBar, Belle, and CLEO Collaborations on B decays to two-body charmless final states are analyzed, with the following consequences. (1) The penguin amplitude which dominates the decay B + →π + K *0 has a magnitude similar to that dominating B + →π + K 0 . (2) The decay B + →π + η, a good candidate for observing direct CP violation, should be detectable at present levels of sensitivity. (3) The decays B + →η ' K + and B + →ηK* + are sufficiently similar in rate to the corresponding decays B 0 →η ' K 0 and B 0 →ηK* 0 , respectively, that one cannot yet infer the need for 'tree' amplitudes t ' contributing to the B + but not the B 0 decays. Statistical requirements for observing this and other examples of tree-penguin interference are given. (4) Whereas the B + →η ' K + and B 0 →η ' K 0 rates cannot be accounted for by the penguin amplitude p ' alone but require an additional flavor-singlet penguin contribution s ' , no such flavor-singlet penguin contribution is yet called for in the decays B + →ηK* + or B 0 →ηK *0 . Predictions for the rates for B + →η ' K* + and B 0 →η ' K* 0 are given which would allow one to gauge the importance of these flavor-singlet penguin amplitudes

  4. 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)

  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. 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

  7. E691 Dalitz plot analysis of Λc+ → pK-π+ decays

    International Nuclear Information System (INIS)

    Cremaldi, L.

    1992-01-01

    The resonant decays of the Λ c + can give insight into the weak decay mechanism of this charmed baryon. We have performed a Dalitz plot analysis of the final state for a clean sample Λ c + → pK - π + decays observed in photoproduction experiment E691 at Fermilab. We measure Γ(pK*(892)degrees)/Γ (pK - π + ) = .35 ± .09 ± .07. We also observe an unexpected enhancement at low pK- invariant mass

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  9. 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)

  10. 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)

  11. 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)

  12. Comparison of decay heat exchangers placing in the primary circuit of pool type fast reactor

    International Nuclear Information System (INIS)

    Birbraer, P.N.; Gorbunov, V.S.; Zotov, V.G.; Kuzavkov, N.G.; Pykhonin, V.A.; Sobolev, V.A.; Ryzhov, V.A.

    1993-01-01

    Description of two alternative arrangements of decay beat exchangers (DHXs) in the fast reactor tank is presented: in 'hot' cavity and in 'cold' cavity. The results of calculation for the two alternative arrangements as regards static and dynamic parameters in the primary circuit on 1-D program are given. (author)

  13. ICRF heating analysis on ASDEX plasmas

    International Nuclear Information System (INIS)

    Itoh, Sanae; Itoh, Kimitaka; Fukuyama, Atsushi; Morishita, Takayuki; Steinmetz, K.; Noterdaeme, J.-M.

    1988-01-01

    ICRF (ion cyclotron range of frequencies) waves heating in an ASDEX tokamak are analyzed. The excitation, propagation and absorption are studied by using a global wave code. This analysis is combined with a Fokker-Planck code. The waveform in the plasma, the loading resistance and the reactance of the antenna are calculated for both the minority ion heating and the second harmonic resonance heating. Attention is given to the change of the antenna loading associated with the L/H transition. Optimum conditions for the loading are discussed. In the minority heating case, the tail generation and thermalization are analyzed. Spatial profiles of the tail-ion temperature and the power transferred to the bulk electrons and ions are obtained. Central as well as off-central heating cases are investigated. The effect of the reactive electric field is discussed in connection with rf losses and impurity production. (author)

  14. General partial wave analysis of the decay of a hyperon of spin 1/2

    International Nuclear Information System (INIS)

    Lee, T.D.; Yang, C.N.

    1983-01-01

    This note is to consider the general problem of the decay of a hyperon of spin 1/2 into a pion and a nucleon under the general assumption of possible violations of parity conservation, charge-conjugation invariance, and time-reversal invariance. The discussion is in essence a partial wave analysis of the decay phenomena and is independent of the dynamics of the decay. Nonrelativistic approximations are not made on either of the decay products. In the reference system in which the hyperon is at rest there are two possible final states of the pion-nucleon system:s/sub 1/2/ and p/sub 1/2/. Denoting the amplitudes of these two states by A and B, one observes that the decay is physically characterized by three real constants specifying the magnitudes and the relative phase between these amplitudes. One of these constants can be taken to be absolute value a 2 + absolute value B 2 , and is evidently proportional to the decay probability per unit time. The other two constants are best defined in terms of experimentally measurable quantities. They discuss three types of experiments: (a) The angular distribution of the decay pion from a completely polarized hyperon at rest. (b) The longitudinal polarization of the nucleon emitted in the decay of unpolarized hyperons at rest. (c) Transverse polarization of the nucleon emitted in a given direction in the decay of a polarized hyperon

  15. 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.

  16. 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.

  17. 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

  18. Analysis of super-allowed Fermi beta-decay

    Energy Technology Data Exchange (ETDEWEB)

    Wilkinson, D H [TRIUMF, Vancouver, BC (Canada); [Sussex Univ., Brighton (United Kingdom)

    1994-11-01

    Analysis of tile J{pi} = 0{sup +} {yields} 0{sup +} super-allowed Fermi transitions within isospin triplets is limited in the precision of its outcome not by the accuracy of the experimental input data nor by the confidence with which the radiative corrections can be applied but by knowledge of the nuclear mismatch: the subversion of the isospin symmetry along the multiplets by the charge-dependence of the forces, both Coulomb and specifically nuclear. Theoretical estimates of the mismatch differ considerably from author to author, their direct application results in clear violation of tile hypothesis of conservation of the vector current and clear inconsistency with unitarity of the Cabibbo-Kobayashi-Maskawa matrix. This paper pursues and elaborates the previous suggestion that, in these unsatisfactory circumstances, the best procedure is to look to the experimental data themselves to determine and eliminate the mismatch by appropriate extrapolation to Z=O. This is done: (i) without any prior correction for mismatch; (ii) after correction for the full theoretical mismatch; (iii) after correction for case-to-case fluctuations in the theoretical mismatch. These three procedures are individually statistically satisfactory and mutually consistent in their extrapolation to Z = 0 despite the variety of the theoretical mismatches on which, in varying degrees, they are based. The resultant unitarity test for the CKM matrix is IV{sub ud}I{sup 2} + IV{sub us}I{sup 2} + IV{sub ub}I{sup 2} = 1.0003 {+-} 0.0014. The associated value for the operational vector coupling constant is: G{sub v}{sup *} / (hc){sup 3} = (1.15155{+-}0.00064) x 10{sup -5} GeV{sup -2}. If unitarity of the CKM matrix is alternatively assumed one may conclude, from a similar analysis, that the mean charge of the fermionic fields between which beta-decay takes place is Q-bar = 0.172{+-}0.060 and that, at the 90% confidence level, b{sub F} < 2.6 x 10{sup -3} were b{sub F} is the relative effective scalar

  19. On the Time-Dependent Analysis of Gamow Decay

    Science.gov (United States)

    Durr, Detlef; Grummt, Robert; Kolb, Martin

    2011-01-01

    Gamow's explanation of the exponential decay law uses complex "eigenvalues" and exponentially growing "eigenfunctions". This raises the question, how Gamow's description fits into the quantum mechanical description of nature, which is based on real eigenvalues and square integrable wavefunctions. Observing that the time evolution of any…

  20. 172 ANALYSIS OF URBAN DECAY FROM LOW RESOLUTION ...

    African Journals Online (AJOL)

    temporal pattern of urban decay in different parts of a traditional organic city through ... c. Twenty percent or more of the residential structures were in need of major ... the body of the city which must be removed through surgical blades of urban ... Figure 1. Pay-off matrix of urban blight development. (Adapted from Sule 1980).

  1. 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

  2. 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.

  3. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    Directory of Open Access Journals (Sweden)

    N. Gierse

    2015-03-01

    The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100 μm are required.

  4. 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.

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

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Performance Analysis of Photovoltaic Water Heating System

    Directory of Open Access Journals (Sweden)

    Tomas Matuska

    2017-01-01

    Full Text Available Performance of solar photovoltaic water heating systems with direct coupling of PV array to DC resistive heating elements has been studied and compared with solar photothermal systems. An analysis of optimum fixed load resistance for different climate conditions has been performed for simple PV heating systems. The optimum value of the fixed load resistance depends on the climate, especially on annual solar irradiation level. Use of maximum power point tracking compared to fixed optimized load resistance increases the annual yield by 20 to 35%. While total annual efficiency of the PV water heating systems in Europe ranges from 10% for PV systems without MPP tracking up to 15% for system with advanced MPP trackers, the efficiency of solar photothermal system for identical hot water load and climate conditions is more than 3 times higher.

  8. 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

  9. Heat Transfer Analysis for a Fixed CST Column

    International Nuclear Information System (INIS)

    Lee, S.Y.

    2004-01-01

    In support of a small column ion exchange (SCIX) process for the Savannah River Site waste processing program, a transient two-dimensional heat transfer model that includes the conduction process neglecting the convection cooling mechanism inside the crystalline silicotitanate (CST) column has been constructed and heat transfer calculations made for the present design configurations. For this situation, a no process flow condition through the column was assumed as one of the reference conditions for the simulation of a loss-of-flow accident. A series of the modeling calculations has been performed using a computational heat transfer approach. Results for the baseline model indicate that transit times to reach 130 degrees Celsius maximum temperature of the CST-salt solution column are about 96 hours when the 20-in CST column with 300 Ci/liter heat generation source and 25 degrees Celsius initial column temperature is cooled by natural convection of external air as a primary heat transfer mechanism. The modeling results for the 28-in column equipped with water jacket systems on the external wall surface of the column and water coolant pipe at the center of the CST column demonstrate that the column loaded with 300 Ci/liter heat source can be maintained non-boiling indefinitely. Sensitivity calculations for several alternate column sizes, heat loads of the packed column, engineered cooling systems, and various ambient conditions at the exterior wall of the column have been performed under the reference conditions of the CST-salt solution to assess the impact of those parameters on the peak temperatures of the packed column for a given transient time. The results indicate that a water-coolant pipe at the center of the CST column filled with salt solution is the most effective one among the potential design parameters related to the thermal energy dissipation of decay heat load. It is noted that the cooling mechanism at the wall boundary of the column has significant

  10. 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).

  11. 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)

  12. 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

  13. 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)

  14. Isospin analysis of charmless B-meson decays

    Energy Technology Data Exchange (ETDEWEB)

    Charles, J. [CNRS, Aix Marseille Univ., Universite de Toulon, CPT, Marseille (France); Deschamps, O.; Niess, V. [CNRS/Universite Clermont Auvergne, UMR 6533, Laboratoire de Physique de Clermont, Aubiere (France); Descotes-Genon, S. [CNRS, Univ. Paris-Sud, Universite Paris-Saclay, Laboratoire de Physique Theorique (UMR 8627), Orsay (France)

    2017-08-15

    We discuss the determination of the CKM angle α using the non-leptonic two-body decays B → ππ, B → ρρ and B → ρπ using the latest data available. We illustrate the methods used in each case and extract the corresponding value of α. Combining all these elements, we obtain the determination α{sub dir} = (86.2{sub -4.0}{sup +4.4} union 178.4{sub -5.1}{sup +3.9}) {sup circle}. We assess the uncertainties associated to the breakdown of the isospin hypothesis and the choice of the statistical framework in detail. We also determine the hadronic amplitudes (tree and penguin) describing the QCD dynamics involved in these decays, briefly comparing our results with theoretical expectations. For each observable of interest in the B → ππ, B → ρρ and B → ρπ systems, we perform an indirect determination based on the constraints from all the other observables available and we discuss the compatibility between indirect and direct determinations. Finally, we review the impact of future improved measurements on the determination of α. (orig.)

  15. 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

  16. Analysis of the heat transfer in double and triple concentric tube heat exchangers

    Science.gov (United States)

    Rădulescu, S.; Negoiţă, L. I.; Onuţu, I.

    2016-08-01

    The tubular heat exchangers (shell and tube heat exchangers and concentric tube heat exchangers) represent an important category of equipment in the petroleum refineries and are used for heating, pre-heating, cooling, condensation and evaporation purposes. The paper presents results of analysis of the heat transfer to cool a petroleum product in two types of concentric tube heat exchangers: double and triple concentric tube heat exchangers. The cooling agent is water. The triple concentric tube heat exchanger is a modified constructive version of double concentric tube heat exchanger by adding an intermediate tube. This intermediate tube improves the heat transfer by increasing the heat area per unit length. The analysis of the heat transfer is made using experimental data obtained during the tests in a double and triple concentric tube heat exchanger. The flow rates of fluids, inlet and outlet temperatures of water and petroleum product are used in determining the performance of both heat exchangers. Principally, for both apparatus are calculated the overall heat transfer coefficients and the heat exchange surfaces. The presented results shows that triple concentric tube heat exchangers provide better heat transfer efficiencies compared to the double concentric tube heat exchangers.

  17. 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

  18. 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

  19. Radionuclide inventory and heat generation analysis in disposal of radioactive waste

    International Nuclear Information System (INIS)

    Suryanto

    1997-01-01

    Radionuclide inventory and heat generation analysis on spent nuclear fuel were done in order to study the potential radionuclides contributing radiological impact to human being caused by spent fuel disposal. The study was carried out using the Bateman equation of radionuclide decay chains for fission products and actinides. the results showed that Cs-137, Sr-90 and Pu-239 dominated inventory of spent fuel, in which Pu-238 and Pu-240 dominated heat generation during disposal. Accordingly, the above radionuclides could be considered as the reference radionuclides for safety analysis of spent nuclear fuel disposal (author)

  20. 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.

  1. 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.

  2. Heating and melting of small icy satellites by the decay of Al-26

    International Nuclear Information System (INIS)

    Prialnik, D.; Bar-Nun, A.

    1990-01-01

    The effect of radiogenic heating due to Al-26 on the thermal evolution of small icy satellites is studied. The object is to find the extent of internal melting as a function of the satellite radius and of the initial Al-26 abundance. The implicit assumption, based on observations of young stars, is that planet and satellite accretion occurred on a time scale of about 10 to the 6th yr (comparable with the lifetime of Al-26). The icy satellites are modeled as spheres of initially amorphous ice, with chondritic abundances of K-40, Th-232, U-235, and U-238, corresponding to an ice/dust mass ratio of 1. Evolutionary calculations are carried out, spanning 4.5 x 10 to the 9th yr, for different combinations of the two free parameters. Heat transfer by subsolidus convection is neglected for these small satellites. The main conclusion is that the initial Al-26 abundance capable of melting icy bodies of satellite size to a significant extent is more than 10 times lower than that prevailing in the interstellar medium (or that inferred from the Ca-Al rich inclusions of the Allende meteorite, about 7 x 10 to the -7th by mass). 34 refs

  3. Heating and melting of small icy satellites by the decay of Al-26

    Science.gov (United States)

    Prialnik, Dina; Bar-Nun, Akiva

    1990-05-01

    The effect of radiogenic heating due to Al-26 on the thermal evolution of small icy satellites is studied. The object is to find the extent of internal melting as a function of the satellite radius and of the initial Al-26 abundance. The implicit assumption, based on observations of young stars, is that planet and satellite accretion occurred on a time scale of about 10 to the 6th yr (comparable with the lifetime of Al-26. The icy satellites are modeled as spheres of initially amorphous ice, with chondritic abundances of K-40, Th-232, U-235, and U-238, corresponding to an ice/dust mass ratio of 1. Evolutionary calculations are carried out, spanning 4.5 x 10 to the 9th yr, for different combinations of the two free parameters. Heat transfer by subsolidus convection is neglected for these small satellites. The main conclusion is that the initial Al-26 abundance capable of melting icy bodies of satellite size to a significant extent is more than 10 times lower than that prevailing in the interstellar medium (or that inferred from the Ca-Al rich inclusions of the Allende meteorite, about 7 x 10 to the -7th by mass).

  4. Angular analysis and CP violation studies in B decays at CMS

    CERN Document Server

    Chen, Kai-Feng

    2014-01-01

    Angular distributions of the decay $B^0 \\to K^{*0} \\mu^+\\mu^-$ are a clean probe of physics beyond the standard model. In particular the forward-backward asymmetry and other variables which can be determined as a function of $q^2$, provide fruitful information from a single decay channel. An angular and proper decay time analysis is applied to the $B_s \\to J/\\psi \\phi$ decays. The $B_s$ signal candidates are reconstructed and used to extract the mixing phase $\\phi_s$. The theoretical prediction of the $\\phi_s$ angle is particularly robust, thus any deviation from the prediction can be a smoking gun signal of new physics.

  5. Treatment of the decay heat production in the reactor dynamics program TINTE

    International Nuclear Information System (INIS)

    Gerwin, H.; Scherer, W.

    1993-07-01

    The TINTE code system deals with the nuclear and the thermal transient behaviour of the primary circuit of an HTGR taking into consideration the mutual feedback effects in two-dimensional r-z-geometry. An update of the treatment of delayed heat production is presented. It is based on the German norm DIN 25485, the rules of which had to be adjusted for use in a dynamics code. For the description of the fuel element power history a substitute-histogram has been constructed from local burnup and optionally from information about shuffling of the fuel balls. As an example the depressurisation accident of a MODUL-HTR is calculated. The results obtained are very similiar to others previously reported. (orig./HP) [de

  6. SWIFT, 3-D Fluid Flow, Heat Transfer, Decay Chain Transport in Geological Media

    International Nuclear Information System (INIS)

    Cranwell, R.M.; Reeves, M.

    2003-01-01

    1 - Description of problem or function: SWIFT solves the coupled or individual equations governing fluid flow, heat transport, brine displacement, and radionuclide displacement in geologic media. Fluid flow may be transient or steady-state. One, two, or three dimensions are available and transport of radionuclides chains is possible. 4. Method of solution: Finite differencing is used to discretize the partial differential equations in space and time. The user may choose centered or backward spatial differencing, coupled with either central or backward temporal differencing. The matrix equations may be solved iteratively (two line successive-over-relaxation) or directly (special matrix banding and Gaussian elimination). 5. Restrictions on the complexity of the problem: On the CDC7600 in direct solution mode, the maximum number of grid blocks allowed is approximately 1400

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

    International Nuclear Information System (INIS)

    Tzanos, C.P.

    1997-01-01

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

  8. 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

  9. 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

  10. Statistical analysis of time-resolved emission from ensembles of semiconductor quantum dots: Interpretation of exponential decay models

    DEFF Research Database (Denmark)

    Van Driel, A.F.; Nikolaev, I.S.; Vergeer, P.

    2007-01-01

    We present a statistical analysis of time-resolved spontaneous emission decay curves from ensembles of emitters, such as semiconductor quantum dots, with the aim of interpreting ubiquitous non-single-exponential decay. Contrary to what is widely assumed, the density of excited emitters...... and the intensity in an emission decay curve are not proportional, but the density is a time integral of the intensity. The integral relation is crucial to correctly interpret non-single-exponential decay. We derive the proper normalization for both a discrete and a continuous distribution of rates, where every...... decay component is multiplied by its radiative decay rate. A central result of our paper is the derivation of the emission decay curve when both radiative and nonradiative decays are independently distributed. In this case, the well-known emission quantum efficiency can no longer be expressed...

  11. HEATING-7, Multidimensional Finite-Difference Heat Conduction Analysis

    International Nuclear Information System (INIS)

    2000-01-01

    1 - Description of program or function: HEATING 7.2i and 7.3 are the most recent developments in a series of heat-transfer codes and obsolete all previous versions distributed by RSICC as SCA-1/HEATING5 and PSR-199/HEATING 6. Note that Unix and PC versions of HEATING7 are available in the CCC-545/SCALE 4.4 package. HEATING can solve steady-state and/or transient heat conduction problems in one-, two-, or three-dimensional Cartesian, cylindrical, or spherical coordinates. A model may include multiple materials, and the thermal conductivity, density, and specific heat of each material may be both time- and temperature-dependent. The thermal conductivity may also be anisotropic. Materials may undergo change of phase. Thermal properties of materials may be input or may be extracted from a material properties library. Heat- generation rates may be dependent on time, temperature, and position, and boundary temperatures may be time- and position-dependent. The boundary conditions, which may be surface-to-environment or surface-to-surface, may be specified temperatures or any combination of prescribed heat flux, forced convection, natural convection, and radiation. The boundary condition parameters may be time- and/or temperature-dependent. General gray body radiation problems may be modeled with user-defined factors for radiant exchange. The mesh spacing may be variable along each axis. HEATING uses a run-time memory allocation scheme to avoid having to recompile to match memory requirements for each specific problem. HEATING utilizes free-form input. In June 1997 HEATING 7.3 was added to the HEATING 7.2i packages, and the Unix and PC versions of both 7.2i and 7.3 were merged into one package. HEATING 7.3 is being released as a beta-test version; therefore, it does not entirely replace HEATING 7.2i. There is no published documentation for HEATING 7.3; but a listing of input specifications, which reflects changes for 7.3, is included in the PSR-199 documentation. For 3-D

  12. Thermodynamic analysis of chemical heat pumps

    International Nuclear Information System (INIS)

    Obermeier, Jonas; Müller, Karsten; Arlt, Wolfgang

    2015-01-01

    Thermal energy storages and heat pump units represent an important part of high efficient renewable energy systems. By using thermally driven, reversible chemical reactions a combination of thermal energy storage and heat pump can be realized. The influences of thermophysical properties of the involved components on the efficiency of a heat pump cycle is analysed and the relevance of the thermodynamic driving force is worked out. In general, the behaviour of energetic and exergetic efficiency is contrary. In a real cycle, higher enthalpies of reaction decrease the energetic efficiency but increase the exergetic efficiency. Higher enthalpies of reaction allow for lower offsets from equilibrium state for a default thermodynamic driving force of the reaction. - Highlights: • A comprehensive efficiency analysis of gas-solid heat pumps is proposed. • Link between thermodynamic driving force and equilibrium drop is shown. • Calculation of the equilibrium drop based on thermochemical properties. • Reaction equilibria of the decomposition reaction of salt hydrates. • Contrary behavior of energetic and exergetic efficiency

  13. Analysis of the radiative decay J/ψ → γωω at BESIII

    Energy Technology Data Exchange (ETDEWEB)

    Albrecht, Malte [Institut fuer Experimentalphysik I, Ruhr-Universitaet Bochum (Germany)

    2015-07-01

    Gluonic bound states like glueballs or hybrids are predicted to be copiously produced in the radiative decays of J/ψ. Especially the radiative decays into two vector mesons, J/ψ → γVV, have been intensively studied, since pseudoscalar (0{sup -}) enhancements have been observed in the ρρ, φφ and ωω final states with low statistics in the past. In this talk we show preliminary results of the analysis of the radiative decay J/ψ → γωω, using the world's largest data sample of ∼ 1.2.10{sup 9} J/ψ events collected with the Beijing Spectrometer III (BESIII) at the Beijing Electron-Positron Collider (BEPCII). The data selection, background analysis and preliminary results of a mass-independent partial wave analysis of the selected sample are presented.

  14. Simultaneous determination of radionuclides separable into natural decay series by use of time-interval analysis

    International Nuclear Information System (INIS)

    Hashimoto, Tetsuo; Sanada, Yukihisa; Uezu, Yasuhiro

    2004-01-01

    A delayed coincidence method, time-interval analysis (TIA), has been applied to successive α-α decay events on the millisecond time-scale. Such decay events are part of the 220 Rn→ 216 Po (T 1/2 145 ms) (Th-series) and 219 Rn→ 215 Po (T 1/2 1.78 ms) (Ac-series). By using TIA in addition to measurement of 226 Ra (U-series) from α-spectrometry by liquid scintillation counting (LSC), two natural decay series could be identified and separated. The TIA detection efficiency was improved by using the pulse-shape discrimination technique (PSD) to reject β-pulses, by solvent extraction of Ra combined with simple chemical separation, and by purging the scintillation solution with dry N 2 gas. The U- and Th-series together with the Ac-series were determined, respectively, from alpha spectra and TIA carried out immediately after Ra-extraction. Using the 221 Fr→ 217 At (T 1/2 32.3 ms) decay process as a tracer, overall yields were estimated from application of TIA to the 225 Ra (Np-decay series) at the time of maximum growth. The present method has proven useful for simultaneous determination of three radioactive decay series in environmental samples. (orig.)

