WorldWideScience

Sample records for maximum decay heat

  1. CRBRP decay heat removal systems

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  2. Decay heat 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.

  3. Development of limiting decay heat values

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  4. 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)

  5. 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%

  6. 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...

  7. 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

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  11. 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

  12. 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)

  13. 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

  14. Targeting the maximum heat recovery for systems with heat losses and heat gains

    International Nuclear Information System (INIS)

    Wan Alwi, Sharifah Rafidah; Lee, Carmen Kar Mun; Lee, Kim Yau; Abd Manan, Zainuddin; Fraser, Duncan M.

    2014-01-01

    Graphical abstract: Illustration of heat gains and losses from process streams. - Highlights: • Maximising energy savings through heat losses or gains. • Identifying location where insulation can be avoided. • Heuristics to maximise heat losses or gains. • Targeting heat losses or gains using the extended STEP technique and HEAT diagram. - Abstract: Process Integration using the Pinch Analysis technique has been widely used as a tool for the optimal design of heat exchanger networks (HENs). The Composite Curves and the Stream Temperature versus Enthalpy Plot (STEP) are among the graphical tools used to target the maximum heat recovery for a HEN. However, these tools assume that heat losses and heat gains are negligible. This work presents an approach that considers heat losses and heat gains during the establishment of the minimum utility targets. The STEP method, which is plotted based on the individual, as opposed to the composite streams, has been extended to consider the effect of heat losses and heat gains during stream matching. Several rules to guide the proper location of pipe insulation, and the appropriate procedure for stream shifting have been introduced in order to minimise the heat losses and maximise the heat gains. Application of the method on two case studies shows that considering heat losses and heat gains yield more realistic utility targets and help reduce both the insulation capital cost and utility cost of a HEN

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. Size dependence of efficiency at maximum power of heat engine

    KAUST Repository

    Izumida, Y.; Ito, N.

    2013-01-01

    We perform a molecular dynamics computer simulation of a heat engine model to study how the engine size difference affects its performance. Upon tactically increasing the size of the model anisotropically, we determine that there exists an optimum size at which the model attains the maximum power for the shortest working period. This optimum size locates between the ballistic heat transport region and the diffusive heat transport one. We also study the size dependence of the efficiency at the maximum power. Interestingly, we find that the efficiency at the maximum power around the optimum size attains a value that has been proposed as a universal upper bound, and it even begins to exceed the bound as the size further increases. We explain this behavior of the efficiency at maximum power by using a linear response theory for the heat engine operating under a finite working period, which naturally extends the low-dissipation Carnot cycle model [M. Esposito, R. Kawai, K. Lindenberg, C. Van den Broeck, Phys. Rev. Lett. 105, 150603 (2010)]. The theory also shows that the efficiency at the maximum power under an extreme condition may reach the Carnot efficiency in principle.© EDP Sciences Società Italiana di Fisica Springer-Verlag 2013.

  2. Size dependence of efficiency at maximum power of heat engine

    KAUST Repository

    Izumida, Y.

    2013-10-01

    We perform a molecular dynamics computer simulation of a heat engine model to study how the engine size difference affects its performance. Upon tactically increasing the size of the model anisotropically, we determine that there exists an optimum size at which the model attains the maximum power for the shortest working period. This optimum size locates between the ballistic heat transport region and the diffusive heat transport one. We also study the size dependence of the efficiency at the maximum power. Interestingly, we find that the efficiency at the maximum power around the optimum size attains a value that has been proposed as a universal upper bound, and it even begins to exceed the bound as the size further increases. We explain this behavior of the efficiency at maximum power by using a linear response theory for the heat engine operating under a finite working period, which naturally extends the low-dissipation Carnot cycle model [M. Esposito, R. Kawai, K. Lindenberg, C. Van den Broeck, Phys. Rev. Lett. 105, 150603 (2010)]. The theory also shows that the efficiency at the maximum power under an extreme condition may reach the Carnot efficiency in principle.© EDP Sciences Società Italiana di Fisica Springer-Verlag 2013.

  3. Maximum allowable heat flux for a submerged horizontal tube bundle

    International Nuclear Information System (INIS)

    McEligot, D.M.

    1995-01-01

    For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose quotes) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration

  4. Heat Convection at the Density Maximum Point of Water

    Science.gov (United States)

    Balta, Nuri; Korganci, Nuri

    2018-01-01

    Water exhibits a maximum in density at normal pressure at around 4° degree temperature. This paper demonstrates that during cooling, at around 4 °C, the temperature remains constant for a while because of heat exchange associated with convective currents inside the water. Superficial approach implies it as a new anomaly of water, but actually it…

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. Maximum-power-point tracking control of solar heating system

    KAUST Repository

    Huang, Bin-Juine

    2012-11-01

    The present study developed a maximum-power point tracking control (MPPT) technology for solar heating system to minimize the pumping power consumption at an optimal heat collection. The net solar energy gain Q net (=Q s-W p/η e) was experimentally found to be the cost function for MPPT with maximum point. The feedback tracking control system was developed to track the optimal Q net (denoted Q max). A tracking filter which was derived from the thermal analytical model of the solar heating system was used to determine the instantaneous tracking target Q max(t). The system transfer-function model of solar heating system was also derived experimentally using a step response test and used in the design of tracking feedback control system. The PI controller was designed for a tracking target Q max(t) with a quadratic time function. The MPPT control system was implemented using a microprocessor-based controller and the test results show good tracking performance with small tracking errors. It is seen that the average mass flow rate for the specific test periods in five different days is between 18.1 and 22.9kg/min with average pumping power between 77 and 140W, which is greatly reduced as compared to the standard flow rate at 31kg/min and pumping power 450W which is based on the flow rate 0.02kg/sm 2 defined in the ANSI/ASHRAE 93-1986 Standard and the total collector area 25.9m 2. The average net solar heat collected Q net is between 8.62 and 14.1kW depending on weather condition. The MPPT control of solar heating system has been verified to be able to minimize the pumping energy consumption with optimal solar heat collection. © 2012 Elsevier Ltd.

  12. Maximum heat flux in boiling in a large volume

    International Nuclear Information System (INIS)

    Bergmans, Dzh.

    1976-01-01

    Relationships are derived for the maximum heat flux qsub(max) without basing on the assumptions of both the critical vapor velocity corresponding to the zero growth rate, and planar interface. The Helmholz nonstability analysis of vapor column has been made to this end. The results of this examination have been used to find maximum heat flux for spherical, cylindric and flat plate heaters. The conventional hydrodynamic theory was found to be incapable of producing a satisfactory explanation of qsub(max) for small heaters. The occurrence of qsub(max) in the present case can be explained by inadequate removal of vapor output from the heater (the force of gravity for cylindrical heaters and surface tension for the spherical ones). In case of flat plate heater the qsub(max) value can be explained with the help of the hydrodynamic theory

  13. 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.)

  14. 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)

  15. 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

  16. Maximum skin hyperaemia induced by local heating: possible mechanisms.

    Science.gov (United States)

    Gooding, Kim M; Hannemann, Michael M; Tooke, John E; Clough, Geraldine F; Shore, Angela C

    2006-01-01

    Maximum skin hyperaemia (MH) induced by heating skin to > or = 42 degrees C is impaired in individuals at risk of diabetes and cardiovascular disease. Interpretation of these findings is hampered by the lack of clarity of the mechanisms involved in the attainment of MH. MH was achieved by local heating of skin to 42-43 degrees C for 30 min, and assessed by laser Doppler fluximetry. Using double-blind, randomized, placebo-controlled crossover study designs, the roles of prostaglandins were investigated by inhibiting their production with aspirin and histamine, with the H1 receptor antagonist cetirizine. The nitric oxide (NO) pathway was blocked by the NO synthase inhibitor, NG-nitro-L-arginine methyl esther (L-NAME), and enhanced by sildenafil (prevents breakdown of cGMP). MH was not altered by aspirin, cetirizine or sildenafil, but was reduced by L-NAME: median placebo 4.48 V (25th, 75th centiles: 3.71, 4.70) versus L-NAME 3.25 V (3.10, 3.80) (p = 0.008, Wilcoxon signed rank test). Inhibition of NO production (L-NAME) resulted in a more rapid reduction in hyperaemia after heating (p = 0.011), whereas hyperaemia was prolonged in the presence of sildenafil (p = 0.003). The increase in skin blood flow was largely confined to the directly heated area, suggesting that the role of heat-induced activation of the axon reflex was small. NO, but not prostaglandins, histamine or an axon reflex, contributes to the increase in blood flow on heating and NO is also a component of the resolution of MH after heating. Copyright 2006 S. Karger AG, Basel.