  15. HEAT TRANSFER ANALYSIS FOR FIXED CST AND RF COLUMNS

    International Nuclear Information System (INIS)

    Lee, S

    2007-01-01

    In support of a small column ion exchange (SCIX) process for the Savannah River Site waste processing program, transient and steady state two-dimensional heat transfer models have been constructed for columns loaded with cesium-saturated crystalline silicotitanate (CST) or spherical Resorcinol-Formaldehyde (RF) beads and 6 molar sodium tank waste supernate. Radiolytic decay of sorbed cesium results in heat generation within the columns. The models consider conductive heat transfer only with no convective cooling and no process flow within the columns (assumed column geometry: 27.375 in ID with a 6.625 in OD center-line cooling pipe). Heat transfer at the column walls was assumed to occur by natural convection cooling with 35 C air. A number of modeling calculations were performed using this computational heat transfer approach. Minimal additional calculations were also conducted to predict temperature increases expected for salt solution processed through columns of various heights at the slowest expected operational flow rate of 5 gpm. Results for the bounding model with no process flow and no active cooling indicate that the time required to reach the boiling point of ∼130 C for a CST-salt solution mixture containing 257 Ci/liter of Cs-137 heat source (maximum expected loading for SCIX applications) at 35 C initial temperature is about 6 days. Modeling results for a column actively cooled with external wall jackets and the internal coolant pipe (inlet coolant water temperature: 25 C) indicate that the CST column can be maintained non-boiling under these conditions indefinitely. The results also show that the maximum temperature of an RF-salt solution column containing 133 Ci/liter of Cs-137 (maximum expected loading) will never reach boiling under any conditions (maximum predicted temperature without cooling: 88 C). The results indicate that a 6-in cooling pipe at the center of the column provides the most effective cooling mechanism for reducing the maximum

  16. A method of the sensitivity analysis of build-up and decay of actinides

    International Nuclear Information System (INIS)

    Mitani, Hiroshi; Koyama, Kinji; Kuroi, Hideo

    1977-07-01

    To make sensitivity analysis of build-up and decay of actinides, mathematical methods related to this problem have been investigated in detail. Application of time-dependent perturbation technique and Bateman method to sensitivity analysis is mainly studied. For the purpose, a basic equation and its adjoint equation for build-up and decay of actinides are systematically solved by introducing Laplace and modified Laplace transforms and their convolution theorems. Then, the mathematical method of sensitivity analyses is formulated by the above technique; its physical significance is also discussed. Finally, application of eigenvalue-method is investigated. Sensitivity coefficients can be directly calculated by this method. (auth.)

  17. Lack of endogenous modulation and reduced decay of prolonged heat pain in older adults.

    Science.gov (United States)

    Riley, Joseph L; King, Christopher D; Wong, Fong; Fillingim, Roger B; Mauderli, Andre P

    2010-07-01

    This study supports the hypothesis that healthy older adults exhibit decreased endogenous pain inhibition compared to younger healthy controls. Twenty-two older adults (56-77years of age) and 27 controls aged 20-49 participated in five experimental sessions following a training session. Each experimental session consisted of five 60-s trials in which the experimental heat stimulus was presented to the thenar eminence of the left palm with or without a conditioning stimulus (cold-water immersion of the foot). The temperature for the palm (44-49 degrees C) and foot (8-16 degrees C) was customized for each subject. The intensity of experimental pain produced by the contact thermode was continuously measured during the 60-s trial with an electronic visual analogue scale. No significant associations were found between subjects rating of concentration and the overall inhibitory effect. Older subjects failed to demonstrate conditioned pain modulation (CPM) and showed facilitation in the trials using painful concurrent immersion of the foot. A novel aspect of the study was that we recorded "pain offset" (i.e., after-sensations) and found that ratings for the older sample decreased at a slower rate than observed for the group of younger adults suggesting increased central sensitization among the older sample. Decrements in CPM could contribute to the greater prevalence of pain in older age. Since a number of neurotransmitter systems are involved in pain modulation, it is possible age-related differences in CPM are due to functional changes in these systems in a number of areas within the neuroaxis. Copyright 2010 International Association for the Study of Pain. Published by Elsevier B.V. All rights reserved.

  18. Spatiotemporal and economic analysis of industrial excess heat as a resource for district heating

    DEFF Research Database (Denmark)

    Bühler, Fabian; Petrović, Stefan; Holm, Fridolin Müller

    2018-01-01

    detailed analysis of the most suitable types of industries and the costs is required, allowing a targeted exploitation of this resource. This work extends the spatial and thermodynamic analysis, to account for the temporal match between industrial excess heat and district heating demands, as well...... as the costs for implementation and operation of the systems. This allows the determination of cost-effective district heating potentials, as well as the analysis of different industries and technological requirements. The results show that the temporal mismatch between excess heat and district heating demand...... and lack of demand, reduces the theoretical substitution potential by almost 30%. If heat storages are introduced, the total potential is reduced by only 10%. A majority of the excess heat can be utilised at socio-economic heating costs lower than the average Danish district heating price and the cost...

  19. 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.

  20. 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.

  1. Computer aided heat transfer analysis in a laboratory scaled heat exchanger unit

    International Nuclear Information System (INIS)

    Gunes, M.

    1998-01-01

    In this study. an explanation of a laboratory scaled heat exchanger unit and a software which is developed to analyze heat transfer. especially to use it in heat transfer courses, are represented. Analyses carried out in the software through sample values measured in the heat exchanger are: (l) Determination of heat transfer rate, logarithmic mean temperature difference and overall heat transfer coefficient; (2)Determination of convection heat transfer coefficient inside and outside the tube and the effect of fluid velocity on these; (3)Investigation of the relationship between Nusselt Number. Reynolds Number and Prandtl Number by using multiple non-linear regression analysis. Results are displayed on the screen graphically

  2. 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)

  3. 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.

  4. 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

  5. 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.

  6. Observation of the B0→ρ0ρ0 decay from an amplitude analysis of B0→(π+π−(π+π− decays

    Directory of Open Access Journals (Sweden)

    R. Aaij

    2015-07-01

    Full Text Available Proton–proton collision data recorded in 2011 and 2012 by the LHCb experiment, corresponding to an integrated luminosity of 3.0 fb−1, are analysed to search for the charmless B0→ρ0ρ0 decay. More than 600 B0→(π+π−(π+π− signal decays are selected and used to perform an amplitude analysis, under the assumption of no CP violation in the decay, from which the B0→ρ0ρ0 decay is observed for the first time with 7.1 standard deviations significance. The fraction of B0→ρ0ρ0 decays yielding a longitudinally polarised final state is measured to be fL=0.745−0.058+0.048(stat±0.034(syst. The B0→ρ0ρ0 branching fraction, using the B0→ϕK⁎(8920 decay as reference, is also reported as B(B0→ρ0ρ0=(0.94±0.17(stat±0.09(syst±0.06(BF×10−6.

  7. 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

  8. The neutral kaon decays to $\\pi^+ \\pi^- \\pi^0$ a detailed analysis of the CPLEAR data

    CERN Document Server

    Angelopoulos, Angelos; Aslanides, Elie; Backenstoss, Gerhard; Bargassa, P; Behnke, O; Benelli, A; Bertin, V; Blanc, F; Bloch, P; Carlson, P J; Carroll, M; Cawley, E; Chertok, M B; Danielsson, M; Dejardin, M; Derré, J; Ealet, A; Eleftheriadis, C; Faravel, L; Fassnacht, P; Fetscher, W; Fidecaro, Maria; Filipcic, A; Francis, D; Fry, J; Gabathuler, Erwin; Gamet, R; Gerber, H J; Go, A; Haselden, A; Hayman, P J; Henry-Coüannier, F; Hollander, R W; Jon-And, K; Kettle, P R; Kokkas, P; Kreuger, R; Le Gac, R; Leimgruber, F; Mandic, I; Manthos, N; Marel, Gérard; Mikuz, M; Miller, J; Montanet, François; Nakada, Tatsuya; Pagels, B; Papadopoulos, I M; Pavlopoulos, P; Polivka, G; Rickenbach, R; Roberts, B L; Ruf, T; Schäfer, M; Schaller, L A; Schietinger, T; Schopper, A; Tauscher, Ludwig; Thibault, C; Touchard, F; Touramanis, C; van Eijk, C W E; Vlachos, S; Weber, P; Wigger, O; Wolter, M; Zavrtanik, D; Zimmerman, D

    1998-01-01

    A detailed analysis of neutral kaons decaying to \\Pgpp \\Pgpm \\Pgpz\\ is presented based on the complete data set containing half a million events. Time-dependent decay rate asymmetries are measured between initially tagged \\PKz\\ and \\PaKz\\ and for different regions of the phase space. These asymmetries, resulting from the interference between the CP-conserving decay amplitude of \\PKzL\\ and the decay amplitude of \\PKzS\\ -- either CP-violating or CP-conserving -- allow the determinationof the \\PKzS\\ parameters \\etapmz\\ (CP-violating) and \\lampmz\\ (CP-conserving), and also of the main i sospin components of the \\PKzS\\ decay amplitude. The branching ratio of \\PKzS\\ $\\rightarrow$ \\Pgpp \\Pgpm \\Pgpz\\ (CP-conserving) is deduced directly from \\lampmz . In addition, we extract the slope parameters describing the energy dependence of the $\\PKzL \\rightarrow \\Pgpp \\Pgpm \\Pgpz $ Dalitz plot. The whole set of our results fits well within the current phenomenological picture of the neut ral-kaon system including CP violation ...

  9. Comparative analysis of nonperturbative effects in B→Xulνl decays

    International Nuclear Information System (INIS)

    Kniehl, Bernd A.; Kramer, Gustav; Yang Jifeng

    2007-01-01

    In order to extract the Cabibbo-Kobayashi-Maskawa matrix element |V ub | from B→X u lν l decays, the overwhelming background from B→X c lν l decays must be reduced by appropriate acceptance cuts. We study the nonperturbative effects due to the motion of the b quark inside the B meson on the phenomenologically relevant decay distributions of B→X u lν l in the presence of such cuts in a comparative analysis based on shape functions and the parton model in the light cone limit. Comparisons with recent data from the CLEO, BABAR, and BELLE collaborations favor the shape-function approach

  10. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

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

  11. β-delayed γ-proton decay in 56Zn: Analysis of the charged-particle spectrum

    International Nuclear Information System (INIS)

    Orrigo, S. E.A.; Ascher, P.; Cakirli, R. B.; Kozer, H. C.; Popescu, L.; Rogers, A. M.; Susoy, G.; Suzuki, T.

    2015-01-01

    A study of the β decay of the proton-rich T z = –2 nucleus 56 Zn has been reported in a recent publication. A rare and exotic decay mode, β-delayed γ-proton decay, has been observed there for the first time in the fp shell. Here, we expand on some of the details of the data analysis, focusing on the charged particle spectrum

  12. Observation of the ${B^0 \\to \\rho^0 \\rho^0}$ decay from an amplitude analysis of ${B^0 \\to (\\pi^+\\pi^-)(\\pi^+\\pi^-)}$ decays

    CERN Document Server

    Aaij, Roel; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casanova Mohr, Raimon; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Di Ruscio, Francesco; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferrari, Fabio; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gascon, David; Gaspar, Clara; Gastaldi, Ugo; Gauld, Rhorry; Gavardi, Laura; Gazzoni, Giulio; Geraci, Angelo; Gerick, David; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Gianì, Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, Vladimir; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lowdon, Peter; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathad, Abhijit; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Mitzel, Dominik Stefan; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Osorio Rodrigues, Bruno; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Aranzazu; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skillicorn, Ian; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Sterpka, Christopher Francis; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Todd, Jacob; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viana Barbosa, Joao Vitor; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wiedner, Dirk; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang

    2015-01-01

    Proton-proton collision data recorded in 2011 and 2012 by the LHCb experiment, corresponding to an integrated luminosity of 3.0 fb$^{-1}$i, are analysed to search for the charmless ${B^0 \\to \\rho^0 \\rho^0}$ decay. More than 600 ${B^0 \\to (\\pi^+\\pi^-)(\\pi^+\\pi^-)}$ signal decays are selected and used to perform an amplitude analysis from which the ${B^0 \\to \\rho^0 \\rho^0}$ decay is observed for the first time with 7.1 standard deviations significance. The fraction of ${B^0 \\to \\rho^0 \\rho^0}$ decays yielding a longitudinally polarised final state is measured to be $fL = 0.745^{+0.048}_{-0.058} ({\\rm stat}) \\pm 0.034 ({\\rm syst})$. The ${B^0 \\to \\rho^0 \\rho^0}$ branching fraction, using the ${B^0 \\to \\phi K^*(892)^{0}}$ decay as reference, is also reported as $\\mathcal B (B^0 \\to \\rho^0 \\rho^0) = (0.94 \\pm 0.17 ({\\rm stat}) \\pm 0.09 ({\\rm syst}) \\pm 0.06 ({\\rm BF})) \\times 10^{-6}$.

  13. Linear stability analysis of heated parallel channels

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Isbin, H.S.

    1982-01-01

    An analyis is presented of thermal hydraulic stability of flow in parallel channels covering the range from inlet subcooling to exit superheat. The model is based on a one-dimensional drift velocity formulation of the two phase flow conservation equations. The system of equations is linearized by assuming small disturbances about the steady state. The dynamic response of the system to an inlet flow perturbation is derived yielding the characteristic equation which predicts the onset of instabilities. A specific application is carried out for homogeneous and regional uniformly heated systems. The particular case of equal characteristic frequencies of two-phase and single phase vapor region is studied in detail. The D-partition method and the Mikhailov stability criterion are used for determining the marginal stability boundary. Stability predictions from the present analysis are compared with the experimental data from the solar test facility. 8 references

  14. Methods for the analysis of complex fluorescence decays: sum of Becquerel functions versus sum of exponentials

    International Nuclear Information System (INIS)

    Menezes, Filipe; Fedorov, Alexander; Baleizão, Carlos; Berberan-Santos, Mário N; Valeur, Bernard

    2013-01-01

    Ensemble fluorescence decays are usually analyzed with a sum of exponentials. However, broad continuous distributions of lifetimes, either unimodal or multimodal, occur in many situations. A simple and flexible fitting function for these cases that encompasses the exponential is the Becquerel function. In this work, the applicability of the Becquerel function for the analysis of complex decays of several kinds is tested. For this purpose, decays of mixtures of four different fluorescence standards (binary, ternary and quaternary mixtures) are measured and analyzed. For binary and ternary mixtures, the expected sum of narrow distributions is well recovered from the Becquerel functions analysis, if the correct number of components is used. For ternary mixtures, however, satisfactory fits are also obtained with a number of Becquerel functions smaller than the true number of fluorophores in the mixture, at the expense of broadening the lifetime distributions of the fictitious components. The quaternary mixture studied is well fitted with both a sum of three exponentials and a sum of two Becquerel functions, showing the inevitable loss of information when the number of components is large. Decays of a fluorophore in a heterogeneous environment, known to be represented by unimodal and broad continuous distributions (as previously obtained by the maximum entropy method), are also measured and analyzed. It is concluded that these distributions can be recovered by the Becquerel function method with an accuracy similar to that of the much more complex maximum entropy method. It is also shown that the polar (or phasor) plot is not always helpful for ascertaining the degree (and kind) of complexity of a fluorescence decay. (paper)

  15. Application of Abaqus to analysis of the temperature field in elements heated by moving heat sources

    Directory of Open Access Journals (Sweden)

    W. Piekarska

    2010-10-01

    Full Text Available Numerical analysis of thermal phenomena occurring during laser beam heating is presented in this paper. Numerical models of surface andvolumetric heat sources were presented and the influence of different laser beam heat source power distribution on temperature field wasanalyzed. Temperature field was obtained by a numerical solution the transient heat transfer equation with activity of inner heat sources using finite element method. Temperature distribution analysis in welded joint was performed in the ABAQUS/Standard solver. The DFLUXsubroutine was used for implementation of the movable welding heat source model. Temperature-depended thermophysical properties for steelwere assumed in computer simulations. Temperature distribution in laser beam surface heated and butt welded plates was numericallyestimated.

  16. Analysis of radiative heat transfer impact in cross-flow tube and fin heat exchangers

    Directory of Open Access Journals (Sweden)

    Hanuszkiewicz-Drapała Małgorzata

    2016-03-01

    Full Text Available A cross-flow, tube and fin heat exchanger of the water – air type is the subject of the analysis. The analysis had experimental and computational form and was aimed for evaluation of radiative heat transfer impact on the heat exchanger performance. The main element of the test facility was an enlarged recurrent segment of the heat exchanger under consideration. The main results of measurements are heat transfer rates, as well as temperature distributions on the surface of the first fin obtained by using the infrared camera. The experimental results have been next compared to computational ones coming from a numerical model of the test station. The model has been elaborated using computational fluid dynamics software. The computations have been accomplished for two cases: without radiative heat transfer and taking this phenomenon into account. Evaluation of the radiative heat transfer impact in considered system has been done by comparing all the received results.

  17. Analysis of decay of radicals induced in irradiated foods during long storage

    International Nuclear Information System (INIS)

    Kishita, Keigo; Kawamura, Shoei; Nakamura, Hideo; Ukai, Mitsuko; Kikuchi, Masahiro; Kobayashi, Yasuhiko

    2013-01-01

    By electron spin resonance (ESR) spectroscopy, we revealed free radicals in γ-ray irradiated foods ; black pepper, green coffee bean, cereal flour and ginseng. We also analyzed the decay behavior of radiation induced free radicals during long storage. The ESR spectrum of experimental irradiated foods consists of a sextet signal centered at g=2.0 and a singlet signal at the same g-value position and a singlet signal at g=4.0. The ESR spectrum of the cereal flour sample showed only singlet signal at g=2.0. The singlet signal at g=2.0 is originated from organic free radicals and its peak intensity showed the dependence of γ-ray radiation dose levels. But the signal intensity was decreased during storage. Only after 3 hours of radiation treatment the peak intensity was decreased fast and after that the intensity was decreased slowly. The analysis of radical decay process using the simulation methods based on the theory of reaction speed, the three decay behavior was showed. It is considered that at least three or more kinds of radicals were induced in irradiated foods and in decay during long time storage. (author)

  18. Dalitz Plot Analysis of $B^0_d \\to K+ \\pi^- \\pi^0$ Decays

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Zhitang; /Wisconsin U., Madison

    2006-05-05

    This thesis describes a Dalitz plot analysis of B{sup 0} {yields} K{sup +}{pi}{sup -}{pi}{sup 0} decays. The data sample comprises 213 million {Upsilon}(4S) {yields} B{bar B} decays collected with the BABAR detector at the PEP-II asymmetric-energy B Factory at the Stanford Linear Accelerator Center in California (SLAC). Preliminary results are presented for measurements of the inclusive branching fraction, quasi-two-body fractions and CP-violating charge asymmetries for intermediate states including K*(892){sup +}{pi}{sup -} and {rho}(770){sup -}K{sup +}. Observations of B{sup 0} decays to the K{pi} S-wave intermediate states, K*{sub 0}(1430){sup +}{pi}{sup -} and K*{sub 0}(1430){sup 0}{pi}{sup 0}, are reported. Evidence of the decay B{sup 0} {yields} K*(892){sup 0}{pi}{sup 0} is seen. We set upper limits at 90% confidence level on branching fractions of the nonresonant and other less significant intermediate states.

  19. Performance analysis of hybrid district heating system

    DEFF Research Database (Denmark)

    Mikulandric, Robert; Krajačić, Goran; Khavin, Gennadii

    2013-01-01

    District heating system could contribute to more efficient heat generation through cogeneration power plants or waste heat utilization facilities and to increase of renewable energy sources share in total energy consumption. In the most developed EU countries, renewable energy sources have been...... as problems related to transportation, storage and environmental impacts of biomass and waste utilisation. Implementation of heat storages in district heating systems could contribute to integration of intermittent energy sources. Hybridisation of heat production facility combines two or more different energy...... more extensively used in district heating systems either separately or as a supplement to traditional fossil fuels in order to achieve national energy policy objectives. However, they are still facing problems such as high intermittences, high energy production costs and low load factors as well...

  20. GIS based analysis of future district heating potential in Denmark

    DEFF Research Database (Denmark)

    Nielsen, Steffen; Möller, Bernd

    2013-01-01

    in Denmark have been mapped in a heat atlas which includes all buildings and their heat demands. This article focuses on developing a method for assessing the costs associated with supplying these buildings with DH. The analysis is based on the existing DH areas in Denmark. By finding the heat production...

  1. Statistical analysis of time-resolved emission from ensembles of semiconductor quantum dots: interpretations of exponantial decay models

    NARCIS (Netherlands)

    van Driel, A.F.; Nikolaev, I.; Vergeer, P.; Lodahl, P.; Vanmaekelbergh, D.; Vos, Willem L.

    2007-01-01

    We present a statistical analysis of time-resolved spontaneous emission decay curves from ensembles of emitters, such as semiconductor quantum dots, with the aim of interpreting ubiquitous non-single-exponential decay. Contrary to what is widely assumed, the density of excited emitters and the

  2. Angular analysis of the B-0 -> K*(0) mu(+) mu(-) decay using 3 fb(-1) of integrated luminosity

    NARCIS (Netherlands)

    Aaij, R.; Beteta, C. Abelian; Adeva, B.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Akar, S.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Cartelle, P. Alvarez; Alves, A. A.; Amato, S.; Arnerio, S.; Arnhis, Y.; An, L.; Anderlini, L.; Andreassi, G.; Andreotti, M.; Andrews, J. E.; Appleby, R. B.; Gutierrez, O. Aquines; Archilli, F.; d'Argent, P.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J. J.; Badalov, A.; Baesso, C.; Baldini, W.; Barlow, R. J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Battista, V.; Beaucourt, L.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Bel, L. J.; Onderwater, C. J. G.; Pellegrino, A.; Tolk, S.

    2016-01-01

    An angular analysis of the B-0 -> K*(0) (-> K+pi(-))mu(+)mu(-) decay is presented. The dataset corresponds to an integrated luminosity of 3.0 fb(-1) of pp collision data collected at the LHCb experiment. The complete angular information from the decay is used to determine CP-averaged observables and

  3. Technical assessment of electric heat boosters in low-temperature district heating based on combined heat and power analysis

    DEFF Research Database (Denmark)

    Cai, Hanmin; You, Shi; Wang, Jiawei

    2018-01-01

    This paper provides a technical assessment of electric heat boosters (EHBs) in low-energy districts. The analysis is based on a hypothetical district with 23 terraced single-family houses supplied by both a lowtemperature district heating (LTDH) network and a low-voltage network (LVN). Two case...

  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. 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)

  6. Comparative analysis of heat pump and biomass boiler for small detached house heating

    Directory of Open Access Journals (Sweden)

    Olkowski Tomasz

    2017-01-01

    Full Text Available The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

  7. Comparative analysis of heat pump and biomass boiler for small detached house heating

    Science.gov (United States)

    Olkowski, Tomasz; Lipiński, Seweryn; Olędzka, Aneta

    2017-10-01

    The purpose of the work is to answer the question - which of the two selected heat sources is more economically beneficial for small detached house: heat pump or biomass boiler fuelled with wood-pellets? The comparative analysis of these sources was carried out to discuss the issue. First, cost of both, equipment and operation of selected heat systems were analysed. Additionally, CO2 emission levels associated with these heat systems were determined. The comparative analysis of the costs of both considered heat systems showed that equipment cost of heat pump system is considerably bigger than the cost of biomass boiler system. The comparison of annual operation costs showed that heat pump operation cost is slightly lower than operation cost of biomass boiler. The analysis of above results shows that lower operation cost of heat pump in comparison with biomass boiler cost lets qualify heat pump as more economically justified only after 38 years of work. For both analysed devices, CO2 emission levels were determined. The considerations take into account the fact that heat pump consumes electricity. It is mostly generated through combustion of coal in Poland. The results show that in Poland biomass boiler can be described as not only more economically justified system but also as considerably more ecological.

  8. Thermal mechanical analysis of applications with internal heat generation

    Science.gov (United States)

    Govindarajan, Srisharan Garg

    The radioactive tracer Technetium-99m is widely used in medical imaging and is derived from its parent isotope Molybedenum-99 (Mo-99) by radioactive decay. The majority of Molybdenum-99 (Mo-99) produced internationally is extracted from high enriched uranium (HEU) dispersion targets that have been irradiated. To alleviate proliferation risks associated with HEU-based targets, the use of non-HEU sources is being mandated. However, the conversion of HEU to LEU based dispersion targets affects the Mo-99 available for chemical extraction. A possible approach to increase the uranium density, to recover the loss in Mo-99 production-per-target, is to use an LEU metal foil placed within an aluminum cladding to form a composite structure. The target is expected to contain the fission products and to dissipate the generated heat to the reactor coolant. In the event of interfacial separation, an increase in the thermal resistance could lead to an unacceptable rise in the LEU temperature and stresses in the target. The target can be deemed structurally safe as long as the thermally induced stresses are within the yield strength of the cladding and welds. As with the thermal and structural safety of the annular target, the thermally induced deflection of the BORALRTM-based control blades, used by the University of Missouri Research Reactor (MURRRTM ), during reactor operation has been analyzed. The boron, which is the neutron absorber in BORAL, and aluminum mixture (BORAL meat) and the aluminum cladding are bonded together through powder metallurgy to establish an adherent bonded plate. As the BORAL absorbs both neutron particles and gamma rays, there is volumetric heat generation and a corresponding rise in temperature. Since the BORAL meat and aluminum cladding materials have different thermal expansion coefficients, the blade may have a tendency to deform as the blade temperature changes and the materials expand at different rates. In addition to the composite nature of the

  9. Fluorescence decay data analysis correcting for detector pulse pile-up at very high count rates

    Science.gov (United States)

    Patting, Matthias; Reisch, Paja; Sackrow, Marcus; Dowler, Rhys; Koenig, Marcelle; Wahl, Michael

    2018-03-01

    Using time-correlated single photon counting for the purpose of fluorescence lifetime measurements is usually limited in speed due to pile-up. With modern instrumentation, this limitation can be lifted significantly, but some artifacts due to frequent merging of closely spaced detector pulses (detector pulse pile-up) remain an issue to be addressed. We propose a data analysis method correcting for this type of artifact and the resulting systematic errors. It physically models the photon losses due to detector pulse pile-up and incorporates the loss in the decay fit model employed to obtain fluorescence lifetimes and relative amplitudes of the decay components. Comparison of results with and without this correction shows a significant reduction of systematic errors at count rates approaching the excitation rate. This allows quantitatively accurate fluorescence lifetime imaging at very high frame rates.

  10. Analysis of heat transfer in plain carbon steels

    International Nuclear Information System (INIS)

    Han, Heung Nam; Lee, Kyung Jong

    1999-01-01

    During cooling of steels, the heat transfer was controlled by radiation, convection, conduction and heat evolution from phase transformation. To analyze the heat transfer during cooling precisely, the material constants such as density, heat capacity and the heat evolved during transformation were obtained as functions of temperature and chemical composition for each phase observed in plain carbon steel using a thermodynamic analysis based on the sublattice model of Fe-C-Mn system. The results were applied to 0.049 wt% and 0.155 wt% carbon steels with an austenitic stainless steel as reference by developing a proper heat transfer governing equation. The equation was solved using the lumped system method. In addition, using a transformation dilatometer with adequate experimental conditions to clarify the individual heat transfer effect, the transformation heat evolved during cooling and the transformation behavior as well as the temperature change were observed. The predicted temperature profiles during cooling were well agreed with the measured ones

  11. Heat demand mapping and district heating grid expansion analysis: Case study of Velika Gorica

    Directory of Open Access Journals (Sweden)

    Dorotić Hrvoje

    2017-01-01

    Full Text Available Highly efficient cogeneration and district heating systems have a significant potential for primary energy savings and the reduction of greenhouse gas emissions through the utilization of a waste heat and renewable energy sources. These potentials are still highly underutilized in most European countries. They also play a key role in the planning of future energy systems due to their positive impact on the increase of integration of intermittent renewable energy sources, for example wind and solar in a combination with power to heat technologies. In order to ensure optimal levels of district heating penetration into an energy system, a comprehensive analysis is necessary to determine the actual demands and the potential energy supply. Economical analysis of the grid expansion by using the GIS based mapping methods hasn’t been demonstrated so far. This paper presents a heat demand mapping methodology and the use of its output for the district heating network expansion analysis. The result are showing that more than 59% of the heat demand could be covered by the district heating in the city of Velika Gorica, which is two times more than the present share. The most important reason of the district heating's unfulfilled potential is already existing natural gas infrastructure.

  12. Heat demand mapping and district heating grid expansion analysis: Case study of Velika Gorica

    Science.gov (United States)

    Dorotić, Hrvoje; Novosel, Tomislav; Duić, Neven; Pukšec, Tomislav

    2017-10-01

    Highly efficient cogeneration and district heating systems have a significant potential for primary energy savings and the reduction of greenhouse gas emissions through the utilization of a waste heat and renewable energy sources. These potentials are still highly underutilized in most European countries. They also play a key role in the planning of future energy systems due to their positive impact on the increase of integration of intermittent renewable energy sources, for example wind and solar in a combination with power to heat technologies. In order to ensure optimal levels of district heating penetration into an energy system, a comprehensive analysis is necessary to determine the actual demands and the potential energy supply. Economical analysis of the grid expansion by using the GIS based mapping methods hasn't been demonstrated so far. This paper presents a heat demand mapping methodology and the use of its output for the district heating network expansion analysis. The result are showing that more than 59% of the heat demand could be covered by the district heating in the city of Velika Gorica, which is two times more than the present share. The most important reason of the district heating's unfulfilled potential is already existing natural gas infrastructure.

  13. 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

  14. Energy and exergy analysis of low temperature district heating network

    International Nuclear Information System (INIS)

    Li, Hongwei; Svendsen, Svend

    2012-01-01

    Low temperature district heating with reduced network supply and return temperature provides better match of the low quality building heating demand and the low quality heating supply from waste heat or renewable energy. In this paper, a hypothetical low temperature district heating network is designed to supply heating for 30 low energy detached residential houses. The network operational supply/return temperature is set as 55 °C/25 °C, which is in line with a pilot project carried out in Denmark. Two types of in-house substations are analyzed to supply the consumer domestic hot water demand. The space heating demand is supplied through floor heating in the bathroom and low temperature radiators in the rest of rooms. The network thermal and hydraulic conditions are simulated under steady state. A district heating network design and simulation code is developed to incorporate the network optimization procedure and the network simultaneous factor. Through the simulation, the overall system energy and exergy efficiencies are calculated and the exergy losses for the major district heating system components are identified. Based on the results, suggestions are given to further reduce the system energy/exergy losses and increase the quality match between the consumer heating demand and the district heating supply. -- Highlights: ► Exergy and energy analysis for low and medium temperature district heating systems. ► Different district heating network dimensioning methods are analyzed. ► Major exergy losses are identified in the district heating network and the in-house substations. ► Advantages to apply low temperature district heating are highlighted through exergy analysis. ► The influence of thermal by-pass on system exergy/energy performance is analyzed.

  15. Numerical Analysis for Heat Transfer Characteristics of Elliptic Fin-Tube Heat Exchanger with Various Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Hwan; Yoon, Jun Kyu [Gachon Univ., Seongnam (Korea, Republic of)

    2013-04-15

    In this study, the characteristics of the heat transfer coefficient and pressure drop were numerically analyzed according to the axis ratio (A R), pitch, location of vortex generator, and bump phase of the tube surface about an elliptical fin-tube heat exchanger. The boundary condition for CAD analysis was decided as a tube surface temperature of 348 K and inlet air velocity of 1.5 m/s. RCM 7th turbulent model was chosen as the numerical analysis for the sensitivity level. The analysis results indicated that the A R and transverse pitch decreased whereas the heat transfer coefficient increased. On the other hand, there was little difference in the longitudinal pitch. Furthermore, the heat transfer rate was more favorable when the vortex generator was located in front of the tube. Also, the bump phase of the tube surface indicated that the pressure drop and heat transfer were more favorable with the circle type than with the serrated type.

  16. Frequency and risk indicators of tooth decay among pregnant women in France: a cross-sectional analysis.

    Science.gov (United States)

    Vergnes, Jean-Noel; Kaminski, Monique; Lelong, Nathalie; Musset, Anne-Marie; Sixou, Michel; Nabet, Cathy

    2012-01-01

    Little is known on the prevalence of tooth decay among pregnant women. Better knowledge of tooth decay risk indicators during pregnancy could help to develop follow-up protocols for women at risk, along with better prevention strategies. The aim of this study was to assess the frequency of tooth decay and the number of decayed teeth per woman in a large sample of pregnant women in France, and to study associated risk indicators. A secondary cross-sectional analysis of data from a French multicentre case-control study was performed. The sample was composed of 1094 at-term women of six maternity units. A dental examination was carried out within 2 to 4 days post-partum. Socio-demographic and behavioural characteristics were obtained through a standardised interview with the women. Medical characteristics were obtained from the women's medical records. Risk indicators associated with tooth decay were identified using a negative binomial hurdle model. 51.6% of the women had tooth decay. The mean number of decayed teeth among women having at least one was 3.1 (s.d. = 2.8). Having tooth decay was statistically associated with lower age (aOR = 1.58, 95%CI [1.03,2.45]), lower educational level (aOR = 1.53, 95%CI [1.06,2.23]) and dental plaque (aOR = 1.75, 95%CI [1.27,2.41]). The number of decayed teeth was associated with the same risk indicators and with non-French nationality and inadequate prenatal care. The frequency of tooth decay and the number of decayed teeth among pregnant women were high. Oral health promotion programmes must continue to inform women and care providers about the importance of dental care before, during and after pregnancy. Future research should also assess the effectiveness of public policies related to oral health in target populations of pregnant women facing challenging social or economic situations.

  17. Heat transfer analysis of underground U-type heat exchanger of ground source heat pump system.

    Science.gov (United States)

    Pei, Guihong; Zhang, Liyin

    2016-01-01

    Ground source heat pumps is a building energy conservation technique. The underground buried pipe heat exchanging system of a ground source heat pump (GSHP) is the basis for the normal operation of an entire heat pump system. Computational-fluid-dynamics (CFD) numerical simulation software, ANSYS-FLUENT17.0 have been performed the calculations under the working conditions of a continuous and intermittent operation over 7 days on a GSHP with a single-well, single-U and double-U heat exchanger and the impact of single-U and double-U buried heat pipes on the surrounding rock-soil temperature field and the impact of intermittent operation and continuous operation on the outlet water temperature. The influence on the rock-soil temperature is approximately 13 % higher for the double-U heat exchanger than that of the single-U heat exchanger. The extracted energy of the intermittent operation is 36.44 kw·h higher than that of the continuous mode, although the running time is lower than that of continuous mode, over the course of 7 days. The thermal interference loss and quantity of heat exchanged for unit well depths at steady-state condition of 2.5 De, 3 De, 4 De, 4.5 De, 5 De, 5.5 De and 6 De of sidetube spacing are detailed in this work. The simulation results of seven working conditions are compared. It is recommended that the side-tube spacing of double-U underground pipes shall be greater than or equal to five times of outer diameter (borehole diameter: 180 mm).

  18. First observation and amplitude analysis of the $B^{-}\\to D^{+}K^{-}\\pi^{-}$ decay

    CERN Document Server

    Aaij, R.; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casanova Mohr, Raimon; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Di Ruscio, Francesco; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gascon, David; Gaspar, Clara; Gastaldi, Ugo; Gauld, Rhorry; Gavardi, Laura; Gazzoni, Giulio; Geraci, Angelo; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Gianì, Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, V.V.; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lowdon, Peter; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Osorio Rodrigues, Bruno; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Arantza; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skillicorn, Ian; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Sterpka, Christopher Francis; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Todd, Jacob; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viana Barbosa, Joao Vitor; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wiedner, Dirk; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Wilschut, Hans; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang

    2016-01-01

    The $B^{-}\\to D^{+}K^{-}\\pi^{-}$ decay is observed in a data sample corresponding to $3.0~\\rm{fb}^{-1}$ of $pp$ collision data recorded by the LHCb experiment during 2011 and 2012. Its branching fraction is measured to be ${\\cal B}(B^{-}\\to D^{+}K^{-}\\pi^{-}) = (7.31 \\pm 0.19 \\pm 0.22 \\pm 0.39) \\times 10^{-5}$ where the uncertainties are statistical, systematic and from the branching fraction of the normalisation channel $B^{-}\\to D^{+}\\pi^{-}\\pi^{-}$, respectively. An amplitude analysis of the resonant structure of the $B^{-}\\to D^{+}K^{-}\\pi^{-}$ decay is used to measure the contributions from quasi-two-body $B^{-}\\to D_{0}^{*}(2400)^{0}K^{-}$, $B^{-}\\to D_{2}^{*}(2460)^{0}K^{-}$, and $B^{-}\\to D_{J}^{*}(2760)^{0}K^{-}$ decays, as well as from nonresonant sources. The $D_{J}^{*}(2760)^{0}$ resonance is determined to have spin 1.

  19. Heat conduction analysis of multi-layered FGMs considering the finite heat wave speed

    International Nuclear Information System (INIS)

    Rahideh, H.; Malekzadeh, P.; Golbahar Haghighi, M.R.

    2012-01-01

    Highlights: ► Using a layerwise-incremental differential quadrature for heat transfer of FGMs. ► Superior accuracy with fewer degrees of freedom of the method with respect to FEM. ► Considering multi-layered functionally graded materials. ► Hyperbolic heat transfer analysis of thermal system with heat generation. ► Showing the effect of heat wave speed on thermal characteristic of the system. - Abstract: In this work, the heat conduction with finite wave heat speed of multi-layered domain made of functionally graded materials (FGMs) subjected to heat generation is simulated. For this purpose, the domain is divided into a set of mathematical layers, the number of which can be equal or greater than those of the physical layers. Then, in each mathematical layer, the non-Fourier heat transfer equations are employed. Since, the governing equations have variable coefficients due to FGM properties, as an efficient and accurate method the differential quadrature method (DQM) is adopted to discretize both spatial and temporal domains in each layer. This results in superior accuracy with fewer degrees of freedom than conventional finite element method (FEM). To verify this advantages through some comparison studies, a finite element solution are also obtained. After demonstrating the convergence and accuracy of the method, the effects of heat wave speed for two different set of boundary conditions on the temperature distribution and heat flux of the domain are studied.

  20. Design and analysis of heat recovery system in bioprocess plant

    International Nuclear Information System (INIS)

    Anastasovski, Aleksandar; Rašković, Predrag; Guzović, Zvonimir

    2015-01-01

    Highlights: • Heat integration of a bioprocess plant is studied. • Bioprocess plant produces yeast and ethyl-alcohol. • The design of a heat recovery system is performed by batch pinch analysis. • Direct and indirect heat integration approaches are used in process design. • The heat recovery system without a heat storage opportunity is more profitable. - Abstract: The paper deals with the heat integration of a bioprocess plant which produces yeast and ethyl-alcohol. The referent plant is considered to be a multiproduct batch plant which operates in a semi-continuous mode. The design of a heat recovery system is performed by batch pinch analysis and by the use of the Time slice model. The results obtained by direct and indirect heat integration approaches are presented in the form of cost-optimal heat exchanger networks and evaluated by different thermodynamic and economic indicators. They signify that the heat recovery system without a heat storage opportunity can be considered to be a more profitable solution for the energy efficiency increase in a plant

  1. Electromagnetic heating processes: analysis and simulations

    OpenAIRE

    Calay, Rajnish Kaur

    1994-01-01

    Electromagnetic heating (EMH) processes are being increasingly used in the industrial and domestic sectors, yet they receive relatively little attention in the thermal engineering domain. Time-temperature characteristics in EMH are qualitatively different from those in conventional heating techniques due to the additional parameters (viz dielectric properties of the material, size and shape of the product and process frequency). From a unified theory perspective, a multi-...

  2. Stability analysis for downflow in heated channels

    International Nuclear Information System (INIS)

    Sampaio, P.A.B. de.

    1985-01-01

    Stability and flow distribution are analysed for downflow in heated channels. It is shown that at low flow rates instabilities associated with the buoyancy forces may appear. A computer code in FORTRAN language to determine downflow distribution among n heated channels is presented. The model used to calculate downflow distribution and the onset of instability is compared with experiments performed in a test section with two parallel channels. (Author) [pt

  3. Thermodynamic analysis and performance assessment of an integrated heat pump system for district heating applications

    International Nuclear Information System (INIS)

    Soltani, Reza; Dincer, Ibrahim; Rosen, Marc A.

    2015-01-01

    A Rankine cycle-driven heat pump system is modeled for district heating applications with superheated steam and hot water as products. Energy and exergy analyses are performed, followed by parametric studies to determine the effects of varying operating conditions and environmental parameters on the system performance. The district heating section is observed to be the most inefficient part of system, exhibiting a relative irreversibility of almost 65%, followed by the steam evaporator and the condenser, with relative irreversibilities of about 18% and 9%, respectively. The ambient temperature is observed to have a significant influence on the overall system exergy destruction. As the ambient temperature decreases, the system exergy efficiency increases. The electricity generated can increase the system exergy efficiency at the expense of a high refrigerant mass flow rate, mainly due to the fact that the available heat source is low quality waste heat. For instance, by adding 2 MW of excess electricity on top of the targeted 6 MW of product heat, the refrigerant mass flow rate increases from 12 kg/s (only heat) to 78 kg/s (heat and electricity), while the production of 8 MW of product heat (same total output, but in form of heat) requires a refrigerant mass flow rate of only 16 kg/s. - Highlights: • A new integrated heat pump system is developed for district heating applications. • An analysis and assessment study is undertaken through exergy analysis methodology. • A comparative efficiency evaluation is performed for practical applications. • A parametric study is conducted to investigate how varying operating conditions and state properties affect energy and exergy efficiencies.

  4. Empirical Analysis for the Heat Exchange Effectiveness of a Thermoelectric Liquid Cooling and Heating Unit

    Directory of Open Access Journals (Sweden)

    Hansol Lim

    2018-03-01

    Full Text Available This study aims to estimate the performance of thermoelectric module (TEM heat pump for simultaneous liquid cooling and heating and propose empirical models for predicting the heat exchange effectiveness. The experiments were conducted to investigate and collect the performance data of TEM heat pump where the working fluid was water. A total of 57 sets of experimental data were statistically analyzed to estimate the effects of each independent variable on the heat exchange effectiveness using analysis of variance (ANOVA. To develop the empirical model, the six design parameters were measured: the number of transfer units (NTU of the heat exchangers (i.e., water blocks, the inlet water temperatures and temperatures of water blocks at the cold and hot sides of the TEM. As a result, two polynomial equations predicting heat exchange effectiveness at the cold and hot sides of the TEM heat pump were derived as a function of the six selected design parameters. Also, the proposed models and theoretical model of conventional condenser and evaporator for heat exchange effectiveness were compared with the additional measurement data to validate the reliability of the proposed models. Consequently, two conclusions have been made: (1 the possibility of using the TEM heat pump for simultaneous cooling and heating was examined with the maximum temperature difference of 30 °C between cold and hot side of TEM, and (2 it is revealed that TEM heat pump has difference with the conventional evaporator and condenser from the comparison results between the proposed models and theoretical model due to the heat conduction and Joule effect in TEM.

  5. 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.

  6. Performance Analysis of a Hybrid District Heating System

    DEFF Research Database (Denmark)

    Mikulandric, Robert; Krajačić, Goran; Duic, Neven

    2015-01-01

    Hybridisation of district heating systems can contribute to more efficient heat generation through cogeneration power plants or through the share increase of renewable energy sources in total energy consumption while reducing negative aspects of particular energy source utilisation. In this work......, the performance of a hybrid district energy system for a small town in Croatia has been analysed. Mathematical model for process analysis and optimisation algorithm for optimal system configuration has been developed and described. The main goal of the system optimisation is to reduce heat production costs....... Several energy sources for heat production have been considered in 8 different simulation cases. Simulation results show that the heat production costs could be reduced with introduction of different energy systems into an existing district heating system. Renewable energy based district heating systems...

  7. 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.

  8. Heat transfer analysis of frictional heat dissipation during articulation of femoral implants.

    Science.gov (United States)

    Davidson, J A; Gir, S; Paul, J P

    1988-12-01

    Previous studies have shown the tendency for frictional heating to occur during articulation of total hip systems in vitro under simulated hip loading conditions. The magnitude of this heating is sufficient to accelerate wear, creep, and oxidation degradation of the UHMWPE bearing surface. It was shown that ceramic articulating systems generate less frictional heating than polished cobalt alloy against UHMWPE. This frictional heating is expected to occur primarily for younger, heavier, and more active patients. Thus, long-term performance of the articulating hip system in these patients may not be that predicted from current, body-temperature wear, creep, and degradation studies. Although the tendency to generate frictional heat has been observed only during in vitro simulated hip loading, a heat transfer analysis of this phenomenon is presented to evaluate the ability of the hip joint to dissipate such heating in vivo. Additional experiments were performed using controlled resistance heaters inside a cobalt femoral head to verify the calculated levels of frictional heat and to assess the heat dissipation under simulated in vivo conditions. The effect of blood perfusion on the effective thermal conductivity of the joint capsule is also discussed. The present study describes and analyzes the various heat dissipation mechanisms present both in vitro and in vivo during articulation of metal and ceramic hip systems. From these tests and analyses, it is concluded that frictional heating in the reconstructed hip cannot be effectively removed, and that degredative elevated temperature processes can be expected to occur in vivo to both the UHMWPE and adjacent tissue under extended periods of excessive patient activity. This is particularly true for metal cobalt alloy femoral heads articulating on UHMWPE versus ceramic heads which generate significantly lower levels of heat.

  9. A detailed description of the analysis of the decay of neutral kaons to $\\pi^+ \\pi^-$ in the CPLEAR experiment

    CERN Document Server

    Apostolakis, Alcibiades J; Backenstoss, Gerhard; Bargassa, P; Behnke, O; Benelli, A; Bertin, V; Blanc, F; Bloch, P; Carlson, P J; Carroll, M; Cawley, E; Chertok, M B; Danielsson, M; Dejardin, M; Derré, J; Ealet, A; Eleftheriadis, C; Fetscher, W; Fidecaro, Maria; Filipcic, A; Francis, D; Fry, J; Gabathuler, Erwin; Gamet, R; Gerber, H J; Go, A; Haselden, A; Hayman, P J; Henry-Coüannier, F; Hollander, R W; Jon-And, K; Kettle, P R; Kokkas, P; Kreuger, R; Le Gac, R; Leimgruber, F; Mandic, I; Manthos, N; Marel, Gérard; Mikuz, M; Miller, J; Montanet, François; Müller, A; Nakada, Tatsuya; Pagels, B; Papadopoulos, I M; Pavlopoulos, P; Polivka, G; Rickenbach, R; Roberts, B L; Ruf, T; Sakelliou, L; Schäfer, M; Schaller, L A; Schietinger, T; Schopper, A; Tauscher, Ludwig; Thibault, C; Touchard, F; Touramanis, C; van Eijk, C W E; Vlachos, S; Weber, P; Wigger, O; Wolter, M; Yéche, C; Zavrtanik, D; Zimmerman, D

    2000-01-01

    A detailed description is given of the analysis of neutral kaons decaying to \\pipi , based on the complete data sample collected with the CPLEAR experiment.Using a novel approach involving initially strangeness-tagged \\kn\\ and \\knb ,the time-dependent decay rate asymmetry has been measured. This asymmetry, resulting from the interference between the \\ks\\and \\kl\\ decay amplitudes, has enabled both the magnitudeand phase of the CP-violation parameter, \\ita , to be measured, with aprecision comparable to that of the current world average values.

  10. A detailed description of the analysis of the decay of neutral kaons to π+π- in the CPLEAR experiment

    International Nuclear Information System (INIS)

    Apostolakis, A.; Aslanides, E.

    2000-01-01

    A detailed description is given of the analysis of neutral kaons decaying to π + π - , based on the complete set of data collected with the CPLEAR experiment. Using a novel approach involving initially strangeness-tagged K 0 and anti K 0 , the time-dependent decay-rate asymmetry has been measured. This asymmetry, resulting from the interference between the K S and K L decay amplitudes, has enabled both the magnitude and phase of the CP-violation parameter, η +- , to be measured, with a precision comparable to that of the current world-average values. (orig.)

  11. Heat transfer analysis of parabolic trough solar receiver

    International Nuclear Information System (INIS)

    Padilla, Ricardo Vasquez; Demirkaya, Gokmen; Goswami, D. Yogi; Stefanakos, Elias; Rahman, Muhammad M.