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

    Science.gov (United States)

    McDaniel, Patrick J.; Perkins, David R.

    1992-07-01

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

  18. 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)

  19. Maximum Langmuir Fields in Planetary Foreshocks Determined from the Electrostatic Decay Threshold

    Science.gov (United States)

    Robinson, P. A.; Cairns, Iver H.

    1995-01-01

    Maximum electric fields of Langmuir waves at planetary foreshocks are estimated from the threshold for electrostatic decay, assuming it saturates beam driven growth, and incorporating heliospheric variation of plasma density and temperature. Comparisons with spacecraft observations yields good quantitative agreement. Observations in type 3 radio sources are also in accord with this interpretation. A single mechanism can thus account for the highest fields of beam driven waves in both contexts.

  20. Evaluation of maximum radionuclide concentration from decay chains migration in aquifers

    International Nuclear Information System (INIS)

    Aquino Branco, O.E. de.

    1983-01-01

    The mathematical formulation of the mechanisms involved in the transport of contaminants in aquifers is presented. The methodology employed is described. A method of calculation the maximum concentration of radionuclides migrating in the underground water, and resulting from one decay chain, is then proposed. As an example, the methodology is applied to a waste basin, built to receive effluents from a hypothectical uranium ore mining and milling facility. (M.A.C.) [pt

  1. Passive decay heat removal by natural circulation

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  2. 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. 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

  4. 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)

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

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1983-01-01

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

  6. 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)

  7. Passive decay heat removal from the core region

    International Nuclear Information System (INIS)

    Hichen, E.F.; Jaegers, H.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Nayak, A.K.

    2015-01-01

    similar to prototypic material. About 50 kg of the molten material was poured into the test section at about 1100 deg C. Decay heat in the melt pool was simulated using 4 numbers of heater catridge, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness 2 and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behavior and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat. (author)

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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.)

  18. 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)

  19. 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)

  20. maximum conversion efficiency of thermionic heat to electricity

    African Journals Online (AJOL)

    DJFLEX

    Dushman constant ... Several attempts on the direct conversion of heat to electricity ... The net current density in the system is equal to jE – jC , which gets over the potential barrier. jE and jC are given by the Richardson-. Dushman equation as. │. ⌋.

  1. Maximum-power-point tracking control of solar heating system

    KAUST Repository

    Huang, Bin-Juine; Ton, Wei-Zhe; Wu, Chen-Chun; Ko, Hua-Wei; Chang, Hsien-Shun; Yen, Rue-Her; Wang, Jiunn-Cherng

    2012-01-01

    was used to determine the instantaneous tracking target Q max(t). The system transfer-function model of solar heating system was also derived experimentally using a step response test and used in the design of tracking feedback control system. The PI

  2. 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)

  3. 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)

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

    International Nuclear Information System (INIS)

    Davies, B.S.J.

    1978-07-01

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

  5. 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)

  6. 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

  7. 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

  8. The η{sub c} decays into light hadrons using the principle of maximum conformality

    Energy Technology Data Exchange (ETDEWEB)

    Du, Bo-Lun; Wu, Xing-Gang; Zeng, Jun; Bu, Shi; Shen, Jian-Ming [Chongqing University, Department of Physics, Chongqing (China)

    2018-01-15

    In the paper, we analyze the η{sub c} decays into light hadrons at the next-to-leading order QCD corrections by applying the principle of maximum conformality (PMC). The relativistic correction at the O(α{sub s}v{sup 2})-order level has been included in the discussion, which gives about 10% contribution to the ratio R. The PMC, which satisfies the renormalization group invariance, is designed to obtain a scale-fixed and scheme-independent prediction at any fixed order. To avoid the confusion of treating n{sub f}-terms, we transform the usual MS pQCD series into the one under the minimal momentum space subtraction scheme. To compare with the prediction under conventional scale setting, R{sub Conv,mMOM-r} = (4.12{sup +0.30}{sub -0.28}) x 10{sup 3}, after applying the PMC, we obtain R{sub PMC,mMOM-r} = (6.09{sup +0.62}{sub -0.55}) x 10{sup 3}, where the errors are squared averages of the ones caused by m{sub c} and Λ{sub mMOM}. The PMC prediction agrees with the recent PDG value within errors, i.e. R{sup exp} = (6.3 ± 0.5) x 10{sup 3}. Thus we think the mismatching of the prediction under conventional scale-setting with the data is due to improper choice of scale, which however can be solved by using the PMC. (orig.)