    2011-01-01

    Highlights: → In this paper a detailed one dimensional numerical heat transfer analysis of a PTC is performed. → The receiver and envelope were divided into several segments and mass and energy balance were applied in each segment. → Improvements either in the heat transfer correlations or radiative heat transfer analysis are presented. → The proposed heat transfer model was validated with experimental data obtained from Sandia National Laboratory. → Our results showed a better agreement with experimental data compared to other models. -- Abstract: Solar Parabolic Trough Collectors (PTCs) are currently used for the production of electricity and applications with relatively higher temperatures. A heat transfer fluid circulates through a metal tube (receiver) with an external selective surface that absorbs solar radiation reflected from the mirror surfaces of the PTC. In order to reduce the heat losses, the receiver is covered by an envelope and the enclosure is usually kept under vacuum pressure. The heat transfer and optical analysis of the PTC is essential to optimize and understand its performance under different operating conditions. In this paper a detailed one dimensional numerical heat transfer analysis of a PTC is performed. The receiver and envelope were divided into several segments and mass and energy balance were applied in each segment. Improvements either in the heat transfer correlations or radiative heat transfer analysis are presented as well. The partial differential equations were discretized and the nonlinear algebraic equations were solved simultaneously. Finally, to validate the numerical results, the model was compared with experimental data obtained from Sandia National Laboratory (SNL) and other one dimensional heat transfer models. Our results showed a better agreement with experimental data compared to other models.

  12. Exergy Analysis of a Ground-Coupled Heat Pump Heating System with Different Terminals

    Directory of Open Access Journals (Sweden)

    Xiao Chen

    2015-04-01

    Full Text Available In order to evaluate and improve the performance of a ground-coupled heat pump (GCHP heating system with radiant floors as terminals, an exergy analysis based on test results is performed in this study. The system is divided into four subsystems, and the exergy loss and exergy efficiency of each subsystem are calculated using the expressions derived based on exergy balance equations. The average values of the measured parameters are used for the exergy analysis. The analysis results show that the two largest exergy losses occur in the heat pump and terminals, with losses of 55.3% and 22.06%, respectively, and the lowest exergy efficiency occurs in the ground heat exchange system. Therefore, GCHP system designers should pay close attention to the selection of heat pumps and terminals, especially in the design of ground heat exchange systems. Compared with the scenario system in which fan coil units (FCUs are substituted for the radiant floors, the adoption of radiant floors can result in a decrease of 12% in heating load, an increase of 3.24% in exergy efficiency of terminals and an increase of 1.18% in total exergy efficiency of the system. The results may point out the direction and ways of optimizing GCHP systems.

  13. Determination and identification of naturally occurring decay series using milli-second order pulse time interval analysis (TIA)

    International Nuclear Information System (INIS)

    Hashimoto, T.; Sanada, Y.; Uezu, Y.

    2003-01-01

    A delayed coincidence method, called a time interval analysis (TIA) method, has been successfully applied to selective determination of the correlated α-α decay events in millisecond order life-time. A main decay process applicable to TIA-treatment is 220 Rn → 216 Po(T 1/2 :145ms) → {Th-series}. The TIA is fundamentally based on the difference of time interval distribution between non-correlated decay events and other events such as background or random events when they were compiled the time interval data within a fixed time (for example, a tenth of concerned half lives). The sensitivity of the TIA-analysis due to correlated α-α decay events could be subsequently improved in respect of background elimination using the pulse shape discrimination technique (PSD with PERALS counter) to reject β/γ-pulses, purging of nitrogen gas into extra scintillator, and applying solvent extraction of Ra. (author)

  14. Heat transfer and thermal stress analysis in grooved tubes

    Indian Academy of Sciences (India)

    Heat transfer and thermal stresses, induced by temperature differencesin the internally grooved tubes of heat transfer equipment, have been analysed numerically. The analysis has been conducted for four different kinds of internally grooved tubes and three different mean inlet water velocities. Constant temperature was ...

  15. GIS Based Analysis of future district heating potential in Denmark

    DEFF Research Database (Denmark)

    Nielsen, Steffen; Möller, Bernd

    2012-01-01

    and their heat demand has in recent years been assessed in a heat atlas including all buildings in Denmark. The focus in this article is to further develop a method for assessing the costs associated with supplying these buildings with DH. The analysis takes departure in the existing DH areas in Denmark...

  16. Model-based analysis and simulation of regenerative heat wheel

    DEFF Research Database (Denmark)

    Wu, Zhuang; Melnik, Roderick V. N.; Borup, F.

    2006-01-01

    The rotary regenerator (also called the heat wheel) is an important component of energy intensive sectors, which is used in many heat recovery systems. In this paper, a model-based analysis of a rotary regenerator is carried out with a major emphasis given to the development and implementation of...

  17. UHS, Ultimate Heat Sink Cooling Pond Analysis

    International Nuclear Information System (INIS)

    Codell, R.; Nuttle, W.K.

    1998-01-01

    1 - Description of program or function: Three programs model performance of an ultimate heat sink cooling pond. National Weather Service data is read and analyzed to predict periods of lowest cooling performance and highest evaporative loss. The data is compared to local site data for significant differences. Then the maximum pond temperature is predicted. Five programs model performance of an ultimate heat sink spray pond. The cooling performance, evaporative water loss, and drift water loss as a function of wind speed are estimated for a spray field. These estimates are used in conjunction with National Weather Service data to predict periods of lowest cooling performance and highest evaporative loss. This data is compared to local site data for significant differences. Then the maximum pond temperature is predicted. 2 - Method of solution: The transfer of heat and water vapor is modeled using an equilibrium temperature procedure for an UHS cooling pond. The UHS spray pond model considers heat, mass, and momentum transfer from a single water drop with the surrounding air, and modification of the surrounding air resulting from the heat, mass, and momentum transfer from many drops in different parts of a spray field. 3 - Restrictions on the complexity of the problem: The program SPRCO uses RANF, a uniform random number generator which is an intrinsic function on the CDC. All programs except COMET use the NAMELIST statement, which is non standard. Otherwise these programs conform to the ANSI Fortran 77 standard. The meteorological data scanning procedure requires tens of years of recorded data to be effective. The models and methods, provided as useful tool for UHS analyses of cooling ponds and spray ponds, are intended as guidelines only. Use of these methods does not automatically assure NRC approval, nor are they required procedures for nuclear-power-plant licensing

  18. Analysis of the flow structure and heat transfer in a vertical mantle heat exchanger

    DEFF Research Database (Denmark)

    Knudsen, Søren; Morrison, GL; Behnia, M

    2005-01-01

    initially mixed and initially stratified inner tank and mantle. The analysis of the heat transfer showed that the flow in the mantle near the inlet is mixed convection flow and that the heat transfer is dependent on the mantle inlet temperature relative to the core tank temperature at the mantle level. (C......The flow structure inside the inner tank and inside the mantle of a vertical mantle heat exchanger was investigated using a full-scale tank designed to facilitate flow visualisation. The flow structure and velocities in the inner tank and in the mantle were measured using a Particle Image...... Velocimetry (PIV) system. A Computational Fluid Dynamics (CFD) model of the vertical mantle heat exchanger was also developed for a detailed evaluation of the heat flux at the mantle wall and at the tank wall. The flow structure was evaluated for both high and low temperature incoming flows and for both...

  19. Optimization of irradiation decay and counting times in nuclear activation analysis using short-lived nuclides

    International Nuclear Information System (INIS)

    Bjoernstad, T.

    This work describes a method and outlines a procedure for optim- ization of an activation analysis with respect to the experimental times, irradiation time, t(subi), decay time and counting time. The method is based on the 'minimum relative standard deviation criterion', and specially designed for the use on short-lived nuclides. A computer program, COMB1, is written in the BASIC language in order to make the calculations easier and faster. It is intended to be understandable, and easily applicable on a computer of modest size. Time and cost are important factors, especially for routine analysis on a service basis. In such cases one can often allow a controlled reduction in the analysis quality (through a higher relative standard deviation). The procedure outlined can therefore help find acceptable conditions by calculation of the 'best practical' (or reasonable) experimental time values, and the minimum number of accumulation cycles necessary to fulfil the requirements given. (Auth.)

  20. Analysis of $B^{0}_{d} \\to K^{*0}\\mu^{+}\\mu^{-}$ Decay with the ATLAS Experiment

    CERN Document Server

    Usanova, Anna

    ATLAS is a general-purpose experiment at the Large Hadron Collider. Beside other goals, it also aims at the study of B -hadrons. B -physics offers a large number of channels that can provide information about some fundamental properties of our universe. Among them, the B 0 d ! K 0 + - decay is sensitive to the potential presence of particles that are not predicted by the Standard Model. Such ”new physics” effects can be observed indirectly by studying angular distributions of the B 0 d ! K 0 + - decay products. This thesis describes the analysis of 4.9 fb - 1 of data produced in the proton-proton collisions at the centre-of-mass energy p s = 7 TeV at the LHC in the year 2011 and recorded by the ATLAS detector. The main steps of analysis are described, such as the selection of the signal events, the data fit procedure and the estimation of uncertainties. The obtained results are compared with other experiments and with the Standard Model prediction.

  1. Dalitz Plot Analysis of the Decay B+ -> K+K+K-

    Energy Technology Data Exchange (ETDEWEB)

    Dvoretskii, Alexei; /SLAC /Caltech

    2006-05-05

    The authors perform an analysis of the three-body charmless decay B{sup {+-}} {yields} K{sup {+-}}K{sup {+-}}K{sup {-+}} using a sample of 226.0 {+-} 2.5 million B{bar B} pairs collected by the BABAR detector and measure the total branching fraction and Cp asymmetry to be {beta} = (35.2 {+-} 0.9 {+-} 1.6) x 10{sup -6} and A{sub CP} = (-1.7 {+-} 2.6 {+-} 1.5)%. They fit the Dalitz plot distribution using an isobar model and report the measured values of magnitudes and phases of the production coefficients. The decay dynamics is dominated by the K{sup +}K{sup -} S-wave, for which we perform a partial-wave analysis in the region m(K{sup +}K{sup -}) < 2 GeV/c{sup 2}. They find no evidence of CP violation for individual components of the isobar model.

  2. Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2005-08-29

    This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

  3. RIBBED DOUBLE PIPE HEAT EXCHANGER: ANALYTICAL ANALYSIS

    Directory of Open Access Journals (Sweden)

    HUSSAIN H. AL-KAYIEM

    2011-02-01

    Full Text Available This paper presents the findings obtained by modeling a Double Pipe Heat Exchanger (DPHE equipped with repeated ribs from the inside for artificial roughing. An analytical procedure was developed to analyze the thermal and hydraulic performance of the DPHE with and without ribbing. The procedure was verified by comparing with experimental reported results and they are in good agreement. Several parameters were investigated in this study including the effect of ribs pitch to height ratios, P/e= 5, 10, 15, and 20, and ribs to hydraulic diameter ratios, e/Dh= 0.0595, 0.0765, and 0.107. These parameters were studied at various operating Reynolds number ranging from 2500 to 150000. Different installation configurations were investigated, too. An enhan-cement of 4 times in the heat transfer in terms of Stanton number was achieved at the expense of 38 times increase of pressure drop across the flow in terms of friction facto values.

  4. Comparisons of experimental beta-ray spectra important to decay heat predictions with ENSDF [Evaluated Nuclear Structure Data File] evaluations

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1990-03-01

    Graphical comparisons of recently obtained experimental beta-ray spectra with predicted beta-ray spectra based on the Evaluated Nuclear Structure Data File are exhibited for 77 fission products having masses 79--99 and 130--146 and lifetimes between 0.17 and 23650 sec. The comparisons range from very poor to excellent. For beta decay of 47 nuclides, estimates are made of ground-state transition intensities. For 14 cases the value in ENSDF gives results in very good agreement with the experimental data. 12 refs., 77 figs., 1 tab

  5. Heat Transfer Analysis of Localized Heat-Treatment for Grade 91 Steel

    Science.gov (United States)

    Walker, Jacob D.

    through an inverse method by collecting actual data from different conditions and temperatures. Then the heat transfer coefficients are used to set up a model to determine the appropriate post-weld heat treatment conditions for Grade 91 steel. This will enable one to use the derived coefficients to run a forward analysis with the specific geometry and conditions they will encounter in the heat treatment process for their application. The analysis will provide a theoretical determination of time and temperatures needed to maintain the temperature for the proper time needed to properly heat treat the welded section in the desired areas that have been joined together through a welding process. Finally time and temperature combinations are compared with experimentally measured data. The forward model code applied to the parameters of the heat-treatment can then appropriately assist to determine the proper post-weld heat treatment conditions for the desired toughness and creep properties. This research is very beneficial to the joining of metals industry because it provides a way to ensure the method used to heat treat the welded section is being properly done, and the required heat treatment is achieved. It is applicable to many different geometries so that it can be modified to specific situations.

  6. Vector-Tensor and Vector-Vector Decay Amplitude Analysis of B0→φK*0

    International Nuclear Information System (INIS)

    Aubert, B.; Bona, M.; Boutigny, D.; Couderc, F.; Karyotakis, Y.; Lees, J. P.; Poireau, V.; Tisserand, V.; Zghiche, A.; Grauges, E.; Palano, A.; Chen, J. C.; Qi, N. D.; Rong, G.; Wang, P.; Zhu, Y. S.; Eigen, G.; Ofte, I.; Stugu, B.; Abrams, G. S.

    2007-01-01

    We perform an amplitude analysis of the decays B 0 →φK 2 * (1430) 0 , φK * (892) 0 , and φ(Kπ) S-wave 0 with a sample of about 384x10 6 BB pairs recorded with the BABAR detector. The fractions of longitudinal polarization f L of the vector-tensor and vector-vector decay modes are measured to be 0.853 -0.069 +0.061 ±0.036 and 0.506±0.040±0.015, respectively. Overall, twelve parameters are measured for the vector-vector decay and seven parameters for the vector-tensor decay, including the branching fractions and parameters sensitive to CP violation

  7. Survival analysis approach to account for non-exponential decay rate effects in lifetime experiments

    Energy Technology Data Exchange (ETDEWEB)

    Coakley, K.J., E-mail: kevincoakley@nist.gov [National Institute of Standards and Technology, 325 Broadway, Boulder, CO 80305 (United States); Dewey, M.S.; Huber, M.G. [National Institute of Standards and Technology, 100 Bureau Drive, Stop 8461, Gaithersburg, MD 20899 (United States); Huffer, C.R.; Huffman, P.R. [North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Marley, D.E. [National Institute of Standards and Technology, 100 Bureau Drive, Stop 8461, Gaithersburg, MD 20899 (United States); North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Mumm, H.P. [National Institute of Standards and Technology, 100 Bureau Drive, Stop 8461, Gaithersburg, MD 20899 (United States); O' Shaughnessy, C.M. [University of North Carolina at Chapel Hill, 120 E. Cameron Ave., CB #3255, Chapel Hill, NC 27599 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Schelhammer, K.W. [North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Thompson, A.K.; Yue, A.T. [National Institute of Standards and Technology, 100 Bureau Drive, Stop 8461, Gaithersburg, MD 20899 (United States)

    2016-03-21

    In experiments that measure the lifetime of trapped particles, in addition to loss mechanisms with exponential survival probability functions, particles can be lost by mechanisms with non-exponential survival probability functions. Failure to account for such loss mechanisms produces systematic measurement error and associated systematic uncertainties in these measurements. In this work, we develop a general competing risks survival analysis method to account for the joint effect of loss mechanisms with either exponential or non-exponential survival probability functions, and a method to quantify the size of systematic effects and associated uncertainties for lifetime estimates. As a case study, we apply our survival analysis formalism and method to the Ultra Cold Neutron lifetime experiment at NIST. In this experiment, neutrons can escape a magnetic trap before they decay due to a wall loss mechanism with an associated non-exponential survival probability function.

  8. Survival analysis approach to account for non-exponential decay rate effects in lifetime experiments

    International Nuclear Information System (INIS)

    Coakley, K.J.; Dewey, M.S.; Huber, M.G.; Huffer, C.R.; Huffman, P.R.; Marley, D.E.; Mumm, H.P.; O'Shaughnessy, C.M.; Schelhammer, K.W.; Thompson, A.K.; Yue, A.T.

    2016-01-01

    In experiments that measure the lifetime of trapped particles, in addition to loss mechanisms with exponential survival probability functions, particles can be lost by mechanisms with non-exponential survival probability functions. Failure to account for such loss mechanisms produces systematic measurement error and associated systematic uncertainties in these measurements. In this work, we develop a general competing risks survival analysis method to account for the joint effect of loss mechanisms with either exponential or non-exponential survival probability functions, and a method to quantify the size of systematic effects and associated uncertainties for lifetime estimates. As a case study, we apply our survival analysis formalism and method to the Ultra Cold Neutron lifetime experiment at NIST. In this experiment, neutrons can escape a magnetic trap before they decay due to a wall loss mechanism with an associated non-exponential survival probability function.

  9. Foundation heat transfer analysis for buildings with thermal piles

    International Nuclear Information System (INIS)

    Almanza Huerta, Luis Enrique; Krarti, Moncef

    2015-01-01

    Highlights: • A numerical transient thermal model for thermo-active foundations is developed. • Thermal interactions between thermal piles and building foundations are evaluated. • A simplified analysis method of thermal interactions between thermal piles and building foundations is developed. - Abstract: Thermal piles or thermo-active foundations utilize heat exchangers embedded within foundation footings to heat and/or cool buildings. In this paper, the impact of thermal piles on building foundation heat transfer is investigated. In particular, a simplified analysis method is developed to estimate the annual ground-coupled foundation heat transfer when buildings are equipped with thermal piles. First, a numerical analysis of the thermal performance of thermo-active building foundations is developed and used to assess the interactions between thermal piles and slab-on-grade building foundations. The impact of various design parameters and operating conditions is evaluated including foundation pile depth, building slab width, foundation insulation configuration, and soil thermal properties. Based on the results of a series of parametric analyses, a simplified analysis method is presented to assess the impact of the thermal piles on the annual heat fluxes toward or from the building foundations. A comparative evaluation of the predictions of the simplified analysis method and those obtained from the detailed numerical analysis indicated good agreement with prediction accuracy lower than 5%. Moreover, it is found that thermal piles can affect annual building foundation heat loss/gain by up to 30% depending on foundation size and insulation level

  10. An analytic analysis of the pion decay constant in three-flavoured chiral perturbation theory

    Energy Technology Data Exchange (ETDEWEB)

    Ananthanarayan, B.; Ghosh, Shayan [Indian Institute of Science, Centre for High Energy Physics, Bangalore, Karnataka (India); Bijnens, Johan [Lund University, Department of Astronomy and Theoretical Physics, Lund (Sweden)

    2017-07-15

    A representation of the two-loop contribution to the pion decay constant in SU(3) chiral perturbation theory is presented. The result is analytic up to the contribution of the three (different) mass sunset integrals, for which an expansion in their external momentum has been taken. We also give an analytic expression for the two-loop contribution to the pion mass based on a renormalized representation and in terms of the physical eta mass. We find an expansion of F{sub π} and M{sub π}{sup 2} in the strange-quark mass in the isospin limit, and we perform the matching of the chiral SU(2) and SU(3) low-energy constants. A numerical analysis demonstrates the high accuracy of our representation, and the strong dependence of the pion decay constant upon the values of the low-energy constants, especially in the chiral limit. Finally, we present a simplified representation that is particularly suitable for fitting with available lattice data. (orig.)

  11. An angular analysis of $\\Lambda_b \\to \\Lambda \\mu^+ \\mu^-$ decays at the LHCb experiment.

    CERN Document Server

    Everett, Christopher; Blake, Thomas

    This work presents an angular analysis of the suppressed $\\Lambda_b^0 \\rightarrow \\Lambda\\mu^+\\mu^-$ decay, performed using proton-proton collision data collected by the LHCb experiment during 2011 and 2012. The distribution of moments describing the decay with a non-zero transverse $\\Lambda_b^0$ baryon polarisation production are extracted for the first time, and used to determine the 5-dimensional angular distribution in the 15 to 20 GeV$^2/c^4$ dimuon invariant mass squared interval for the first time. The forward-backward asymmetry of the $\\Lambda$ baryon and $\\mu^+\\mu^-$ produced from the $\\Lambda_b^0$ baryon ($A_{FB}^{lh}$ ) is measured, along with the previously published $f_L$, $A^l_{FB}$ and $A^h_{FB}$. These values are found to be $f_L = 0.546 \\pm 0.132$, $A^l_{FB} = -0.004 \\pm 0.077$, $A^h_{FB} = -0.312 \\pm 0.150$ and $A_{FB}^{lh} = -0.007 ± 0.071$, where uncertainties are statistical, and account for correlation between the moments used to determine them.

  12. Amplitude Analysis of the Decay Ds+ → π+π-π+ in the Experiment E831/FOCUS

    International Nuclear Information System (INIS)

    Schilithz, Anderson Correa

    2005-01-01

    We present in this thesis the Dalitz Plot analysis of the D s + → π + π - π + decay, with the data of the E831/FOCUS, that took data in 1996 and 1997. The masses and widhts of f 0 (980) and f 0 (1370) are free parametres of the fit on Dalitz Plot, objectiving to study in detail these resonances. After this analysis we present the Spectator Model study on the S wave in this decay. For this study we used the formalism developed by M. Svec [2] for scattering. We present the comparison between the Isobar Model, frequently used in Dalitz Plot analysis, and this formalism.

  13. Analysis of heat transfer under high heat flux nucleate boiling conditions

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y.; Dinh, N. [3145 Burlington Laboratories, Raleigh, NC (United States)

    2016-07-15

    Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

  14. Analysis of heat transfer under high heat flux nucleate boiling conditions

    International Nuclear Information System (INIS)

    Liu, Y.; Dinh, N.

    2016-01-01

    Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

  15. Application of dynamic response analysis to JET heat pulse data

    International Nuclear Information System (INIS)

    Griguoli, A.; Sips, A.C.C.

    1993-09-01

    The plasma dynamic response can be used to study transport processes in a tokamak plasma. A method has been developed for the application of dynamic response analysis to study perturbations away from the plasma equilibrium. In this report perturbations on the electron temperature following a sawtooth collapse in the center of the plasma are considered. The method has been used to find mathematical description of a series of heat pulses at the Joint European Torus project (JET). From the plasma dynamic response, the time constants which characterise the heat pulse are obtained. These time constants are compared to the transport coefficients found in previous analysis of the JET heat pulse data. Various methods are discussed for applying dynamic response analysis to JET heat pulse data. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  17. Physical explosion analysis in heat exchanger network design

    Science.gov (United States)

    Pasha, M.; Zaini, D.; Shariff, A. M.

    2016-06-01

    The failure of shell and tube heat exchangers is being extensively experienced by the chemical process industries. This failure can create a loss of production for long time duration. Moreover, loss of containment through heat exchanger could potentially lead to a credible event such as fire, explosion and toxic release. There is a need to analyse the possible worst case effect originated from the loss of containment of the heat exchanger at the early design stage. Physical explosion analysis during the heat exchanger network design is presented in this work. Baker and Prugh explosion models are deployed for assessing the explosion effect. Microsoft Excel integrated with process design simulator through object linking and embedded (OLE) automation for this analysis. Aspen HYSYS V (8.0) used as a simulation platform in this work. A typical heat exchanger network of steam reforming and shift conversion process was presented as a case study. It is investigated from this analysis that overpressure generated from the physical explosion of each heat exchanger can be estimated in a more precise manner by using Prugh model. The present work could potentially assist the design engineer to identify the critical heat exchanger in the network at the preliminary design stage.

  18. Analysis of the rare decay B0 → K*0e+e- at LHCb

    International Nuclear Information System (INIS)

    Nicol, M.