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

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

    1999-10-01

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

  10. 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)

  11. Maximum Power Output of Quantum Heat Engine with Energy Bath

    Directory of Open Access Journals (Sweden)

    Shengnan Liu

    2016-05-01

    Full Text Available The difference between quantum isoenergetic process and quantum isothermal process comes from the violation of the law of equipartition of energy in the quantum regime. To reveal an important physical meaning of this fact, here we study a special type of quantum heat engine consisting of three processes: isoenergetic, isothermal and adiabatic processes. Therefore, this engine works between the energy and heat baths. Combining two engines of this kind, it is possible to realize the quantum Carnot engine. Furthermore, considering finite velocity of change of the potential shape, here an infinite square well with moving walls, the power output of the engine is discussed. It is found that the efficiency and power output are both closely dependent on the initial and final states of the quantum isothermal process. The performance of the engine cycle is shown to be optimized by control of the occupation probability of the ground state, which is determined by the temperature and the potential width. The relation between the efficiency and power output is also discussed.

  12. 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

  13. 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.

  14. 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)

  15. Prediction of transient maximum heat flux based on a simple liquid layer evaporation model

    International Nuclear Information System (INIS)

    Serizawa, A.; Kataoka, I.

    1981-01-01

    A model of liquid layer evaporation with considerable supply of liquid has been formulated to predict burnout characteristics (maximum heat flux, life, etc.) during an increase of the power. The analytical description of the model is built upon the visual and photographic observations of the boiling configuration at near peak heat flux reported by other investigators. The prediction compares very favourably with water data presently available. It is suggested from the work reported here that the maximum heat flux occurs because of a balance between the consumption of the liquid film on the heated surface and the supply of liquid. Thickness of the liquid film is also very important. (author)

  16. 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.)

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. Performance analysis and comparison of an Atkinson cycle coupled to variable temperature heat reservoirs under maximum power and maximum power density conditions

    International Nuclear Information System (INIS)

    Wang, P.-Y.; Hou, S.-S.

    2005-01-01

    In this paper, performance analysis and comparison based on the maximum power and maximum power density conditions have been conducted for an Atkinson cycle coupled to variable temperature heat reservoirs. The Atkinson cycle is internally reversible but externally irreversible, since there is external irreversibility of heat transfer during the processes of constant volume heat addition and constant pressure heat rejection. This study is based purely on classical thermodynamic analysis methodology. It should be especially emphasized that all the results and conclusions are based on classical thermodynamics. The power density, defined as the ratio of power output to maximum specific volume in the cycle, is taken as the optimization objective because it considers the effects of engine size as related to investment cost. The results show that an engine design based on maximum power density with constant effectiveness of the hot and cold side heat exchangers or constant inlet temperature ratio of the heat reservoirs will have smaller size but higher efficiency, compression ratio, expansion ratio and maximum temperature than one based on maximum power. From the view points of engine size and thermal efficiency, an engine design based on maximum power density is better than one based on maximum power conditions. However, due to the higher compression ratio and maximum temperature in the cycle, an engine design based on maximum power density conditions requires tougher materials for engine construction than one based on maximum power conditions

  2. 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.)

  3. Getting maximum information from incomplete data on B → charmonium-KS decays

    International Nuclear Information System (INIS)

    Lipkin, H.J.

    1989-01-01

    Tests of CP violation using B decays into CP eigenstates can be improved by using events normally rejected because of incomplete information. A search for lepton asymmetry in decays Υ(4S) → B + bar B → (K S + J/ψ) + (lepton ± + X) can be improved by including other (c bar c)K S events where the (c bar c) pair is not bound in a J/ψ but in some other state like ψ' or η c and where the lepton asymmetry is predicted to be the same as for (K S + J/ψ), other (c bar c)K S events which are not fully reconstructed and (c bar c)K L events where the K L pair is not detected and which are predicted to have the opposite lepton asymmetry from corresponding K S events. The information from these additional events can give improved statistics if suitable cuts can be found to improve signal/noise. The opposite asymmetry predicted for K L events can test spurious lepton asymmetries due to systematic errors. 3 refs

  4. 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)

  5. 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

  6. 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.

  7. A thermoelectric generator using loop heat pipe and design match for maximum-power generation

    KAUST Repository

    Huang, Bin-Juine

    2015-09-05

    The present study focuses on the thermoelectric generator (TEG) using loop heat pipe (LHP) and design match for maximum-power generation. The TEG uses loop heat pipe, a passive cooling device, to dissipate heat without consuming power and free of noise. The experiments for a TEG with 4W rated power show that the LHP performs very well with overall thermal resistance 0.35 K W-1, from the cold side of TEG module to the ambient. The LHP is able to dissipate heat up to 110W and is maintenance free. The TEG design match for maximum-power generation, called “near maximum-power point operation (nMPPO)”, is studied to eliminate the MPPT (maximum-power point tracking controller). nMPPO is simply a system design which properly matches the output voltage of TEG with the battery. It is experimentally shown that TEG using design match for maximum-power generation (nMPPO) performs better than TEG with MPPT.

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

    established that a single plate core catcher can safely accommodate decay heat arising due to ∼70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5 h after the accident, demonstrating the benefit of large thermal inertia of the pool.

  10. Thermoelectric automotive waste heat energy recovery using maximum power point tracking

    International Nuclear Information System (INIS)

    Yu Chuang; Chau, K.T.

    2009-01-01

    This paper proposes and implements a thermoelectric waste heat energy recovery system for internal combustion engine automobiles, including gasoline vehicles and hybrid electric vehicles. The key is to directly convert the heat energy from automotive waste heat to electrical energy using a thermoelectric generator, which is then regulated by a DC-DC Cuk converter to charge a battery using maximum power point tracking. Hence, the electrical power stored in the battery can be maximized. Both analysis and experimental results demonstrate that the proposed system can work well under different working conditions, and is promising for automotive industry.

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  4. 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

  5. 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)

  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. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. Sensitivity analysis for maximum heat removal from debris in the lower head

    International Nuclear Information System (INIS)

    Kim, Yong Hoon; Suh, Kune Y.

    2000-01-01

    Sensitivity analyses were performed to determine the maximum heat removal capability from the debris and the reactor pressure vessel (RPV) wall through the gap that may be formed during a core melt relocation accident. Cases studied included four different nuclear power plant (TMI-2,KORI-2,YGN 3and4 and KNGR) per the thermal opower output. Results of the analysis show that the heat removal through gap cooling relative to flooding is efficacious as much as about 40% of the core material accumulated in the lower plenum in case of the TMI-2 reactor. In excess of 40%, however, the gap cooling alone was found not to be enough for heat removal from the core debris. There being uncertaainties aoboout the assumptions made in the present study,the analyses yield consistent results. If different cooling effects are considered, heat removal may be greatly enhanced. The LAVA experiements were performed at the Korea Atomic Energy Research Institute (KAERI) using al 2 O 3 /Fe thermite melt relocating down to the scaled vessel of a reactor lower head filled with preheated water. Test results indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices. If the cooling capacity of the intra-debris pores and crevices is comparable to debris-to-vessel heat removal capability, heat removal from the debris will be greatly augmented than heat removal by the gap cooling alone. The three nuclear reactor (KORI-2, YGN 3and4 and KNGR) calculation results for heat removal through the debris-to-vessel gap size of about 1mm were compared with the TMI-2 reactor calculation results for the case of gap cooling alone. (author)

  15. 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

  16. 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

  17. 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...

  18. 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)

  19. 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)

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  3. 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

  4. 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

  5. Relations between the efficiency, power and dissipation for linear irreversible heat engine at maximum trade-off figure of merit

    Science.gov (United States)

    Iyyappan, I.; Ponmurugan, M.

    2018-03-01

    A trade of figure of merit (\\dotΩ ) criterion accounts the best compromise between the useful input energy and the lost input energy of the heat devices. When the heat engine is working at maximum \\dotΩ criterion its efficiency increases significantly from the efficiency at maximum power. We derive the general relations between the power, efficiency at maximum \\dotΩ criterion and minimum dissipation for the linear irreversible heat engine. The efficiency at maximum \\dotΩ criterion has the lower bound \

  6. 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

  7. 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

  8. 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

  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. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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)

  2. 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)

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. Radioactivity and decay heat generation in precambrian magmatic rocks (with the South Pamirs as an example)

    International Nuclear Information System (INIS)

    Batyrmurzaev, A.S.; Alibekov, G.I.; Bekieva, A.A.