    2012-01-01

    The high bb-bar cross section produced by the LHC offers an excellent opportunity for the study of flavour changing neutral current B decays, where the effect of new physics can be probed. This thesis presents an analysis of the rare decay B 0 → K *0 e + e - , which can be used to measure the polarisation of the photon in the b → sγ transition. When the dilepton mass is low, the e + e - pair comes predominantly from a virtual photon, and the polarisation can be accessed via an angular analysis. It is predicted to be predominantly left handed in the Standard Model, and therefore an enhanced right handed amplitude would be a sign of new physics. A first step is to measure the branching fraction in the dilepton mass range, 30 MeV/c 2 to 1 GeV. This decay has not yet been observed in this region, due to its small branching ratio. The analysis involves electrons with low transverse momentum, and is thus experimentally complex in the hadronic environment at the LHC. The branching ratio is measured relative to that of B 0 → J/ψ(e + e - )K *0 , which eliminates both certain experimental effects, and the need to determine absolute efficiencies. The result is obtained with an integrated luminosity of 1 fb -1 of pp collisions, collected by LHCb during 2011 and is found to be (within the [30-1000 MeV/C 2 ] mass range): Br(B 0 → K *0 e + e - ) = [3.19+0.75-0.68(stat) ± 0.21(syst) ± 0.15(PDG)]*10 -7 when using the PDG value for the B 0 → J/ψ(e + e - )K *0 branching ratio. The last part of the thesis presents Monte Carlo studies, showing that with the inclusion of the 2012 data sample, the expected sensitivity on the fraction of right handed polarisation is approximately 0.1, which is comparable with the world average obtained with different methods. (author)

  19. Three-dimensional calculation analysis of ICRF heating in LHD

    International Nuclear Information System (INIS)

    Seki, Tetsuo; Kumazawa, Ryuhei; Mutoh, Takashi

    2004-01-01

    Ion cyclotron range of frequencies (ICRF) heating is one of the heating methods for the fusion plasma experiments and also effective for the helical plasmas. For the purpose of analysis of the ICRF heating in the helical plasmas, the three-dimensional full-wave code has been developed. The feature of the helical system compared with the tokamak device is the strong coupling of the toroidal harmonic modes. They cannot be treated independently. Dependence of the power absorption on the position of the ion cyclotron resonance layer is calculated including all toroidal modes. Strong power absorption was obtained when the position of the resonance layer is slightly different from the experimental results. Difference of the position of the resonance layer in different toroidal angle is thought to be important to achieve the good heating efficiency in the ICRF heating for the helical plasmas. (author)

  20. Heat transfer analysis of liquid piston compressor for hydrogen applications

    DEFF Research Database (Denmark)

    Kermani, Nasrin Arjomand; Rokni, Masoud

    2015-01-01

    A hydrogen compression technology using liquid as the compression piston is investigated from heat transfer point of view. A thermodynamic model, simulating a single compression stroke, is developed to investigate the heat transfer phenomena inside the compression chamber. The model is developed...... and through the walls, is investigated and compared with the adiabatic case. The results show that depending on heat transfer correlation, the hydrogen temperature reduces slightly between 0.2% and 0.4% compared to the adiabatic case, at 500bar, due to the large wall resistance and small contact area...... at the interface. Moreover, the results of the sensitivity analysis illustrates that increasing the total heat transfer coefficients at the interface and the wall, together with compression time, play key roles in reducing the hydrogen temperature. Increasing the total heat transfer coefficient at the interface...

  1. Economic analysis of electric heating based on critical electricity price

    Science.gov (United States)

    Xie, Feng; Sun, Zhijie; Zhou, Xinnan; Fu, Chengran; Yang, Jie

    2018-06-01

    The State Grid Corporation of China proposes an alternative energy strategy, which will make electric heating an important task in the field of residential electricity consumption. This article takes this as the background, has made the detailed introduction to the inhabitant electric heating technology, and take the Zhangjiakou electric panels heating technology as an example, from the expense angle, has carried on the analysis to the electric panels heating economy. In the field of residential heating, electric panels operating costs less than gas boilers. After customers implying energy-saving behavior, electric panels operating cost is even lower than coal-fired boilers. The critical price is higher than the execution price, which indicates that the economic performance of the electric panels is significantly higher than that of the coal boiler.

  2. 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)

  3. 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

  4. Multivariate regression analysis for determining short-term values of radon and its decay products from filter measurements

    International Nuclear Information System (INIS)

    Kraut, W.; Schwarz, W.; Wilhelm, A.

    1994-01-01

    A multivariate regression analysis is applied to decay measurements of α-resp. β-filter activcity. Activity concentrations for Po-218, Pb-214 and Bi-214, resp. for the Rn-222 equilibrium equivalent concentration are obtained explicitly. The regression analysis takes into account properly the variances of the measured count rates and their influence on the resulting activity concentrations. (orig.) [de

  5. Constraints on the unitarity triangle angle gamma from Dalitz plot analysis of B-0 -> DK+pi(-) decays

    NARCIS (Netherlands)

    Aaij, R.; Beteta, C. Abellan; Adeva, B.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Akar, S.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Cartelle, P. Alvarez; Alves, A. A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Andreassi, G.; Andreotti, M.; Andrews, J. E.; Appleby, R. B.; Gutierrez, O. Aquines; Archilli, F.; d'Argent, P.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J. J.; Badalov, A.; Baesso, C.; Baldini, W.; Barlow, R. J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Battista, V.; Beaucourt, L.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Bel, L. J.; Onderwater, C. J. G.; Pellegrino, A.; Tolk, S.

    2016-01-01

    The first study is presented of CP violation with an amplitude analysis of the Dalitz plot of B-0 -> DK+pi(-) decays, with D -> K+pi(-), K+K-, and pi(+)pi(-). The analysis is based on a data sample corresponding to 3.0 fb(-1) of pp collisions collected with the LHCb detector. No significant CP

  6. Decay heat measurement of 235U in the time region from 10 to 1,000 seconds

    International Nuclear Information System (INIS)

    Baumung, K.

    1976-01-01

    In order to provide a good cooling of the fuel samples during irradiation in the thermal column of the FR2 reactor, vehicles for a pneumatic transfer system were developed and constructed which keep the fuel temperature at about 10 K above the temperature of the cooling gas (N 2 at 1 atm and about 300 K). For these 'rabbits' a new transport system has to be designed. Together with the mechanical system inside the vacuum vessel, which contains the calorimeter, the samples can be transfered from the irradiation position into the adiabatic jacket automatically within 2 to 3 seconds. For the measurement of the decay energy escaping the samples as γ-radiation, a MOXON-RAE-type detector was constructed. As its sensitivity increases linearly with the energy of the γ-quanta detected, this counter directly measures the energy flux of an unknown γ-spectrum. (orig./RW) [de

  7. Heat Retreat Locations in Cities - The Survey-Based Location Analysis of Heat Relief

    Science.gov (United States)

    Neht, Alice; Maximini, Claudia; Prenger-Berninghoff, Kathrin

    2017-12-01

    The adaptation of cities to climate change effects is one of the major strategies in urban planning to encounter the challenges of climate change (IPCC 2014). One of the fields of climate change adaption is dealing with heat events that occur more frequently and with greater intensity. Cities in particular are vulnerable to these events due to high population and infrastructure density. Proceeding urbanization calls for the existence of sufficient heat retreat locations (HRL) to enable relief for the population from heat in summer. This is why an extensive analysis of HRL is needed. This paper aims at the development of a survey-based location analysis of heat relief by identifying user groups, locations and characteristics of HRL based on a home survey that was conducted in three German cities. Key results of the study show that the majority of the participants of the survey are users of existing HRL, are affected by heat, and perceive heat as a burden in summer. Moreover, HRL that are located in close proximity are preferred by most users while their effect depends on the regional context that has to be considered in the analysis. Hence, this research presents an approach to heat relief that underlines the importance of HRL in cities by referring to selected examples of HRL types in densely populated areas of cities. HRL should especially be established and secured in densely built-up areas of cities. According to results of the survey, most HRL are located in public spaces, and the overall accessibility of HRL turned out to be an issue.

  8. Dalitz plot analysis of $B_s^0 \\rightarrow \\overline{D}^0 K^- \\pi^+$ decays

    CERN Document Server

    Aaij, Roel; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassen, Rolf; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Belogurov, Sergey; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Bjørnstad, Pål Marius; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Brambach, Tobias; van den Brand, Johannes; Bressieux, Joël; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Brown, Henry; Bursche, Albert; Busetto, Giovanni; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Ciba, Krzystof; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pascal; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Di Canto, Angelo; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gavrilov, Gennadii; Geraci, Angelo; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Gianì, Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, Vladimir; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Hunt, Philip; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jaton, Pierre; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kaballo, Michael; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kozlinskiy, Alexandr; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leo, Sabato; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lopez-March, Neus; Lowdon, Peter; Lu, Haiting; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martens, Aurelien; Martín Sánchez, Alexandra; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Moggi, Niccolò; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Nicol, Michelle; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Oggero, Serena; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Orlandea, Marius; Otalora Goicochea, Juan Martin; Owen, Patrick; Oyanguren, Maria Arantza; Pal, Bilas Kanti; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Parkinson, Christopher John; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pazos Alvarez, Antonio; Pearce, Alex; Pellegrino, Antonio; Pepe Altarelli, Monica; Perazzini, Stefano; Perez Trigo, Eliseo; Perret, Pascal; Perrin-Terrin, Mathieu; Pescatore, Luca; Pesen, Erhan; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Polci, Francesco; Poluektov, Anton; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Roa Romero, Diego; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruffini, Fabrizio; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrie, Mauro; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Seco, Marcos; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Sparkes, Ailsa; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Subbiah, Vijay Kartik; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szilard, Daniela; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Tran, Minh Tâm; Tresch, Marco; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; Voss, Helge; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Whitehead, Mark; Wicht, Jean; Wiedner, Dirk; Wilkinson, Guy; Williams, Matthew; Williams, Mike; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wu, Suzhi; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Wen Chao; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zvyagin, Alexander

    2014-01-01

    The resonant substructure of $B_s^0 \\rightarrow \\overline{D}^0 K^- \\pi^+$ decays is studied with the Dalitz plot analysis technique. The study is based on a data sample corresponding to an integrated luminosity of $3.0\\,{\\rm fb}^{-1}$ of $pp$ collision data recorded by LHCb. A structure at $m(\\overline{D}^0 K^-) \\approx 2.86 {\\rm GeV}/c^2$ is found to be an admixture of spin-1 and spin-3 resonances. The masses and widths of these states and of the $D^*_{s2}(2573)^-$ meson are measured, as are the complex amplitudes and fit fractions for all the $\\overline{D}^0 K^-$ and $K^-\\pi^+$ components included in the amplitude model. In addition, the $D^*_{s2}(2573)^-$ resonance is confirmed to be spin-2.

  9. Analysis of the Intermediate-State Contributions to Neutrinoless Double β− Decays

    Directory of Open Access Journals (Sweden)

    Juhani Hyvärinen

    2016-01-01

    Full Text Available A comprehensive analysis of the structure of the nuclear matrix elements (NMEs of neutrinoless double beta-minus (0νβ-β- decays to the 0+ ground and first excited states is performed in terms of the contributing multipole states in the intermediate nuclei of 0νβ-β- transitions. We concentrate on the transitions mediated by the light (l-NMEs Majorana neutrinos. As nuclear model we use the proton-neutron quasiparticle random-phase approximation (pnQRPA with a realistic two-nucleon interaction based on the Bonn one-boson-exchange G matrix. In the computations we include the appropriate short-range correlations, nucleon form factors, and higher-order nucleonic weak currents and restore the isospin symmetry by the isoscalar-isovector decomposition of the particle-particle proton-neutron interaction parameter gpp.

  10. Amplitude analysis of the anomalous decay η'→π+π-γ

    Science.gov (United States)

    Dai, Ling-Yun; Kang, Xian-Wei; Meißner, Ulf-G.; Song, Xin-Ying; Yao, De-Liang

    2018-02-01

    In this paper we perform an amplitude analysis of η'→π+π-γ and confront it with the latest BESIII data. Based on the final-state interaction theorem, we represent the amplitude in terms of an Omnés function multiplied by a form factor that corresponds to the contributions from left-hand cuts and right-hand cuts in the inelastic channels. We also take into account the isospin violation effect induced by ρ -ω mixing. Our results show that the anomaly contribution is mandatory in order to explain the data. Its contribution to the decay width of Γ (η'→π π γ ) is larger than that induced by isospin violation. Finally we extract the pole positions of the ρ and ω as well as their corresponding residues.

  11. Frequency and risk indicators of tooth decay among pregnant women in France: a cross-sectional analysis.

    Directory of Open Access Journals (Sweden)

    Jean-Noel Vergnes

    Full Text Available INTRODUCTION: Little is known on the prevalence of tooth decay among pregnant women. Better knowledge of tooth decay risk indicators during pregnancy could help to develop follow-up protocols for women at risk, along with better prevention strategies. The aim of this study was to assess the frequency of tooth decay and the number of decayed teeth per woman in a large sample of pregnant women in France, and to study associated risk indicators. METHODS: A secondary cross-sectional analysis of data from a French multicentre case-control study was performed. The sample was composed of 1094 at-term women of six maternity units. A dental examination was carried out within 2 to 4 days post-partum. Socio-demographic and behavioural characteristics were obtained through a standardised interview with the women. Medical characteristics were obtained from the women's medical records. Risk indicators associated with tooth decay were identified using a negative binomial hurdle model. RESULTS: 51.6% of the women had tooth decay. The mean number of decayed teeth among women having at least one was 3.1 (s.d. = 2.8. Having tooth decay was statistically associated with lower age (aOR = 1.58, 95%CI [1.03,2.45], lower educational level (aOR = 1.53, 95%CI [1.06,2.23] and dental plaque (aOR = 1.75, 95%CI [1.27,2.41]. The number of decayed teeth was associated with the same risk indicators and with non-French nationality and inadequate prenatal care. DISCUSSION: The frequency of tooth decay and the number of decayed teeth among pregnant women were high. Oral health promotion programmes must continue to inform women and care providers about the importance of dental care before, during and after pregnancy. Future research should also assess the effectiveness of public policies related to oral health in target populations of pregnant women facing challenging social or economic situations.

  12. Fourier analysis of conductive heat transfer for glazed roofing materials

    Energy Technology Data Exchange (ETDEWEB)

    Roslan, Nurhana Lyana; Bahaman, Nurfaradila; Almanan, Raja Noorliyana Raja; Ismail, Razidah [Faculty of Computer and Mathematical Sciences, Universiti Teknologi MARA, 40450 Shah Alam, Selangor (Malaysia); Zakaria, Nor Zaini [Faculty of Applied Sciences, Universiti Teknologi MARA, 40450 Shah Alam, Selangor (Malaysia)

    2014-07-10

    For low-rise buildings, roof is the most exposed surface to solar radiation. The main mode of heat transfer from outdoor via the roof is conduction. The rate of heat transfer and the thermal impact is dependent on the thermophysical properties of roofing materials. Thus, it is important to analyze the heat distribution for the various types of roofing materials. The objectives of this paper are to obtain the Fourier series for the conductive heat transfer for two types of glazed roofing materials, namely polycarbonate and polyfilled, and also to determine the relationship between the ambient temperature and the conductive heat transfer for these materials. Ambient and surface temperature data were collected from an empirical field investigation in the campus of Universiti Teknologi MARA Shah Alam. The roofing materials were installed on free-standing structures in natural ventilation. Since the temperature data are generally periodic, Fourier series and numerical harmonic analysis are applied. Based on the 24-point harmonic analysis, the eleventh order harmonics is found to generate an adequate Fourier series expansion for both glazed roofing materials. In addition, there exists a linear relationship between the ambient temperature and the conductive heat transfer for both glazed roofing materials. Based on the gradient of the graphs, lower heat transfer is indicated through polyfilled. Thus polyfilled would have a lower thermal impact compared to polycarbonate.

  13. Hybrid Ventilation with Innovative Heat Recovery—A System Analysis

    Directory of Open Access Journals (Sweden)

    Bengt Hellström

    2013-02-01

    Full Text Available One of the most important factors when low energy houses are built is to have good heat recovery on the ventilation system. However, standard ventilation units use a considerable amount of electricity. This article discusses the consequences on a system level of using hybrid ventilation with heat recovery. The simulation program TRNSYS was used in order to investigate a ventilation system with heat recovery. The system also includes a ground source storage and waste water heat recovery system. The result of the analysis shows that the annual energy gain from ground source storage is limited. However, this is partly a consequence of the fact that the well functioning hybrid ventilation system leaves little room for improvements. The analysis shows that the hybrid ventilation system has potential to be an attractive solution for low energy buildings with a very low need for electrical energy.

  14. Comparative technical-economic analysis of the low temperature heating systems

    International Nuclear Information System (INIS)

    Sharevski, Vasko; Sharevski, Milan

    1994-01-01

    A method for comparative technical-economic analysis between low temperature heating systems and heating systems with fossil fuel boiler plant, heat pump heating system and electrical heating systems is presented. The single and combined heating systems are analyzed. The technical-economic priority application of the heating system is determined according to the prices of the low temperature heat energy, fossil fuel heat energy, electrical energy, as well as to the coefficient of the annual use of the installed heating capacity, investment expenses, structure of the combined heating system and coefficient of performances of the heat pump. The combined heating system, composed with a low temperature heating subsystem, which is used to cover the base heat demands, and a oil boiler plant heating subsystem, for the top heat demands, have technical-economic justification and wide range of priority application, in comparison with single heating systems. (author)

  15. Angular analysis of B→φK{sup *} decays and search for CP violation at the Belle experiment

    Energy Technology Data Exchange (ETDEWEB)

    Prim, Michael

    2013-11-08

    The thesis presents an angular analysis of B → φK{sup *} decays and search for CP violation at the Belle experiment. A partial wave analysis is used to separate individual K{sup *} components and perform a coherent amplitude analysis. Further, multivariate methods are applied to enrich the signal contribution. The thesis also presents methods to optimize numeric integrations in amplitude analysis and discusses methods to quantify correlations in multivariate data sets.

  16. Experimental analysis of the double Cabibbo-suppressed weak decay D+ -> φ K+

    International Nuclear Information System (INIS)

    Lima, Jose Guilherme Rocha de.

    1991-03-01

    A phenomenological estimate of the branching fraction of the weak decay D + -> θ K + is made, which cannot be explained only through spectator processes and which is a rare process because it is a doubly Cabibbo suppressed decay. After describing the Charm Photoproduction Experiment E691, we present the results obtained in the search for this decay in the data collected by E691. These results suggest a significant contribution from W annihilation processes for the decays of the charmed meson D + . (author). 41 refs, 37 figs, 17 tabs

  17. Spatial temporal analysis of urban heat hazard in Tangerang City

    Science.gov (United States)

    Wibowo, Adi; Kuswantoro; Ardiansyah; Rustanto, Andry; Putut Ash Shidiq, Iqbal

    2016-11-01

    Urban heat is a natural phenomenon which might caused by human activities. The human activities were represented by various types of land-use such as urban and non-urban area. The aim of this study is to identify the urban heat behavior in Tangerang City as it might threats the urban environment. This study used three types of remote sensing data namely, Landsat TM, Landsat ETM+ and Landsat OLI-TIRS, to capture the urban heat behavior and to analysis the urban heat signature of Tangerang City in 2001, 2012, 2013, 2014, 2015 and 2016. The result showed that urban heat signature change dynamically each month based on the sun radiation. The urban heat island covered only small part of Tangerang City in 2001, but it was significantly increased and reached 50% of the area in 2012. Based on the result on urban heat signature, the threshold for threatening condition is 30 oC which recognized from land surface temperature (LST). The effective temperature (ET) index explains that condition as warm, uncomfortable, increase stress due to sweating and blood flow and may causing cardiovascular disorder.

  18. Application of transient analysis methodology to heat exchanger performance monitoring

    International Nuclear Information System (INIS)

    Rampall, I.; Soler, A.I.; Singh, K.P.; Scott, B.H.

    1994-01-01

    A transient testing technique is developed to evaluate the thermal performance of industrial scale heat exchangers. A Galerkin-based numerical method with a choice of spectral basis elements to account for spatial temperature variations in heat exchangers is developed to solve the transient heat exchanger model equations. Testing a heat exchanger in the transient state may be the only viable alternative where conventional steady state testing procedures are impossible or infeasible. For example, this methodology is particularly suited to the determination of fouling levels in component cooling water system heat exchangers in nuclear power plants. The heat load on these so-called component coolers under steady state conditions is too small to permit meaningful testing. An adequate heat load develops immediately after a reactor shutdown when the exchanger inlet temperatures are highly time-dependent. The application of the analysis methodology is illustrated herein with reference to an in-situ transient testing carried out at a nuclear power plant. The method, however, is applicable to any transient testing application

  19. An immersed-boundary method for conjugate heat transfer analysis

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jeong Chul; Lee, Joon Sik [Seoul National University, Seoul (Korea, Republic of); Ahn, Joon [Kookmin University, Seoul (Korea, Republic of)

    2017-05-15

    An immersed-boundary method is proposed for the analysis of conjugate problems of convective heat transfer in conducting solids. In- side the solid body, momentum forcing is applied to set the velocity to zero. A thermal conductivity ratio and a heat capacity ratio, between the solid body and the fluid, are introduced so that the energy equation is reduced to the heat diffusion equation. At the solid fluid interface, an effective conductivity is introduced to satisfy the heat flux continuity. The effective thermal conductivity is obtained by considering the heat balance at the interface or by using a harmonic mean formulation. The method is first validated against the analytic solution to the heat transfer problem in a fully developed laminar channel flow with conducting solid walls. Then it is applied to a laminar channel flow with a heated, block-shaped obstacle to show its validity for geometry with sharp edges. Finally the validation for a curvilinear solid body is accomplished with a laminar flow through arrayed cylinders.

  20. To the safety conception of the high temperature reactor with natural heat removal decay in teh case of accidents

    International Nuclear Information System (INIS)

    Petersen, K.

    1983-10-01

    On September 22, 1970, for the first time an accident simulation experiment with complete failure of the forced core cooling and the nuclear shut-down system was performed in the AVR-reactor: Due to a small heat-up of the fuel the nuclear chain-reaction was interrupted and an overheating of the core and structure was prevented due to the natural heat-convection. On the basis of the meanwhile developed computer-methods and accompanying experimental investigations it is now possible to determine exactly the behaviour of the non actively controlled core of the high temperature reactor, and to understand better the course of the AVR-experiments. On the same basis the potential and the limits of the safety conception realized in the AVR with self-stabilization in the case of accident can be determined. Such a small high temperature reactor as for example the HTR-modul of the KWU, which is characterized by a reliable and simple safety-technique with a minimum of expensive active systems, can be realized using a 2-zone-core up to a unit size of nearly 250 MW(th). (orig.) [de

  1. Reaction theory for analysis of nuclear giant resonances production and decay processes

    International Nuclear Information System (INIS)

    Foglia, G.A.

    1991-01-01

    The existence of mixing parameters connected to the different decay forms of the giant resonances was theoretically justified, and their energy dependence determined as well using a reaction theory which treats in a consistent manner the giant multipolar resonances formation and their different decay modes. (L.C.J.A.)

  2. Startup analysis for a high temperature gas loaded heat pipe

    Science.gov (United States)

    Sockol, P. M.

    1973-01-01

    A model for the rapid startup of a high-temperature gas-loaded heat pipe is presented. A two-dimensional diffusion analysis is used to determine the rate of energy transport by the vapor between the hot and cold zones of the pipe. The vapor transport rate is then incorporated in a simple thermal model of the startup of a radiation-cooled heat pipe. Numerical results for an argon-lithium system show that radial diffusion to the cold wall can produce large vapor flow rates during a rapid startup. The results also show that startup is not initiated until the vapor pressure p sub v in the hot zone reaches a precise value proportional to the initial gas pressure p sub i. Through proper choice of p sub i, startup can be delayed until p sub v is large enough to support a heat-transfer rate sufficient to overcome a thermal load on the heat pipe.

  3. Performance Analysis of Waste Heat Driven Pressurized Adsorption Chiller

    KAUST Repository

    LOH, Wai Soong

    2010-01-01

    This article presents the transient modeling and performance of waste heat driven pressurized adsorption chillers for refrigeration at subzero applications. This innovative adsorption chiller employs pitch-based activated carbon of type Maxsorb III (adsorbent) with refrigerant R134a as the adsorbent-adsorbate pair. It consists of an evaporator, a condenser and two adsorber/desorber beds, and it utilizes a low-grade heat source to power the batch-operated cycle. The ranges of heat source temperatures are between 55 to 90°C whilst the cooling water temperature needed to reject heat is at 30°C. A parametric analysis is presented in the study where the effects of inlet temperature, adsorption/desorption cycle time and switching time on the system performance are reported in terms of cooling capacity and coefficient of performance. © 2010 by JSME.

  4. Analysis of Hydrodynamics and Heat Transfer in a Thin Liquid Film Flowing over a Rotating Disk by Integral Method

    Science.gov (United States)

    Basu, S.; Cetegen, B. M.

    2005-01-01

    An integral analysis of hydrodynamics and heat transfer in a thin liquid film flowing over a rotating disk surface is presented for both constant temperature and constant heat flux boundary conditions. The model is found to capture the correct trends of the liquid film thickness variation over the disk surface and compare reasonably well with experimental results over the range of Reynolds and Rossby numbers covering both inertia and rotation dominated regimes. Nusselt number variation over the disk surface shows two types of behavior. At low rotation rates, the Nusselt number exhibits a radial decay with Nusselt number magnitudes increasing with higher inlet Reynolds number for both constant wall temperature and heat flux cases. At high rotation rates, the Nusselt number profiles exhibit a peak whose location advances radially outward with increasing film Reynolds number or inertia. The results also compare favorably with the full numerical simulation results from an earlier study as well as with the reported experimental results.