    2003-01-01

    The evaluation of the heat generation share in the results of the long-living radioactive elements (RAE) decay in the Earth surface layers is accomplished on the basis of the data on the uranium and thorium concentration in the precambrian magmatic rocks of the South Pamirs. It was supposed by the calculations, that the value of the heat flux, generated by the rocks, is determined mainly by the RAE content in the Earth upper layer crust itself of 10-15 km. It is shown that the radioheat generation share is within the range of 5-10% from the measured values of the geothermal flows [ru

  14. 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)

  15. 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.

  16. 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.

  17. 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

  18. 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

  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. 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. 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)

  2. 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.

  3. 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

  4. 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

  5. The maximum power condition of the brayton cycle with heat exchange processes

    International Nuclear Information System (INIS)

    Jung, Pyung Suk; Cha, Jin Girl; Ro, Sung Tack

    1985-01-01

    The ideal brayton cycle has been analyzed with the heat exchange processes between the working fluid and the heat source and the sink while their heat capacity rates are constant. The power of the cycle can be expressed in terms of a temperature of the cycle and the heat capacity rate of the working fluid. There exists an optimum power condition where the heat capacity rate of the working fluid has a value between those of the heat source and the heat sink, and the cycle efficiency is determined by the inlet temperatures of the heat source and the sink. (Author)

  6. 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

  7. '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

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 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.)

  13. Maximum Expected Wall Heat Flux and Maximum Pressure After Sudden Loss of Vacuum Insulation on the Stratospheric Observatory for Infrared Astronomy (SOFIA) Liquid Helium (LHe) Dewars

    Science.gov (United States)

    Ungar, Eugene K.

    2014-01-01

    The aircraft-based Stratospheric Observatory for Infrared Astronomy (SOFIA) is a platform for multiple infrared observation experiments. The experiments carry sensors cooled to liquid helium (LHe) temperatures. A question arose regarding the heat input and peak pressure that would result from a sudden loss of the dewar vacuum insulation. Owing to concerns about the adequacy of dewar pressure relief in the event of a sudden loss of the dewar vacuum insulation, the SOFIA Program engaged the NASA Engineering and Safety Center (NESC). This report summarizes and assesses the experiments that have been performed to measure the heat flux into LHe dewars following a sudden vacuum insulation failure, describes the physical limits of heat input to the dewar, and provides an NESC recommendation for the wall heat flux that should be used to assess the sudden loss of vacuum insulation case. This report also assesses the methodology used by the SOFIA Program to predict the maximum pressure that would occur following a loss of vacuum event.

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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.)

  2. 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.

  3. Simultaneous measurement of the maximum oscillation amplitude and the transient decay time constant of the QCM reveals stiffness changes of the adlayer.

    Science.gov (United States)

    Marxer, C Galli; Coen, M Collaud; Bissig, H; Greber, U F; Schlapbach, L

    2003-10-01

    Interpretation of adsorption kinetics measured with a quartz crystal microbalance (QCM) can be difficult for adlayers undergoing modification of their mechanical properties. We have studied the behavior of the oscillation amplitude, A(0), and the decay time constant, tau, of quartz during adsorption of proteins and cells, by use of a home-made QCM. We are able to measure simultaneously the frequency, f, the dissipation factor, D, the maximum amplitude, A(0), and the transient decay time constant, tau, every 300 ms in liquid, gaseous, or vacuum environments. This analysis enables adsorption and modification of liquid/mass properties to be distinguished. Moreover the surface coverage and the stiffness of the adlayer can be estimated. These improvements promise to increase the appeal of QCM methodology for any applications measuring intimate contact of a dynamic material with a solid surface.

  4. Use of MICRAS code on the evaluation of the maximum radionuclides concentrations due to transport/migration of decay chain in groundwaters

    International Nuclear Information System (INIS)

    Aquino Branco, O.E. de

    1995-01-01

    This paper presents a methodology for the evaluation of the maximum radionuclides concentrations in groundwaters, due to the transport/migration of decay chains. Analytical solution of the equations system is difficult, even if only three elements of the decay chain are considered. Therefore, a numerical solution is most convenient. An application of the MICRAS code, developed to assess maximum concentrations of each radionuclide, starting with the initial concentrations, is presented. The maximum concentration profile for 226 Ra, calculated using MICRAS, is compared with the results obtained through an analytical and a numerical model. The fitness of results is considered good. Simplified models, like the one represented by the application of MICRAS, are largely employed in the section in the selection and characterization of sites for radioactive wastes repositories and in studies of safety evaluation for the same purpose. A detailed analysis of the transport/migration of contaminants in aquifers requires a large quantify of data from the site and from the installation as well, which makes this analysis expensive and inviable during the preliminary phases of the studies. (author). 6 refs, 1 fig, 1 tab

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

    International Nuclear Information System (INIS)

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

    1994-09-01

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

  6. The effects of linear accelerations on the maximum heat transfer capacity of micro pipes with triangular grooves

    International Nuclear Information System (INIS)

    Shokouhmand, H.; Kahrobaian, A.; Tabandeh, N.; Jalilvand, A.

    2002-01-01

    Micro heat pipes are widely used for the thermal control of spacecraft and their electronic components. In this paper the influence of linear accelerations in micro grooves has been studied. A mathematical model for predicating the minimum meniscus radius and the maximum heat transport in triangular groove under the influence of linear acceleration is presented and method for determining the theoretical minimum meniscus radius is developed. It is shown that both, the direction and the magnitude of the acceleration have a great effect upon heat transfer capability of micro heat pipes. The analysis presented here provides a mechanism where by the groove geometry can be optimized with respect to the length of the heat pipe and direction and magnitude of linear acceleration

  7. 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

  8. The maximum and minimum values of the heat Q transmitted from metal to boiling water under atmospheric pressure

    International Nuclear Information System (INIS)

    Nukiyama, S.

    1991-01-01

    The quantity of heat transmitted from a metal surface to boiling water increases as the temperature difference ΔT is increased, but after the ΔT has reached a certain limit, quantity Q decreases with further increase in ΔT. This turning point is the maximum value of heat transmitted. The existence of this point was actually observed in the experiment. Under atmospheric pressure, ΔT corresponding to the maximum value of heat transfer for water at 100 degrees C falls between 20-40 degrees C, and Q is between 1,080,000 and 1,800,000 kcal/m 2 h (i.e. between 2,000 and 3,000 kg/m 2 h, if expressed in constant evaporation rate at 100 degrees C); this figure is larger than the maximum value of heat transfer as was previously considered. In this paper the minimum value of heat transfer was obtained, and in the Q-ΔT curve for the high temperature region, the burn-out effect is discussed

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

    International Nuclear Information System (INIS)

    Epiney, A.S.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Epiney, A. S.

    2010-09-01

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

  12. Development of a mathematical model of the heating phase of rubber mixture and development of the synthesis of the heating control algorithm using the Pontryagin maximum principle

    Directory of Open Access Journals (Sweden)

    V. S. Kudryashov

    2017-01-01

    Full Text Available The article is devoted to the development of the algorithm of the heating phase control of a rubber compound for CJSC “Voronezh tyre plant”. The algorithm is designed for implementation on basis of controller Siemens S-300 to control the RS-270 mixer. To compile the algorithm a systematic analysis of the heating process has been performed as a control object, also the mathematical model of the heating phase has been developed on the basis of the heat balance equation, which describes the process of heating of a heat-transfer agent in the heat exchanger and further heating of the mixture in the mixer. The dynamic characteristics of temperature of the heat exchanger and the rubber mixer have been obtained. Taking into account the complexity and nonlinearity of the control object – a rubber mixer, as well as the availability of methods and great experience in managing this machine in an industrial environment, the algorithm has been implemented using the Pontryagin maximum principle. The optimization problem is reduced to determining the optimal control (heating steam supply and the optimal path of the object’s output coordinates (the temperature of the mixture which ensure the least flow of steam while heating a rubber compound in a limited time. To do this, the mathematical model of the heating phase has been written in matrix form. Coefficients matrices for each state of the control, control and disturbance vectors have been created, the Hamilton function has been obtained and time switching points have been found for constructing an optimal control and escape path of the object. Analysis of the model experiments and practical research results in the process of programming of the controller have showed a decrease in the heating steam consumption by 24.4% during the heating phase of the rubber compound.

  13. 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.

  14. 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.

  15. 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.

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

    International Nuclear Information System (INIS)

    Tobias, A.

    1976-07-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  18. 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

  19. 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.

  20. 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)

  1. 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)

  2. 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)

  3. 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.