  5. Heat Transfer Analysis of a Diesel Engine Head

    Directory of Open Access Journals (Sweden)

    M. Diviš

    2003-01-01

    Full Text Available This paper documents the research carried out at the Josef Božek Research Center of Engine and Automotive Engineering dealing with extended numerical stress/deformation analyses of engines parts loaded by heat and mechanical forces. It contains a detailed description of a C/28 series diesel engine head FE model and a discussion of heat transfer analysis tunning and results. The head model consisting of several parts allows a description of contact interaction in both thermal and mechanical analysis.

  6. Dalitz plot analysis of B-s(0) -> (D)over-bar(0)K(-)pi(+) decays

    NARCIS (Netherlands)

    Aaij, R.; Adeva, B.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Akar, S.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Alvarez Cartelle, P.; Alves, A. A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Anderson, J.; Andreassen, R.; Andreotti, M.; Andrews, J. E.; Appleby, R. B.; Gutierrez, O. Aquines; Archilli, F.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J. J.; Badalov, A.; Baesso, C.; Baldini, W.; Barlow, R. J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Battista, V.; Bay, A.; Beaucourt, L.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Belogurov, S.; Belous, K.; Onderwater, G.; Pellegrino, A.

    2014-01-01

    The resonant substructure of B-s(0) -> (D) over bar K-0(-)pi(+) decays is studied with the Dalitz plot analysis technique. The study is based on a data sample corresponding to an integrated luminosity of 3.0 fb(-1) of pp collision data recorded by LHCb. A structure at m((D) over bar K-0(-))

  7. Angular analysis of the B-0 -> K*(0) e(+) e(-) decay in the low-q(2) region

    NARCIS (Netherlands)

    Aaij, R.; Adeva, B.; Adinolfi, M.; Affolder, A.; Ajaltouni, Z.; Akar, S.; Albrecht, J.; Alessio, F.; Alexander, M.; Ali, S.; Alkhazov, G.; Alvarez Cartelle, P.; Alves, A. A.; Amato, S.; Amerio, S.; Amhis, Y.; An, L.; Anderlini, L.; Andreassen, R.; Andreotti, M.; Andrews, J. E.; Appleby, R. B.; Gutierrez, O. Aquines; Archilli, F.; Artamonov, A.; Artuso, M.; Aslanides, E.; Auriemma, G.; Baalouch, M.; Bachmann, S.; Back, J. J.; Badalov, A.; Baesso, C.; Baldini, W.; Barlow, R. J.; Barschel, C.; Barsuk, S.; Barter, W.; Batozskaya, V.; Battista, V.; Beaucourt, L.; Beddow, J.; Bedeschi, F.; Bediaga, I.; Belogurov, S.; Belyaev, I.; Ben-Haim, E.; Bencivenni, G.; Benson, S.; Benton, J.; Berezhnoy, A.; Bernet, R.; Bertolin, A.; Bettler, M. -O.; van Beuzekom, M.; Bien, A.; Bifani, S.; Bird, T.; Bizzeti, A.; Blake, T.; Blanc, F.; Blouw, J.; Blusk, S.; Bocci, V.; Bondar, A.; Bondar, N.; Bonivento, W.; Borghi, S.; Borgia, A.; Borsato, M.; Bowcock, T. J. V.; Bowen, E.; Bozzi, C.; Brett, D.; Britsch, M.; Britton, T.; Brodzicka, J.; Brook, N. H.; Bursche, A.; Buytaert, J.; Cadeddu, S.; Calabrese, R.; Calvi, M.; Calvo Gomez, M.; Campana, P.; Perez, D. Campora; Capriotti, L.; Carbone, A.; Carboni, G.; Cardinale, R.; Cardini, A.; Carson, L.; Carvalho Akiba, K.; Casanova Mohr, R. C. M.; Casse, G.; Cassina, L.; Garcia, L. Castillo; Cattaneo, M.; Cauet, Ch.; Cenci, R.; Charles, M.; Charpentier, Ph.; Chefdeville, M.; Cheung, S. -F.; Chiapolini, N.; Chrzaszcz, M.; Vidal, X. Cid; Ciezarek, G.; Clarke, P. E. L.; Clemencic, M.; Cliff, H. V.; Closier, J.; Coco, V.; Cogan, J.; Cogneras, E.; Cogoni, V.; Cojocariu, L.; Collazuol, G.; Collins, P.; Comerma-Montells, A.; Contu, A.; Cook, A.; Coombes, M.; Coquereau, S.; Corti, G.; Corvo, M.; Counts, I.; Couturier, B.; Cowan, G. A.; Craik, D. C.; Crocombe, A. C.; Cruz Torres, M.; Cunliffe, S.; Currie, R.; D'Ambrosio, C.; Dalseno, J.; David, P.; David, P. N. Y.; Davis, A.; De Bruyn, K.; De Capua, S.; De Cian, M.; De Miranda, J. M.; De Paula, L.; De Silva, W.; De Simone, P.; Dean, C. -T.; Decamp, D.; Deckenhoff, M.; Del Buono, L.; Deleage, N.; Derkach, D.; Deschamps, O.; Dettori, F.; Dey, B.; Di Canto, A.; Di Domenico, A.; Di Ruscio, F.; Donleavy, S.; Dordei, F.; Dorigo, M.; Dosil Suarez, A.; Dossett, D.; Dovbnya, A.; Dreimanis, K.; Dujany, G.; Dupertuis, F.; Durante, P.; Dzhelyadin, R.; Dziurda, A.; Dzyuba, A.; Easo, S.; Egede, U.; Egorychev, V.; Eidelman, S.; Eisenhardt, S.; Eitschberger, U.; Ekelhof, R.; Eklund, L.; El Rifai, I.; Elsasser, Ch.; Ely, S.; Esen, S.; Evans, H. M.; Evans, T.; Falabella, A.; Faerber, C.; Farinelli, C.; Farley, N.; Farry, S.; Fay, R.; Ferguson, D.; Fernandez Albor, V.; Ferreira Rodrigues, F.; Ferro-Luzzi, M.; Filippov, S.; Fiore, M.; Fiorini, M.; Firlej, M.; Fitzpatrick, C.; Fiutowski, T.; Fol, P.; Fontana, M.; Fontanelli, F.; Forty, R.; Francisco, O.; Frei, C.; Frosini, M.; Furfaro, E.; Gallas Torreira, A.; Galli, D.; Gallorini, S.; Gambetta, S.; Gandelman, M.; Gandini, P.; Garcia Pardinas, J.; Garofoli, J.; Tico, J. Garra; Garrido, L.; Gascon, D.; Gaspar, C.; Gastaldi, U.; Gauld, R.; Gavardi, L.; Gazzoni, G.; Geraci, A.; Gersabeck, E.; Gersabeck, M.; Gershon, T.; Ghez, Ph.; Gianelle, A.; Giani, S.; Gibson, V.; Giubega, L.; Gligorov, V. V.; Goebel, C.; Golubkov, D.; Golutvin, A.; Gotti, C.; Gandara, M. Grabalosa; Graciani Diaz, R.; Cardoso, L. A. Granado; Grauges, E.; Graverini, E.; Graziani, G.; Grecu, A.; Greening, E.; Gregson, S.; Griffith, P.; Grillo, L.; Gruenberg, O.; Gui, B.; Gushchin, E.; Guz, Yu.; Gys, T.; Hadjivasiliou, C.; Haefeli, G.; Haen, C.; Haines, S. C.; Hall, S.; Hamilton, B.; Hampson, T.; Han, X.; Hansmann-Menzemer, S.; Harnew, N.; Harnew, S. T.; Harrison, J.; He, J.; Head, T.; Heijne, V.; Hennessy, K.; Henrard, P.; Henry, L.; Hernando Morata, J. A.; van Herwijnen, E.; Hess, M.; Hicheur, A.; Hill, D.; Hoballah, M.; Hombach, C.; Hulsbergen, W.; Hussain, N.; Hutchcroft, D.; Hynds, D.; Idzik, M.; Ilten, P.; Jacobsson, R.; Jalocha, J.; Jans, E.; Jawahery, A.; Jing, F.; John, M.; Johnson, D.; Jones, C. R.; Joram, C.; Jost, B.; Jurik, N.; Kandybei, S.; Kanso, W.; Karacson, M.; Karbach, T. M.; Karodia, S.; Kelsey, M.; Kenyon, I. R.; Kenzie, M.; Ketel, T.; Khanji, B.; Khurewathanakul, C.; Klaver, S.; Klimaszewski, K.; Kochebina, O.; Kolpin, M.; Komarov, I.; Koppenburg, P.; Korolev, M.; Kravchuk, L.; Kreplin, K.; Kreps, M.; Krocker, G.; Krokovny, P.; Kucewicz, W.; Kucharczyk, M.; Kudryavtsev, V.; Kurek, K.; Kvaratskheliya, T.; La Thi, V. N.; Lacarrere, D.; Lafferty, G.; Lai, A.; Lambert, D.; Lambert, R. W.; Lanfranchi, G.; Langenbruch, C.; Langhans, B.; Latham, T.; Lazzeroni, C.; Le Gac, R.; van Leerdam, J.; Lees, J. -P.; Lefevre, R.; Leflat, A.; Lefrancois, J.; Leroy, O.; Lesiak, T.; Leverington, B.; Likhomanenko, T.; Liles, M.; Lindner, R.; Linn, C.; Lionetto, F.; Lohn, S.; Longstaff, I.; Lopes, J. H.; Lowdon, P.; Lucchesi, D.; Luo, H.; Lupato, A.; Luppi, E.; Lupton, O.; Machefert, F.; Machikhiliyan, I. V.; Maciuc, F.; Maev, O.; Malde, S.; Malinin, A.; Manca, G.; Mancinelli, G.; Manning, P.; Mapelli, A.; Maratas, J.; Marchand, J. F.; Marconi, U.; Marin Benito, C.; Marino, P.; Maerki, R.; Marks, J.; Martellotti, G.; Martinelli, M.; Santos, D. Martinez; Martinez Vidal, F.; Martins Tostes, D.; Massafferri, A.; Matev, R.; Mathe, Z.; Matteuzzi, C.; Maurin, B.; Mazurov, A.; McCann, M.; McCarthy, J.; McNab, A.; McNulty, R.; McSkelly, B.; Meadows, B.; Meier, F.; Merk, M.; Milanes, D. A.; Minard, M. -N.; Moggi, N.; Molina Rodriguez, J.; Monteil, S.; Morandin, M.; Morawski, P.; Morda, A.; Morello, M. J.; Moron, J.; Morris, A. -B.; Mountain, R.; Muheim, F.; Mueller, K.; Mussini, M.; Muster, B.; Naik, P.; Nakada, T.; Nandakumar, R.; Nasteva, I.; Needham, M.; Neri, N.; Neubert, S.; Neufeld, N.; Neuner, M.; Nguyen, A. D.; Nguyen-Mau, C.; Nicol, M.; Niess, V.; Niet, R.; Nikitin, N.; Nikodem, T.; Novoselov, A.; O'Hanlon, D. P.; Oblakowska-Mucha, A.; Obraztsov, V.; Ogilvy, S.; Okhrimenko, O.; Oldeman, R.; Onderwater, C. J. G.; Orlandea, M.; Osorio Rodrigues, B.; Otalora Goicochea, J. M.; Otto, A.; Owen, P.; Oyanguren, A.; Pal, B. K.; Palano, A.; Palombo, F.; Palutan, M.; Papanestis, A.; Pappagallo, M.; Pappalardo, L. L.; Parkes, C.; Parkinson, C. J.; Passaleva, G.; Patel, G. D.; Patrignani, C.; Pearce, A.; Pellegrino, A.; Penso, G.; Altarelli, M. Pepe; Perazzini, S.; Perret, P.; Pescatore, L.; Pesen, E.; Petridis, K.; Petrolini, A.; Picatoste Olloqui, E.; Pietrzyk, B.; Pilar, T.; Pinci, D.; Pistone, A.; Playfer, S.; Plo Casasus, M.; Polci, F.; Poluektov, A.; Polyakov, I.; Polycarpo, E.; Popov, A.; Popov, D.; Popovici, B.; Potterat, C.; Price, E.; Price, J. D.; Prisciandaro, J.; Pritchard, A.; Prouve, C.; Pugatch, V.; Navarro, A. Puig; Punzi, G.; Qian, W.; Quagliani, R.; Rachwal, B.; Rademacker, J. H.; Rakotomiaramanana, B.; Rama, M.; Rangel, M. S.; Raniuk, I.; Rauschmayr, N.; Raven, G.; Redi, F.; Reichert, S.; Reid, M. M.; dos Reis, A. C.; Ricciardi, S.; Richards, S.; Rihl, M.; Rinnert, K.; Rives Molina, V.; Rodrigues, A. B.; Rodrigues, E.; Perez, P. Rodriguez; Roiser, S.; Romanovsky, V.; Romero Vidal, A.; Rotondo, M.; Rouvinet, J.; Ruf, T.; Ruiz, H.; Ruiz Valls, P.; Saborido Silva, J. J.; Sagidova, N.; Sail, P.; Saitta, B.; Salustino Guimaraes, V.; Sanchez Mayordomo, C.; Sanmartin Sedes, B.; Santacesaria, R.; Santamarina Rios, C.; Santovetti, E.; Sarti, A.; Satriano, C.; Satta, A.; Saunders, D. M.; Savrina, D.; Schindler, H.; Schlupp, M.; Schmelling, M.; Schmidt, B.; Schneider, O.; Schopper, A.; Schune, M. -H.; Schwemmer, R.; Sciascia, B.; Sciubba, A.; Semennikov, A.; Sepp, I.; Serra, N.; Serrano, J.; Sestini, L.; Seyfert, P.; Shapkin, M.; Shapoval, I.; Shcheglov, Y.; Shears, T.; Shekhtman, L.; Shevchenko, V.; Shires, A.; Coutinho, R. Silva; Simi, G.; Sirendi, M.; Skidmore, N.; Skillicorn, I.; Skwarnicki, T.; Smith, N. A.; Snoek, H.; Sokoloff, M. D.; Soler, F. J. P.; Soomro, F.; Souza, D.; Souza De Paula, B.; Spaan, B.; Spradlin, P.; Sridharan, S.; Stagni, F.; Stahl, M.; Stahl, S.; Steinkamp, O.; Stenyakin, O.; Sterpka, F.; Stevenson, S.; Stone, S.; Storaci, B.; Stracka, S.; Straticiuc, M.; Straumann, U.; Stroili, R.; Sutcliffe, W.; Swientek, K.; Swientek, S.; Syropoulos, V.; Szczekowski, M.; Szczypka, P.; Szumlak, T.; T'Jampens, S.; Teklishyn, M.; Tellarini, G.; Teubert, F.; van Tilburg, J.; Tisserand, V.; Tobin, M.; Todd, J.; Tolk, S.; Tomassetti, L.; Tonelli, D.; Topp-Joergensen, S.; Torr, N.; Tournefier, E.; Tourneur, S.; Trabelsi, K.; Tresch, M.; Trisovic, A.; Tsaregorodtsev, A.; Tsopelas, P.; Tuning, N.; Garcia, M. Ubeda; Ukleja, A.; Ustyuzhanin, A.; Uwer, U.; Vacca, C.; Vagnoni, V.; Valenti, G.; Vallier, A.; Gomez, R. Vazquez; Vazquez Regueiro, P.; Vazquez Sierra, C.; Vecchi, S.; Velthuis, J. J.; Veltri, M.; Veneziano, G.; Vesterinen, M.; Barbosa, J. V. V. B. Viana; Viaud, B.; Vieira, D.; Vieites Diaz, M.; Vilasis-Cardona, X.; Vollhardt, A.; Volyanskyy, D.; Voong, D.; Vorobyev, A.; Vorobyev, V.; Voss, C.; Waldi, R.; Wallace, C.; Wallace, R.; Wandernoth, S.; Ward, D. R.; Watson, N. K.; Websdale, D.; Whitehead, M.; Wiedner, D.; Wilkinson, G.; Wilschut, H. W.; Wilson, F. F.; Wimberley, J.; Wishahi, J.; Wislicki, W.; Witek, M.; Wormser, G.; Wotton, S. A.; Wright, S.; Wyllie, K.; Xie, Y.; Xing, Z.; Yang, Z.; Yuan, X.; Yushchenko, O.; Zangoli, M.; Zavertyaev, M.; Zhang, W. C.; Zhelezov, A.; Zhokhov, A.; Zhong, L.

    2015-01-01

    An angular analysis of the B-0 -> K(*0)e(+) e(-) decay is performed using a data sample, corresponding to an integrated luminosity of 3.0 fb(-1), collected by the LHCb experiment in pp collisions at centre-of-mass energies of 7 and 8TeV during 2011 and 2012. For the first time several observables

  8. The Monitoring and Analysis of the Chinese Traditional Architecture Painting Decay - the Case of Jilan Pavilion in the Summer Palace

    Science.gov (United States)

    Zhan, L.; Wang, Z.; Zhao, X.; Liu, Z.; Liu, G.

    2017-08-01

    The painting is one of the most decorative elements of ancient Chinese architecture and it is sensitive to the natural environment. The outdoor painting is very easily to be affected by the natural environment and its color is easily to decay. In order to study the relationship between the painting decay and the physical environment, the two paintings on both east and west inner eaves of Jilan Pavilion beside the Kunming Lake in Summer Palace are taken as the research objects, because these two paintings are painted at the same period but their decay degrees are greatly different. Since 2013, the research group regularly monitors the color decay of these two paintings. Meanwhile, the physical environment around these two paintings has been monitored. Based on the analysis of the monitor data, it can be seen that the illumination is the major factor that causes the decay of the painting. Meanwhile, by using the ECOTECT software to stimulate the light environment around the Jilan Pavilion, it can be seen that it is the mirror reflection of the lake that causes the result that the illumination of the west eave (face east) is stronger than that of the east eave (face west).

  9. Feasibility analysis of reciprocating magnetic heat pumps

    Science.gov (United States)

    Larson, A. V.; Hartley, J. G.; Shelton, Sam V.; Smith, M. M.

    1989-01-01

    A reciprocating gadolinium core in a regeneration fluid column in the warm bore of a superconducting solenoidal magnet is considered for magnetic refrigeration in 3.517 MW (1000 ton) applications. A procedure is presented to minimize the amount of superconducting cable needed in the magnet design. Estimated system capital costs for an ideal magnetic refrigerator of this type become comparable to conventional chillers as the frequency of reciprocation approaches 10 Hertz. A 1-D finite difference analysis of a regenerator cycling at 0.027 Hertz is presented which exhibits some of the features seen in the experiments of G. V. Brown.

  10. 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.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-02-01

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

  13. 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.

  14. Design of cycler trajectories and analysis of solar influences on radioactive decay rates during space missions

    Science.gov (United States)

    Rogers, Blake A.

    investigated to determine if they can be used to find new cycler trajectories, as well as those previously discovered. First order approximations to the relative motion equations are unfruitful for Earth-Mars cyclers because the variation in radial distance from the Sun is too large. However, using optimization techniques, cycling trajectories are found for the Earth-Mars, Earth-Ceres, and Mars-Ceres systems. Experiments showing a seasonal variation of the nuclear decay rates of a number of different nuclei and decay anomalies--- apparently related to solar flares and solar rotation--- have suggested that the Sun may somehow be influencing nuclear decay processes. Recently, there have been searches for such an effect in 238Pu nuclei contained in the radioisotope thermoelectric generators on board the Cassini spacecraft. In this work, that analysis is modified and extended to obtain constraints on anomalous decays of 238Pu over a wider range of models, but these limits cannot be applied to other nuclei if the anomaly is composition-dependent. It is also shown that it may require very high sensitivity for terrestrial experiments to discriminate among some models if such a decay anomaly exists, motivating the consideration of future spacecraft experiments which would require less precision. A mission on which such an experiment could be run is proposed. The proposed mission will take various isotopes on a spacecraft that has a large variation in radial distance and return them to Earth. Two different types of trajectories are considered: one with intermediate Venus flybys and one that injects directly into an Earth-resonant orbit. It is shown that each of these types of trajectories have their relative merits with regards to the scientific objective. The suitability of the upcoming Solar Probe Plus and Solar Orbiter missions to perform this experiment is also investigated.

  15. Two-dimensional heat flow analysis applied to heat sterilization of ponderosa pine and Douglas-fir square timbers

    Science.gov (United States)

    William T. Simpson

    2004-01-01

    Equations for a two-dimensional finite difference heat flow analysis were developed and applied to ponderosa pine and Douglas-fir square timbers to calculate the time required to heat the center of the squares to target temperature. The squares were solid piled, which made their surfaces inaccessible to the heating air, and thus surface temperatures failed to attain...

  16. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  17. Feasibility analysis of heat pump dryer to dry hawthorn cake

    Energy Technology Data Exchange (ETDEWEB)

    Wang, D.C., E-mail: wdechang@163.com [College of Electromechanical Engineering, Qingdao University, Qingdao 266071 (China); Zhang, G.; Han, Y.P.; Zhang, J.P.; Tian, X.L. [College of Electromechanical Engineering, Qingdao University, Qingdao 266071 (China)

    2011-08-15

    Highlights: {yields} A heat pump dryer (HPD) is effectively proposed to dry hawthorn cake-likely materials. {yields} Low drying temperature and high COP of heat pump are obtained in drying beginning. {yields} HPD is more effective, economic than a traditional hot air dryer. {yields} Feasibility of the HPD is also validated by the operation economy estimation. - Abstract: A heat pump dryer (HPD) would be an economic, environmentally friendly, hygienic drying machine used to dry some food, such as hawthorn cakes. Based on the production process of the hawthorn cake, a HPD is proposed and its basic principle is introduced. The experimental drying curves of the hawthorn cake using the heat pump drying method and the traditional hot air drying method are compared and analyzed. The drying process of hawthorn cakes is similar to that of the other drying materials. The higher drying temperature causes a faster drying process. But in the initial stage of the heat pump drying process, the water content of the hawthorn cake is not sensitive to the drying temperature, so a lower drying air temperature can be available in order to get a higher coefficient of performance (COP) of the heat pump (HP). The experimental results and the economic analysis indicate that the HPD is feasibly used to dry hawthorn cakes.

  18. Feasibility analysis of heat pump dryer to dry hawthorn cake

    International Nuclear Information System (INIS)

    Wang, D.C.; Zhang, G.; Han, Y.P.; Zhang, J.P.; Tian, X.L.

    2011-01-01

    Highlights: → A heat pump dryer (HPD) is effectively proposed to dry hawthorn cake-likely materials. → Low drying temperature and high COP of heat pump are obtained in drying beginning. → HPD is more effective, economic than a traditional hot air dryer. → Feasibility of the HPD is also validated by the operation economy estimation. - Abstract: A heat pump dryer (HPD) would be an economic, environmentally friendly, hygienic drying machine used to dry some food, such as hawthorn cakes. Based on the production process of the hawthorn cake, a HPD is proposed and its basic principle is introduced. The experimental drying curves of the hawthorn cake using the heat pump drying method and the traditional hot air drying method are compared and analyzed. The drying process of hawthorn cakes is similar to that of the other drying materials. The higher drying temperature causes a faster drying process. But in the initial stage of the heat pump drying process, the water content of the hawthorn cake is not sensitive to the drying temperature, so a lower drying air temperature can be available in order to get a higher coefficient of performance (COP) of the heat pump (HP). The experimental results and the economic analysis indicate that the HPD is feasibly used to dry hawthorn cakes.

  19. Exergy analysis of an experimental heat transformer for water purification

    International Nuclear Information System (INIS)

    Rivera, W.; Huicochea, A.; Martinez, H.; Siqueiros, J.; Juarez, D.; Cadenas, E.

    2011-01-01

    First and second law of thermodynamics have been used to analyze the performance of an experimental heat transformer used for water purification. The pure water is produced in the auxiliary condenser delivering an amount of heat, which is recycled into the heat transformer increasing the heat source temperatures and also the internal, external and exergy coefficients of performance. The theoretical and experimental study was divided into two parts. In the first part, a second law analysis was carried out to the experimental system showing that the absorber and the condenser are the components with the highest irreversibilities. In the second part, with the results obtained from the second law analysis, new test runs were carried out at similar conditions than the former but varying only one selected temperature at the time. Comparing the COP (coefficient of performance) between the old and new test runs, it was shown that higher internal, external and exergy coefficients of performance were obtained in all the new test runs. Also it was shown that the ECOP (exergy coefficient of performance) increases with an increment of the amount of the purified water produced and with the decrease of the flow ratio. -- Research highlights: → By the first time an experimental results of a heat transformer for water purification with heat recycling has been presented. → An exergy analysis has been carried out in order to identify the irreversibilities in the main components of the system. → With the results obtained of the second law analysis new experimental test runs were carried out minimizing the system irreversibilities and furthermore increasing the system efficiency.

  20. Heat analysis of the magnetic limiter plate for JT-60

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ninomiya, Hiromasa; Shimizu, Masatsugu; Ohta, Mitsuru

    1977-03-01

    Heat analysis has been made of the magnetic limiter plate for JT-60. Test materials of the magnetic limiter plate are molybdenum, graphite, pyrolytic graphite and silicon carbide. It is assumed in calculation of the heat analysis that 10MW is deposited on the 2 cm wide surface of the magnetic limiter plate in about 10 sec. The magnetic limiter plate of pyrolytic graphite is a stack of pyrolytic graphite sheets, heat input is in the deposition plane to take advantage of the large heat conductivity along this plane. Pyrolytic graphite is the best in terms of temperature rise. The temperature of molybdenum and graphite rise up to 1800 0 C and 620 0 C, respectively, in an deposition of 10 MWx10sec. Silicon carbide is not suitable for the magnetic limiter plate. Because the plasma of the JT-60 discharges every 10 min, the average heat flux decreases to 17 w/cm 2 during the each interval. When the magnetic limiter plate has the above heat inflow, a maximum of above 1000 0 C occurs at the edge far from the joint to the thick ring of the vacuum vessel. To reduce heat load of the magnetic limiter plate, an alternating current (2 -- 5Hz) is superposed on the magnetic limiter coil current. The intersection of separatrix line and magnetic limiter plate then moves cyclically more than 10 cm. Concerning temperature distribution of the multi-groove magnetic limiter plate, its dimensions are determined by the limitation in vapor pressure to prevent the impurity inflow. (auth.)

  1. Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

    Science.gov (United States)

    Rickman, S. L.; Iamello, C. J.

    2016-01-01

    Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

  2. Convective heat transfer analysis in aggregates rotary drum reactor

    International Nuclear Information System (INIS)

    Le Guen, Laurédan; Huchet, Florian; Dumoulin, Jean; Baudru, Yvan; Tamagny, Philippe

    2013-01-01

    Heat transport characterisation inside rotary drum dryer has a considerable importance linked to many industrial applications. The present paper deals with the heat transfer analysis from experimental apparatus installed in a large-scale rotary drum reactor applied to the asphalt materials production. The equipment including in-situ thermal probes and external visualization by mean of infrared thermography gives rise to the longitudinal evaluation of inner and external temperatures. The assessment of the heat transfer coefficients by an inverse methodology is resolved in order to accomplish a fin analysis of the convective mechanism inside baffled (or flights) rotary drum. The results are discussed and compared with major results of the literature. -- Highlights: ► A thermal and flow experimentation is performed on a large-scale rotary drum. ► Four working points is chosen in the frame of asphalt materials production. ► Evaluation of the convective transfer mechanisms is calculated by inverse method. ► The drying stage is performed in the combustion area. ► Wall/aggregates heat exchanges have a major contribution in the heating stage

  3. Heat transfer and performance analysis of thermoelectric stoves

    International Nuclear Information System (INIS)

    Najjar, Yousef S.H.; Kseibi, Musaab M.

    2016-01-01

    Highlights: • Design and testing of a thermo electric stove. • Three biofuels namely: wood, peat and manure are used. • Heat transfer analysis is detailed. • Resulting thermoelectric energy for vital purposes in remote poor regions. • Evaluation of performance of the stove subcomponents. - Abstract: Access to electricity is one of the important challenges for remote poor regions of the world. Adding TEG (thermoelectric generators) to stoves can provide electricity for the basic benefits such as: operating radio, light, phones, medical instruments and other small electronic devices. Heat transfer analysis of a multi-purpose stove coupled with 12 TEG modules is presented. This analysis comprises a well aerodynamically designed combustor, finned TEG base plate, cooker and water heater beside the outer surface for space heating. Heat transfer analysis was also carried out for all the subcomponents of the stove, and performance predicted against the experimental results. It was found that the maximum power obtained is about 7.88 W using wood, manure or peat with an average overall efficiency of the stove about 60%.

  4. Computational heat transfer analysis and combined ANN–GA ...

    Indian Academy of Sciences (India)

    The analysis using the numerical simulation and neural network ... Optimization is the process of finding the most plausible and desirable solution to a problem. ... increased heat transfer and compared the results of regular non-fuzzy model and fuzzy model. ..... network is designed using MATLAB Neural network toolbox.

  5. 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.

  6. Using Evidence Credibility Decay Model for dependence assessment in human reliability analysis

    International Nuclear Information System (INIS)

    Guo, Xingfeng; Zhou, Yanhui; Qian, Jin; Deng, Yong

    2017-01-01

    Highlights: • A new computational model is proposed for dependence assessment in HRA. • We combined three factors of “CT”, “TR” and “SP” within Dempster–Shafer theory. • The BBA of “SP” is reconstructed by discounting rate based on the ECDM. • Simulation experiments are illustrated to show the efficiency of the proposed method. - Abstract: Dependence assessment among human errors plays an important role in human reliability analysis. When dependence between two sequent tasks exists in human reliability analysis, if the preceding task fails, the failure probability of the following task is higher than success. Typically, three major factors are considered: “Closeness in Time” (CT), “Task Relatedness” (TR) and “Similarity of Performers” (SP). Assume TR is not changed, both SP and CT influence the degree of dependence level and SP is discounted by the time as the result of combine two factors in this paper. In this paper, a new computational model is proposed based on the Dempster–Shafer Evidence Theory (DSET) and Evidence Credibility Decay Model (ECDM) to assess the dependence between tasks in human reliability analysis. First, the influenced factors among human tasks are identified and the basic belief assignments (BBAs) of each factor are constructed based on expert evaluation. Then, the BBA of SP is discounted as the result of combining two factors and reconstructed by using the ECDM, the factors are integrated into a fused BBA. Finally, the dependence level is calculated based on fused BBA. Experimental results demonstrate that the proposed model not only quantitatively describe the fact that the input factors influence the dependence level, but also exactly show how the dependence level regular changes with different situations of input factors.

  7. Technoeconomic analysis of a biomass based district heating system

    International Nuclear Information System (INIS)

    Zhang, H.; Ugursal, V.I.; Fung, A.

    2005-01-01

    This paper discussed a proposed biomass-based district heating system to be built for the Pictou Landing First Nation Community in Nova Scotia. The community centre consists of 6 buildings and a connecting arcade. The methodology used to size and design heating, ventilating and air conditioning (HVAC) systems, as well as biomass district energy systems (DES) were discussed. Annual energy requirements and biomass fuel consumption predictions were presented, along with cost estimates. A comparative assessment of the system with that of a conventional oil fired system was also conducted. It was suggested that the design and analysis methodology could be used for any similar application. The buildings were modelled and simulated using the Hourly Analysis Program (HAP), a detailed 2-in-1 software program which can be used both for HVAC system sizing and building energy consumption estimation. A techno-economics analysis was conducted to justify the viability of the biomass combustion system. Heating load calculations were performed assuming that the thermostat was set constantly at 22 degrees C. Community centre space heating loads due to individual envelope components for 3 different scenarios were summarized, as the design architecture for the buildings was not yet finalized. It was suggested that efforts should be made to ensure air-tightness and insulation levels of the interior arcade glass wall. A hydronic distribution system with baseboard space heating units was selected, comprising of a woodchip boiler, hot water distribution system, convective heating units and control systems. The community has its own logging operation which will provide the wood fuel required by the proposed system. An outline of the annual allowable harvest covered by the Pictou Landing Forestry Management Plan was presented, with details of proposed wood-chippers for the creation of biomass. It was concluded that the woodchip combustion system is economically preferable to the

  8. Flow analysis of an innovative compact heat exchanger channel geometry

    International Nuclear Information System (INIS)

    Vitillo, F.; Cachon, L.; Reulet, F.; Millan, P.

    2016-01-01

    Highlights: • An innovative compact heat transfer technology is proposed. • Experimental measurements are shown to validate the CFD model. • CFD simulations show various flow mechanisms. • Flow analysis is performed to study physical phenomena enhancing heat transfer. - Abstract: In the framework of CEA R&D program to develop an industrial prototype of sodium-cooled fast reactor named ASTRID, the present work aims to propose an innovative compact heat exchanger technology to provide solid technological basis for the utilization of a Brayton gas-power conversion system, in order to avoid the energetic sodium–water interaction if a traditional Rankine cycle was used. The aim of the present work is to propose an innovative compact heat exchanger channel geometry to potentially enhance heat transfer in such components. Hence, before studying the innovative channel performance, a solid experimental and numerical database is necessary to perform a preliminary thermal–hydraulic analysis. To do that, two experimental test sections are used: a Laser Doppler Velocimetry (LDV) test section and a Particle Image Velocimetry (PIV) test section. The acquired experimental database is used to validate the Anisotropic Shear Stress Transport (ASST) turbulence model. Results show a good agreement between LDV, PIV and ASST data for the pure aerodynamic flow. Once validated the numerical model, the innovative channel flow analysis is performed. Principal and secondary flow has been analyzed, showing a high swirling flow in the bend region and demonstrating that mixing actually occurs in the mixing zone. This work has to be considered as a step forward the preposition of a reliable high-performance component for application to ASTRID reactor as well as to any other industrial power plant dealing needing compact heat exchangers.

  9. Thermal diffusivity measurement of erythritol and numerical analysis of heat storage performance on a fin-type heat exchanger

    International Nuclear Information System (INIS)

    Zamengo, Massimiliano; Funada, Tomohiro; Morikawa, Junko

    2017-01-01

    Highlights: • Thermal diffusivity of Erythritol was measured by temperature wave method. • Thermal diffusivity was measured in function of temperature and during phase change. • Database of temperature-dependent thermal properties is used for numerical analysis. • Heat transfer and heat storage were analyzed in a fin-type heat exchanger. • Use of temperature-dependent properties in calculations lead to longer melting time. - Abstract: Temperature dependency of thermal diffusivity of erythritol was measured by temperature wave analysis (TWA) method. This modulating technique allowed measuring thermal diffusivity continuously, even during the phase transition solid-liquid. Together with specific heat capacity and specific enthalpy measured by differential scanning calorimetry, the values of measured properties were utilized in a bi-dimensional numerical model for analysis of heat transfer and heat storage performance. The geometry of the model is representative of a cross section of a fin-type heat exchanger, in which erythritol is filling the interspaces between fins. Time-dependent temperature change and heat storage performance were analyzed by considering the variation of thermophysical properties as a function of temperature. The numerical method can be utilized for a fast parametric analysis of heat transfer and heat storage performance into heat storage systems of phase-change materials and composites.

  10. Absorbed decay-photon dose analysis of the IVVS/GDC plug in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Leichtle, D.; Serikov, A.; Fischer, U. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (DE). Inst. for Neutron Physics and Reactor Technology (INR)

    2011-07-01

    The In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit share a common port at the equatorial level of the ITER tokamak. The plug consists mainly of the IWS probe, capable of performing the laser-based in-vessel viewing and metrology, the GDC electrode, capable of producing glow discharge in the vacuum vessel during intermediate maintenance and wall conditioning periods, and their respective deployment systems to move the electrodes. The plug extends over a length of about 11 m from the GDC tip to the rear end at the bioshield level. At the present stage of the conceptual design a neutronics analysis has been requested to provide valuable input to the design strategy. To this end, a first assessment has been performed focusing on operational loads on the GDC electrode head in the so-called shielding position and on absorbed decay-photon dose rate levels in the structural components of the entire system. In this contribution we are reporting on the absorbed dose rates after the ITER life time irradiation at several cooling times. Gamma sources from activated materials of the IVVS/GDC and surrounding structures, like blanket, vacuum vessel, toroidal and poloidal field coils, have been taken into account. (orig.)

  11. Analysis of K → 3π decays in chiral perturbation theory

    International Nuclear Information System (INIS)

    Cheng Haiyang; Cheung, C.Y.; Yeung Waibong

    1989-01-01

    Using the recently proposed higher-order chiral Lagrangians determined from the integration of nontopological chiral anomalies, we calculate corrections to the current-algebra analysis of K→3π decay amplitudes expanded in powers of the Dalitz variables. Effects of quartic-derivative weak chiral Lagrangians are determined through the use of short-distance effective weak Hamiltonian and the factorization method. We find that (1) the constant and linear terms in the amplitude for ΔI=1/2 K→3π are in excellent agreement with experiment; the previous discrepancy of (20-35)% between current algebra and data is thus accounted for by the higher-order effective Lagrangians, (2) the penguin interaction does not play an essential role in the ΔI=1/2 rule, for otherwise it will lead to a large disagreement for the constant and linear terms, (3) one of the two quadratic terms in the ΔI=1/2 process, which arise from the quartic chiral Lagrangians, is in accord with data within experimental errors, while the other is off by four standard deviations, (4) the linear term in the ΔI=3/2 transitions is in good agreement with experiment and contributions from quadratic terms are sizable. (orig.)

  12. Analysis of discrepancies in Dalitz plot parameters in η→3π decay

    International Nuclear Information System (INIS)

    Kolesár, Marián

    2011-01-01

    We analyze the Dalitz plot parameters of η→3π decay in the framework of resummed chiral perturbation theory. This approach allows us to keep the uncertainties in the NNLO and higher orders under better control and estimate their influence. We cannot confirm the suspected discrepancy in the case of the charged decay parameter b, where even small uncertainties in higher orders could accommodate the difference. On the other hand, we find the experimental value of the neutral decay parameter α incompatible with an assumption of good convergence properties in the center of the Dalitz plot. We calculate ππ rescattering bubble corrections up to three loops and show that these might explain the discrepancy, especially for a low value of the pseudoscalar decay constant in the chiral limit. However, that could indicate a failure of convergence of the chiral series in this channel already at low energies around 500MeV.

  13. Epidemiological Analysis of the Reduction of the Prevalence of Tooth Decay in Infancy

    OpenAIRE

    近藤, 武

    1989-01-01

    It is evident that the prevalence of tooth decay in infancy has been gradually declining throughont Japan. But the cause of this reduction has not been established scientifically. Accordingly, I tried to find some cause of this reduction in tooth decay by comparing the patterns of reduction for several areas with differing populations. It is clear that the reduction rate was very different according to the size of population. The inhabitants living in cities can generally receive good health ...

  14. Inclusive Analysis of the b Quark Fragmentation Function in Z Decays at LEP

    CERN Document Server

    Abbiendi, G.; Akesson, P.F.; Alexander, G.; Allison, John; Amaral, P.; Anagnostou, G.; Anderson, K.J.; Arcelli, S.; Asai, S.; Axen, D.; Azuelos, G.; Bailey, I.; Barberio, E.; Barlow, R.J.; Batley, R.J.; Bechtle, P.; Behnke, T.; Bell, Kenneth Watson; Bell, P.J.; Bella, G.; Bellerive, A.; Benelli, G.; Bethke, S.; Biebel, O.; Bloodworth, I.J.; Boeriu, O.; Bock, P.; Bonacorsi, D.; Boutemeur, M.; Braibant, S.; Brigliadori, L.; Brown, Robert M.; Buesser, K.; Burckhart, H.J.; Cammin, J.; Campana, S.; Carnegie, R.K.; Caron, B.; Carter, A.A.; Carter, J.R.; Chang, C.Y.; Charlton, David G.; Cohen, I.; Csilling, A.; Cuffiani, M.; Dado, S.; Dallavalle, G.Marco; Dallison, S.; De Roeck, A.; De Wolf, E.A.; Desch, K.; Donkers, M.; Dubbert, J.; Duchovni, E.; Duckeck, G.; Duerdoth, I.P.; Elfgren, E.; Etzion, E.; Fabbri, F.; Feld, L.; Ferrari, P.; Fiedler, F.; Fleck, I.; Ford, M.; Frey, A.; Furtjes, A.; Gagnon, P.; Gary, John William; Gaycken, G.; Geich-Gimbel, C.; Giacomelli, G.; Giacomelli, P.; Giunta, Marina; Goldberg, J.; Gross, E.; Grunhaus, J.; Gruwe, M.; Gunther, P.O.; Gupta, A.; Hajdu, C.; Hamann, M.; Hanson, G.G.; Harder, K.; Harel, A.; Harin-Dirac, M.; Hauschild, M.; Hauschildt, J.; Hawkes, C.M.; Hawkings, R.; Hemingway, R.J.; Hensel, C.; Herten, G.; Heuer, R.D.; Hill, J.C.; Hoffman, Kara Dion; Homer, R.J.; Horvath, D.; Howard, R.; Huntemeyer, P.; Igo-Kemenes, P.; Ishii, K.; Jeremie, H.; Jovanovic, P.; Junk, T.R.; Kanaya, N.; Kanzaki, J.; Karapetian, G.; Karlen, D.; Kartvelishvili, V.; Kawagoe, K.; Kawamoto, T.; Keeler, R.K.; Kellogg, R.G.; Kennedy, B.W.; Kim, D.H.; Klein, K.; Klier, A.; Kluth, S.; Kobayashi, T.; Kobel, M.; Kokott, T.P.; Komamiya, S.; Kormos, Laura L.; Kowalewski, Robert V.; Kramer, T.; Kress, T.; Krieger, P.; von Krogh, J.; Krop, D.; Kupper, M.; Kyberd, P.; Lafferty, G.D.; Landsman, H.; Lanske, D.; Layter, J.G.; Leins, A.; Lellouch, D.; Letts, J.; Levinson, L.; Lillich, J.; Lloyd, S.L.; Loebinger, F.K.; Lu, J.; Ludwig, J.; Macpherson, A.; Mader, W.; Marcellini, S.; Marchant, T.E.; Martin, A.J.; Martin, J.P.; Masetti, G.; Mashimo, T.; Mattig, Peter; McDonald, W.J.; McKenna, J.; McMahon, T.J.; McPherson, R.A.; Meijers, F.; Mendez-Lorenzo, P.; Menges, W.; Merritt, F.S.; Mes, H.; Michelini, A.; Mihara, S.; Mikenberg, G.; Miller, D.J.; Moed, S.; Mohr, W.; Mori, T.; Mutter, A.; Nagai, K.; Nakamura, I.; Neal, H.A.; Nisius, R.; O'Neale, S.W.; Oh, A.; Okpara, A.; Oreglia, M.J.; Orito, S.; Pahl, C.; Pasztor, G.; Pater, J.R.; Patrick, G.N.; Pilcher, J.E.; Pinfold, J.; Plane, David E.; Poli, B.; Polok, J.; Pooth, O.; Przybycien, M.; Quadt, A.; Rabbertz, K.; Rembser, C.; Renkel, P.; Rick, H.; Roney, J.M.; Rosati, S.; Rozen, Y.; Runge, K.; Rust, D.R.; Sachs, K.; Saeki, T.; Sahr, O.; Sarkisyan, E.K.G.; Schaile, A.D.; Schaile, O.; Scharff-Hansen, P.; Schieck, J.; Schoerner-Sadenius, Thomas; Schroder, Matthias; Schumacher, M.; Schwick, C.; Scott, W.G.; Seuster, R.; Shears, T.G.; Shen, B.C.; Shepherd-Themistocleous, C.H.; Sherwood, P.; Siroli, G.; Skuja, A.; Smith, A.M.; Sobie, R.; Soldner-Rembold, S.; Spagnolo, S.; Spano, F.; Stahl, A.; Stephens, K.; Strom, David M.; Strohmer, R.; Tarem, S.; Tasevsky, M.; Taylor, R.J.; Teuscher, R.; Thomson, M.A.; Torrence, E.; Toya, D.; Tran, P.; Trefzger, T.; Tricoli, A.; Trigger, I.; Trocsanyi, Z.; Tsur, E.; Turner-Watson, M.F.; Ueda, I.; Ujvari, B.; Vachon, B.; Vollmer, C.F.; Vannerem, P.; Verzocchi, M.; Voss, H.; Vossebeld, J.; Waller, D.; Ward, C.P.; Ward, D.R.; Watkins, P.M.; Watson, A.T.; Watson, N.K.; Wells, P.S.; Wengler, T.; Wermes, N.; Wetterling, D.; Wilson, G.W.; Wilson, J.A.; Wolf, G.; Wyatt, T.R.; Yamashita, S.; Zacek, V.; Zer-Zion, D.; Zivkovic, Lidija

    2003-01-01

    A study of b quark hadronisation is presented using inclusively reconstructed B hadrons in about four million hadronic Z decays recorded in 1992-2000 with the OPAL detector at LEP. The data are compared to different theoretical models, and fragmentation function parameters of these models are fitted. The average scaled energy of weakly decaying B hadrons is determined to be =0.7193+-0.0016(stat)+0.0036-0.0031(syst)

  15. Analysis on flow characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin

    1997-06-01

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

  16. Amplitude Analysis of the Decay $D_s^+ \\to \\pi^+ \\pi^- \\pi^+$ in the Experiment E831/FOCUS

    Energy Technology Data Exchange (ETDEWEB)

    Schilithz, Anderson Correa; /Rio de Janeiro, CBPF

    2005-01-01

    We present in this thesis the Dalitz Plot analysis of the D{sub s}{sup +} {yields} {pi}{sup +}{pi}{sup -}{pi}{sup +} decay, with the data of the E831/FOCUS, that took data in 1996 and 1997. The masses and widhts of f{sub 0}(980) and f{sub 0}(1370) are free parametres of the fit on Dalitz Plot, objectiving to study in detail these resonances. After this analysis we present the Spectator Model study on the S wave in this decay. For this study we used the formalism developed by M. Svec [2] for scattering. We present the comparison between the Isobar Model, frequently used in Dalitz Plot analysis, and this formalism.

  17. Preliminary Analysis on Heat Removal Capacity of Passive Air-Water Combined Cooling Heat Exchanger Using MARS

    International Nuclear Information System (INIS)

    Kim, Seung-Sin; Jeon, Seong-Su; Hong, Soon-Joon; Bae, Sung-Won; Kwon, Tae-Soon

    2015-01-01

    Current design requirement for working time of PAFS heat exchanger is about 8 hours. Thus, it is not satisfied with the required cooling capability for the long term SBO(Station Black-Out) situation that is required to over 72 hours cooling. Therefore PAFS is needed to change of design for 72 hours cooling. In order to acquirement of long terms cooling using PAFS, heat exchanger tube has to be submerged in water tank for long time. However, water in the tank is evaporated by transferred heat from heat exchanger tubes, so water level is gradually lowered as time goes on. The heat removal capacity of air cooling heat exchanger is core parameter that is used for decision of applicability on passive air-water combined cooling system using PAFS in long term cooling. In this study, the development of MARS input model and plant accident analysis are performed for the prediction of the heat removal capacity of air cooling heat exchanger. From analysis result, it is known that inflow air velocity is the decisive factor of the heat removal capacity and predicted air velocity is lower than required air velocity. But present heat transfer model and predicted air velocity have uncertainty. So, if changed design of PAFS that has over 4.6 kW heat removal capacity in each tube, this type heat exchanger can be applied to long term cooling of the nuclear power plant

  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. Numerical Analysis of Microwave Heating on Saponification Reaction

    Science.gov (United States)

    Huang, Kama; Jia, Kun

    2005-01-01

    Currently, microwave is widely used in chemical industry to accelerate chemical reactions. Saponification reaction has important applications in industry; some research results have shown that microwave heating can significantly accelerate the reaction [1]. But so far, no efficient method has been reported for the analysis of the heating process and design of an efficient reactor powered by microwave. In this paper, we present a method to study the microwave heating process on saponification reaction, where the reactant in a test tube is considered as a mixture of dilute solution. According to the preliminary measurement results, the effective permittivity of the mixture is approximately the permittivity of water, but the conductivity, which could change with the reaction, is derived from the reaction equation (RE). The electromagnetic field equation and reaction equation are coupled by the conductivity. Following that, the whole heating processes, which is described by Maxwell's equations, the reaction equation and heat transport equation (HTE), is analyzed by finite difference time domain (FDTD) method. The temperature rising in the test tube are measured and compared with the computational results. Good agreement can be seen between the measured and calculated results.

  20. Parametrical analysis of latent heat and cold storage for heating and cooling of rooms

    International Nuclear Information System (INIS)

    Osterman, E.; Hagel, K.; Rathgeber, C.; Butala, V.; Stritih, U.

    2015-01-01

    One of the problems we are facing today is the energy consumption minimization, while maintaining the indoor thermal comfort in buildings. A potential solution to this issue is use of phase change materials (PCMs) in thermal energy storage (TES), where cold gets accumulated during the summer nights in order to reduce cooling load during the day. In winter, on the other hand, heat from solar air collector is stored for evening and morning hours when solar radiation is not available. The main objective of the paper is to examine experimentally whether it is possible to use such a storage unit for heating as well as for cooling. For this purpose 30 plates filled with paraffin (melting point around 22°C) were positioned into TES and applied with the same initial and boundary conditions as they are expected in reality. Experimental work covered flow visualization, measurements of air velocity in the channels between the plates, parametric analysis in conjunction with TES thermal response and measurements of the pressure drops. The results indicate that this type of storage technology could be advantageously used in real conditions. For optimized thermal behavior, only plate thickness should be reduced. - Highlights: • Thermal properties of paraffin RT22HC were measured. • Flow visualization was carried out and velocity between plates was measured. • Thermal and pressure drop analysis were performed. • Melting times are too long however, use of storage tank for heating and cooling looks promising

  1. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    Energy Technology Data Exchange (ETDEWEB)

    Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

    2017-03-15

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  2. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    International Nuclear Information System (INIS)

    Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

    2017-01-01

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  3. Amplitude Analysis of the Decay $D^+ \\to K_s \\pi^- \\pi^+ \\pi^+$ in the {FOCUS} Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Otalora Goicochea, Juan Martin [Rio de Janeiro, Pont. U. Catol.

    2007-07-01

    This thesis is devoted to the study of the decay D+ → KS π−π+π+ with data collected from the FOCUS experiment, at Fermilab. The fi state composed of 4 pseudo-scalars can be produced through a number of resonant sub-structures. The purpose of this analysis is to fi the contributing intermediate states by measuring their relative strenghts and phases. For that, the Amplitude Analysis formalism is used, with the so-called Isobar Model. The decay dynamics is described through a function which has the features of the contributing channels (functional forms of the resonances, angular distribution, etc) and which domain is a phase space determined by 5 invariants (due to the kinematical constraints of a 4-body spinless decay). The data sample is thus fi to this function. Our results show a dominant contribution of the axial-vector meson a1(1260) (52%), followed by the K1(1400) axial-vector (34%). Moreover, the model presents a contribution from the σ meson (about 8% as a1(σπ)KS and σKS π) and a significant contribution from the scalar κ−. The κ state has been reported in its neutral mode in other charm decays but not is its charged mode. We fi no significant contribution from the direct 4-body decay (non-resonant). This work adds to the effort in the understanding of the strong-interaction dynamics at low energies, which in recent years have been receiving an important contribution from charm meson physics.

  4. Control characteristics and heating performance analysis of automatic thermostatic valves for radiant slab heating system in residential apartments

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Byung-Cheon [Department of Building Equipment System Engineering, Kyungwon University, Seongnam City (Korea); Song, Jae-Yeob [Graduate School, Building Equipment System Engineering, Kyungwon University, Seongnam City (Korea)

    2010-04-15

    Computer simulations and experiments are carried out to research the control characteristics and heating performances for a radiant slab heating system with automatic thermostatic valves in residential apartments. An electrical equivalent R-C circuit is applied to analyze the unsteady heat transfer in the house. In addition, the radiant heat transfer between slabs, ceilings and walls in the room is evaluated by enclosure analysis method. Results of heating performance and control characteristics were determined from control methods such as automatic thermostatic valves, room air temperature-sensing method, water-temperature-sensing method, proportional control method, and On-Off control method. (author)

  5. Angular analysis of the $B^{0}\\rightarrow K^{*0}\\mu^{+}\\mu^{-}$ decay using 3 fb$^{-1}$ of integrated luminosity

    CERN Document Server

    Aaij, Roel; Adeva, Bernardo; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Andreassi, Guido; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; d'Argent, Philippe; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Bel, Lennaert; Bellee, Violaine; Belloli, Nicoletta; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bertolin, Alessandro; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bifani, Simone; Billoir, Pierre; Bird, Thomas; Birnkraut, Alex; Bizzeti, Andrea; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borisyak, Maxim; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Braun, Svende; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Buchanan, Emma; Burr, Christopher; Bursche, Albert; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Capriotti, Lorenzo; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carniti, Paolo; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cavallero, Giovanni; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collazuol, Gianmaria; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Crocombe, Andrew; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dall'Occo, Elena; Dalseno, Jeremy; David, Pieter; Davis, Adam; De Aguiar Francisco, Oscar; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Simone, Patrizia; Dean, Cameron Thomas; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Demmer, Moritz; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Dey, Biplab; Di Canto, Angelo; Di Ruscio, Francesco; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dovbnya, Anatoliy; Dreimanis, Karlis; Dufour, Laurent; Dujany, Giulio; Dungs, Kevin; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Fabianska, Maria; Falabella, Antonio; Färber, Christian; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferrari, Fabio; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fleuret, Frederic; Fohl, Klaus; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forshaw, Dean Charles; Forty, Roger; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garra Tico, Jordi; Garrido, Lluis; Gascon, David; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gazzoni, Giulio; Gerick, David; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianì, Sebastiana; Gibson, Valerie; Girard, Olivier Göran; Giubega, Lavinia-Helena; Gligorov, V.V.; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graverini, Elena; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadavizadeh, Thomas; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Heister, Arno; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Humair, Thibaud; Hushchyn, Mikhail; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jawahery, Abolhassan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kecke, Matthieu; Kelsey, Matthew; Kenyon, Ian; Kenzie, Matthew; Ketel, Tjeerd; Khairullin, Egor; Khanji, Basem; Khurewathanakul, Chitsanu; Kirn, Thomas; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Kozeiha, Mohamad; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krokovny, Pavel; Kruse, Florian; Krzemien, Wojciech; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kuonen, Axel Kevin; Kurek, Krzysztof; Kvaratskheliya, Tengiz; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Lemos Cid, Edgar; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Liu, Xuesong; Loh, David; Longstaff, Iain; Lopes, Jose; Lucchesi, Donatella; Lucio Martinez, Miriam; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Lusiani, Alberto; Machefert, Frederic; Maciuc, Florin; Maev, Oleg; Maguire, Kevin; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Manning, Peter Michael; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Marks, Jörg; Martellotti, Giuseppe; Martin, Morgan; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massacrier, Laure Marie; Massafferri, André; Matev, Rosen; Mathad, Abhijit; Mathe, Zoltan; Matteuzzi, Clara; Mauri, Andrea; Maurin, Brice; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; Meadows, Brian; Meier, Frank; Meissner, Marco; Melnychuk, Dmytro; Merk, Marcel; Michielin, Emanuele; Milanes, Diego Alejandro; Minard, Marie-Noelle; Mitzel, Dominik Stefan; Molina Rodriguez, Josue; Monroy, Ignacio Alberto; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Dominik; Müller, Janine; Müller, Katharina; Müller, Vanessa; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nandi, Anita; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Osorio Rodrigues, Bruno; Otalora Goicochea, Juan Martin; Otto, Adam; Owen, Patrick; Oyanguren, Maria Aranzazu; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Pappenheimer, Cheryl; Parker, William; Parkes, Christopher; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pearce, Alex; Pellegrino, Antonio; Penso, Gianni; Pepe Altarelli, Monica; Perazzini, Stefano; Perret, Pascal; Pescatore, Luca; Petridis, Konstantinos; Petrolini, Alessandro; Petruzzo, Marco; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pikies, Malgorzata; Pinci, Davide; Pistone, Alessandro; Piucci, Alessio; Playfer, Stephen; Plo Casasus, Maximo; Poikela, Tuomas; Polci, Francesco; Poluektov, Anton; Polyakov, Ivan; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Quagliani, Renato; Rachwal, Bartolomiej; Rademacker, Jonas; Rama, Matteo; Ramos Pernas, Miguel; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; dos Reis, Alberto; Renaudin, Victor; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Lopez, Jairo Alexis; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Ronayne, John William; Rotondo, Marcello; Ruf, Thomas; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santimaria, Marco; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrina, Darya; Schael, Stefan; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmelzer, Timon; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schubiger, Maxime; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Semennikov, Alexander; Sergi, Antonino; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Siddi, Benedetto Gianluca; Silva Coutinho, Rafael; Silva de Oliveira, Luiz Gustavo; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Edmund; Smith, Eluned; Smith, Iwan Thomas; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Stefkova, Slavomira; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szumlak, Tomasz; T'Jampens, Stephane; Tayduganov, Andrey; Tekampe, Tobias; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Todd, Jacob; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Trabelsi, Karim; Traill, Murdo; Tran, Minh Tâm; Tresch, Marco; Trisovic, Ana; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; van Veghel, Maarten; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Volkov, Vladimir; Vollhardt, Achim; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Weiden, Andreas; Whitehead, Mark; Wicht, Jean; Wilkinson, Guy; Wilkinson, Michael; Williams, Mark Richard James; Williams, Matthew; Williams, Mike; Williams, Timothy; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wraight, Kenneth; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xu, Zhirui; Yang, Zhenwei; Yu, Jiesheng; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zhukov, Valery; Zucchelli, Stefano

    2016-02-16

    An angular analysis of the $B^{0}\\rightarrow K^{*0}(\\rightarrow K^{+}\\pi^{-})\\mu^{+}\\mu^{-}$ decay is presented. The dataset corresponds to an integrated luminosity of $3.0\\,{\\mbox{fb}^{-1}}$ of $pp$ collision data collected at the LHCb experiment. The complete angular information from the decay is used to determine $C\\!P$-averaged observables and $C\\!P$ asymmetries, taking account of possible contamination from decays with the $K^{+}\\pi^{-}$ system in an S-wave configuration. The angular observables and their correlations are reported in bins of $q^2$, the invariant mass squared of the dimuon system. The observables are determined both from an unbinned maximum likelihood fit and by using the principal moments of the angular distribution. In addition, by fitting for $q^2$-dependent decay amplitudes in the region $1.1< q^{2} <6.0 \\mathrm{\\,Ge\\kern -0.1em V}^{2}/c^{4}$, the zero-crossing points of several angular observables are computed. A global fit is performed to the complete set of $C\\!P$-averaged ob...

  6. Applications of artificial neural networks for thermal analysis of heat exchangers - A review

    International Nuclear Information System (INIS)

    Mohanraj, M.; Jayaraj, S.; Muraleedharan, C.

    2015-01-01

    Artificial neural networks (ANN) have been widely used for thermal analysis of heat exchangers during the last two decades. In this paper, the applications of ANN for thermal analysis of heat exchangers are reviewed. The reported investigations on thermal analysis of heat exchangers are categorized into four major groups, namely (i) modeling of heat exchangers, (ii) estimation of heat exchanger parameters, (iii) estimation of phase change characteristics in heat exchangers and (iv) control of heat exchangers. Most of the papers related to the applications of ANN for thermal analysis of heat exchangers are discussed. The limitations of ANN for thermal analysis of heat exchangers and its further research needs in this field are highlighted. ANN is gaining popularity as a tool, which can be successfully used for the thermal analysis of heat exchangers with acceptable accuracy. (authors)

  7. Thermodynamic performance analysis and algorithm model of multi-pressure heat recovery steam generators (HRSG) based on heat exchangers layout

    International Nuclear Information System (INIS)

    Feng, Hongcui; Zhong, Wei; Wu, Yanling; Tong, Shuiguang

    2014-01-01

    Highlights: • A general model of multi-pressure HRSG based on heat exchangers layout is built. • The minimum temperature difference is introduced to replace pinch point analysis. • Effects of layout on dual pressure HRSG thermodynamic performances are analyzed. - Abstract: Changes of heat exchangers layout in heat recovery steam generator (HRSG) will modify the amount of waste heat recovered from flue gas; this brings forward a desire for the optimization of the design of HRSG. In this paper the model of multi-pressure HRSG is built, and an instance of a dual pressure HRSG under three different layouts of Taihu Boiler Co., Ltd. is discussed, with specified values of inlet temperature, mass flow rate, composition of flue gas and water/steam parameters as temperature, pressure etc., steam mass flow rate and heat efficiency of different heat exchangers layout of HRSG are analyzed. This analysis is based on the laws of thermodynamics and incorporated into the energy balance equations for the heat exchangers. In the conclusion, the results of the steam mass flow rate, heat efficiency obtained for three heat exchangers layout of HRSGs are compared. The results show that the optimization of heat exchangers layout of HRSGs has a great significance for waste heat recovery and energy conservation

  8. The Interactomic Analysis Reveals Pathogenic Protein Networks in Phomopsis longicolla Underlying Seed Decay of Soybean

    Directory of Open Access Journals (Sweden)

    Shuxian Li

    2018-04-01

    Full Text Available Phomopsis longicolla T. W. Hobbs (syn. Diaporthe longicolla is the primary cause of Phomopsis seed decay (PSD in soybean, Glycine max (L. Merrill. This disease results in poor seed quality and is one of the most economically important seed diseases in soybean. The objectives of this study were to infer protein–protein interactions (PPI and to identify conserved global networks and pathogenicity subnetworks in P. longicolla including orthologous pathways for cell signaling and pathogenesis. The interlog method used in the study identified 215,255 unique PPIs among 3,868 proteins. There were 1,414 pathogenicity related genes in P. longicolla identified using the pathogen host interaction (PHI database. Additionally, 149 plant cell wall degrading enzymes (PCWDE were detected. The network captured five different classes of carbohydrate degrading enzymes, including the auxiliary activities, carbohydrate esterases, glycoside hydrolases, glycosyl transferases, and carbohydrate binding molecules. From the PPI analysis, novel interacting partners were determined for each of the PCWDE classes. The most predominant class of PCWDE was a group of 60 glycoside hydrolases proteins. The glycoside hydrolase subnetwork was found to be interacting with 1,442 proteins within the network and was among the largest clusters. The orthologous proteins FUS3, HOG, CYP1, SGE1, and the g5566t.1 gene identified in this study could play an important role in pathogenicity. Therefore, the P. longicolla protein interactome (PiPhom generated in this study can lead to a better understanding of PPIs in soybean pathogens. Furthermore, the PPI may aid in targeting of genes and proteins for further studies of the pathogenicity mechanisms.

  9. Transient thermal performance analysis of micro heat pipes

    International Nuclear Information System (INIS)

    Liu, Xiangdong; Chen, Yongping

    2013-01-01

    A theoretical analysis of transient fluid flow and heat transfer in a triangular micro heat pipes (MHP) has been conducted to study the thermal response characteristics. By introducing the system identification theory, the quantitative evaluation of the MHP's transient thermal performance is realized. The results indicate that the evaporation and condensation processes are both extended into the adiabatic section. During the start-up process, the capillary radius along axial direction of MHP decreases drastically while the liquid velocity increases quickly at the early transient stage and an approximately linear decrease in wall temperature arises along the axial direction. The MHP behaves as a first-order LTI control system with the constant input power as the 'step input' and the evaporator wall temperature as the 'output'. Two corresponding evaluation criteria derived from the control theory, time constant and temperature constant, are able to quantitatively evaluate the thermal response speed and temperature level of MHP under start-up, which show that a larger triangular groove's hydraulic diameter within 0.18–0.42 mm is able to accelerate the start-up and decrease the start-up temperature level of MHP. Additionally, the MHP starts up fastest using the fluid of ethanol and most slowly using the working fluid of methanol, and the start-up temperature reaches maximum level for acetone and minimum level for the methanol. -- Highlights: • Transient thermal response of micro heat pipe is simulated by an improved model. • Control theory is introduced to quantify the thermal response of micro heat pipe. • Evaluation criteria are proposed to represent thermal response of micro heat pipe. • Effects of groove dimensions and working fluids on start-up of micro heat pipe are evaluated

  10. Global transcriptome analysis of the heat shock response ofshewanella oneidensis

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Haichun; Wang, Sarah; Liu, Xueduan; Yan, Tinfeng; Wu, Liyou; Alm, Eric; Arkin, Adam P.; Thompson, Dorothea K.; Zhou, Jizhong

    2004-04-30

    Shewanella oneidensis is an important model organism for bioremediation studies because of its diverse respiratory capabilities. However, the genetic basis and regulatory mechanisms underlying the ability of S. oneidensis to survive and adapt to various environmentally relevant stresses is poorly understood. To define this organism's molecular response to elevated growth temperatures, temporal gene expression profiles were examined in cells subjected to heat stress using whole-genome DNA microarrays for S. oneidensis MR-1. Approximately 15 percent (711) of the predicted S. oneidensis genes represented on the microarray were significantly up- or down-regulated (P < 0.05) over a 25-min period following shift to the heat shock temperature (42 C). As expected, the majority of S. oneidensis genes exhibiting homology to known chaperones and heat shock proteins (Hsps) were highly and transiently induced. In addition, a number of predicted genes encoding enzymes in glycolys is and the pentose cycle, [NiFe] dehydrogenase, serine proteases, transcriptional regulators (MerR, LysR, and TetR families), histidine kinases, and hypothetical proteins were induced in response to heat stress. Genes encoding membrane proteins were differentially expressed, suggesting that cells possibly alter their membrane composition or structure in response to variations in growth temperature. A substantial number of the genes encoding ribosomal proteins displayed down-regulated co-expression patterns in response to heat stress, as did genes encoding prophage and flagellar proteins. Finally, based on computational comparative analysis of the upstream promoter regions of S.oneidensis heat-inducible genes, a putative regulatory motif, showing high conservation to the Escherichia coli sigma 32-binding consensus sequence, was identified.

  11. Retrospective analysis of RF heating measurements of passive medical implants.

    Science.gov (United States)

    Song, Ting; Xu, Zhiheng; Iacono, Maria Ida; Angelone, Leonardo M; Rajan, Sunder

    2018-05-09

    The test reports for the RF-induced heating of metallic devices of hundreds of medical implants have been provided to the U.S. Food and Drug Administration as a part of premarket submissions. The main purpose of this study is to perform a retrospective analysis of the RF-induced heating data provided in the reports to analyze the trends and correlate them with implant geometric characteristics. The ASTM-based RF heating test reports from 86 premarket U.S. Food and Drug Administration submissions were reviewed by three U.S. Food and Drug Administration reviewers. From each test report, the dimensions and RF-induced heating values for a given whole-body (WB) specific absorption rate (SAR) and local background (LB) SAR were extracted and analyzed. The data from 56 stents were analyzed as a subset to further understand heating trends and length dependence. For a given WB SAR, the LB/WB SAR ratio varied significantly across the test labs, from 2.3 to 11.3. There was an increasing trend on the temperature change per LB SAR with device length. The maximum heating for stents occurred at lengths of approximately 100 mm at 3 T, and beyond 150 mm at 1.5 T. Differences in the LB/WB SAR ratios across testing labs and various MRI scanners could lead to inconsistent WB SAR labeling. Magnetic resonance (MR) conditional labeling based on WB SAR should be derived from a conservative estimate of global LB/WB ratios. Published 2018. This article is a U.S. Government work and is in the public domain in the USA.

  12. An analysis of the vapor flow and the heat conduction through the liquid-wick and pipe wall in a heat pipe with single or multiple heat sources

    Science.gov (United States)

    Chen, Ming-Ming; Faghri, Amir

    1990-01-01

    A numerical analysis is presented for the overall performance of heat pipes with single or multiple heat sources. The analysis includes the heat conduction in the wall and liquid-wick regions as well as the compressibility effect of the vapor inside the heat pipe. The two-dimensional elliptic governing equations in conjunction with the thermodynamic equilibrium relation and appropriate boundary conditions are solved numerically. The solutions are in agreement with existing experimental data for the vapor and wall temperatures at both low and high operating temperatures.

  13. Optimal observable analysis for the decay b → s plus missing energy

    Energy Technology Data Exchange (ETDEWEB)

    Calcuttawala, Zaineb; Kundu, Anirban [University of Calcutta, Department of Physics, Kolkata (India); Nandi, Soumitra; Patra, Sunando Kumar [Indian Institute of Technology, Department of Physics, Guwahati (India)

    2017-09-15

    The decay b → sν anti ν has received comparatively less attention than the semileptonic decay b → sl{sup +}l{sup -}, because neutrinos pass undetected and hence the process offers lesser number of observables. We show how the decay b → s + invisible(s) can shed light, even with a limited number of observables, on possible new physics beyond the Standard Model and also show, quantitatively, the reach of future B factories like SuperBelle to uncover such new physics. Depending on the operator structure of new physics, different channels may act as the best possible probe. We show, using the optimal observable technique, how almost the entire parameter space allowed till now can successfully be probed at a high-luminosity B factory. (orig.)

  14. Consistent Dalitz plot analysis of Cabibbo-favored D+ → K bar ππ+ decays

    Science.gov (United States)

    Niecknig, Franz; Kubis, Bastian

    2018-05-01

    We resume the study of the Cabibbo-favored charmed-meson decays D+ → K bar ππ+ in a dispersive framework that satisfies unitarity, analyticity, and crossing symmetry by construction. The formalism explicitly describes the strong final-state interactions between all three decay products and relies on pion-pion and pion-kaon phase shift input. For the first time, we show that the D+ →KSπ0π+ Dalitz plot obtained by the BESIII collaboration as well as the D+ →K-π+π+ Dalitz plot data by CLEO and FOCUS can be described consistently, exploiting the isospin relation between the two coupled decay channels that provides better constraints on the subtraction constants.

  15. Mathematical analysis of the Photovoltage Decay (PVD) method for minority carrier lifetime measurements

    Science.gov (United States)

    Vonroos, O. H.

    1982-01-01

    When the diffusion length of minority carriers becomes comparable with or larger than the thickness of a p-n junction solar cell, the characteristic decay of the photon-generated voltage results from a mixture of contributions with different time constants. The minority carrier recombination lifetime tau and the time constant l(2)/D, where l is essentially the thickness of the cell and D the minority carrier diffusion length, determine the signal as a function of time. It is shown that for ordinary solar cells (n(+)-p junctions), particularly when the diffusion length L of the minority carriers is larger than the cell thickness l, the excess carrier density decays according to exp (-t/tau-pi(2)Dt/4l(2)), tau being the lifetime. Therefore, tau can be readily determined by the photovoltage decay method once D and L are known.

  16. Comparative analysis of internal friction and natural frequency measured by free decay and forced vibration

    International Nuclear Information System (INIS)

    Wang, Y. Z.; Ding, X. D.; Xiong, X. M.; Zhang, J. X.

    2007-01-01

    Relations between various values of the internal friction (tgδ, Q -1 , Q -1* , and Λ/π) measured by free decay and forced vibration are analyzed systemically based on a fundamental mechanical model in this paper. Additionally, relations between various natural frequencies, such as vibration frequency of free decay ω FD , displacement-resonant frequency of forced vibration ω d , and velocity-resonant frequency of forced vibration ω 0 are calculated. Moreover, measurement of natural frequencies of a copper specimen of 99.9% purity has been made to demonstrate the relation between the measured natural frequencies of the system by forced vibration and free decay. These results are of importance for not only more accurate measurement of the elastic modulus of materials but also the data conversion between different internal friction measurements

  17. Analysis of the neutralino system in three-body leptonic decays of neutralinos

    International Nuclear Information System (INIS)

    Choi, S.Y.; Chung, B.C.; Kalinowski, J.; Rolbiecki, K.; Kim, Y.G.

    2006-01-01

    Neutralinos χ 0 in supersymmetric theories, the spin-1/2 Majorana-type superpartners of the U(1) and SU(2) neutral electroweak gauge bosons and SU(2) neutral Higgs bosons, are expected to be among light supersymmetric particles so that they can be produced copiously via direct pair production and/or from cascade decays of other sparticles such as sleptons at the planned Large Hadron Collider and the prospective International Linear Collider. Considering the prospects of having both highly polarized neutralinos and possibility of reconstructing their decay rest frames, we provide a systematic investigation of the three-body leptonic decays of the neutralinos in the minimal supersymmetric standard model and demonstrate alternative ways for probing the Majorana nature of the neutralinos and CP violation in the neutralino system. (orig.)

  18. Heating Performance Analysis of a Geothermal Heat Pump Working with Different Zeotropic and Azeotropic Mixtures

    OpenAIRE

    Robert Bedoić; Veljko Filipan

    2018-01-01

    The aim of the paper is to examine the possibility of application of the spreadsheet calculator and Reference Fluid Thermodynamic and Transport Properties database to a thermodynamic process. The heating process of a real soil-to-water heat pump, including heat transfer in the borehole heat exchanger has been analysed. How the changes of condensing temperature, at constant evaporating temperature, influence the following: heating capacity, compressor effective power, heat supplied to evaporat...