  4. The Greenland ice sheet - a model for its culmination and decay during and after the last glacial maximum

    DEFF Research Database (Denmark)

    Funder, Svend Visby; Hansen, Louise

    1996-01-01

    there was little change at all. The driving factor during this step was calving caused by rising sea level. This lasted until c. 10 ka, but may have been consumated before the Younger Dryas. The second step began with a glacier-readvance between 10 and 9.5 ka, and after this the fjord glaciers began to retreat....... Maximum Holocene uplift was attained in areas of the 10 ka ice margin, indicating that the uplift is essentially a response to the melting and unloading of ice that began at this time. In suppport of this, recent results in West, North and East Greenland indicate that the...

  5. 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)

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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).

  12. A thermoelectric generator using loop heat pipe and design match for maximum-power generation

    KAUST Repository

    Huang, Bin-Juine; Hsu, Po-Chien; Tsai, Rung-Je; Hussain, Muhammad Mustafa

    2015-01-01

    of noise. The experiments for a TEG with 4W rated power show that the LHP performs very well with overall thermal resistance 0.35 K W-1, from the cold side of TEG module to the ambient. The LHP is able to dissipate heat up to 110W and is maintenance free

  13. 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)

  14. Maximum Efficiency of Thermoelectric Heat Conversion in High-Temperature Power Devices

    Directory of Open Access Journals (Sweden)

    V. I. Khvesyuk

    2016-01-01

    Full Text Available Modern trends in development of aircraft engineering go with development of vehicles of the fifth generation. The features of aircrafts of the fifth generation are motivation to use new high-performance systems of onboard power supply. The operating temperature of the outer walls of engines is of 800–1000 K. This corresponds to radiation heat flux of 10 kW/m2 . The thermal energy including radiation of the engine wall may potentially be converted into electricity. The main objective of this paper is to analyze if it is possible to use a high efficiency thermoelectric conversion of heat into electricity. The paper considers issues such as working processes, choice of materials, and optimization of thermoelectric conversion. It presents the analysis results of operating conditions of thermoelectric generator (TEG used in advanced hightemperature power devices. A high-temperature heat source is a favorable factor for the thermoelectric conversion of heat. It is shown that for existing thermoelectric materials a theoretical conversion efficiency can reach the level of 15–20% at temperatures up to 1500 K and available values of Ioffe parameter being ZT = 2–3 (Z is figure of merit, T is temperature. To ensure temperature regime and high efficiency thermoelectric conversion simultaneously it is necessary to have a certain match between TEG power, temperature of hot and cold surfaces, and heat transfer coefficient of the cooling system. The paper discusses a concept of radiation absorber on the TEG hot surface. The analysis has demonstrated a number of potentialities for highly efficient conversion through using the TEG in high-temperature power devices. This work has been implemented under support of the Ministry of Education and Science of the Russian Federation; project No. 1145 (the programme “Organization of Research Engineering Activities”.

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. Optimal piston motion for maximum net output work of Daniel cam engines with low heat rejection

    International Nuclear Information System (INIS)

    Badescu, Viorel

    2015-01-01

    Highlights: • The piston motion of low heat rejection compression ignition engines is optimized. • A realistic model taking into account the cooling system is developed. • The optimized cam is smaller for cylinders without thermal insulation. • The optimized cam size depends on ignition moment and cooling process intensity. - Abstract: Compression ignition engines based on classical tapper-crank systems cannot provide optimal piston motion. Cam engines are more appropriate for this purpose. In this paper the piston motion of a Daniel cam engine is optimized. Piston acceleration is taken as a control. The objective is to maximize the net output work during the compression and power strokes. A major research effort has been allocated in the last two decades for the development of low heat rejection engines. A thermally insulated cylinder is considered and a realistic model taking into account the cooling system is developed. The sinusoidal approximation of piston motion in the classical tapper-crank system overestimates the engine efficiency. The exact description of the piston motion in tapper-crank system is used here as a reference. The radiation process has negligible effects during the optimization. The approach with no constraint on piston acceleration is a reasonable approximation. The net output work is much larger (by 12–13%) for the optimized system than for the classical tapper-crank system, for similar thickness of cylinder walls and thermal insulation. Low heat rejection measures are not of significant importance for optimized cam engines. The optimized cam is smaller for a cylinder without thermal insulation than for an insulated cylinder (by up to 8%, depending on the local polar radius). The auto-ignition moment is not a parameter of significant importance for optimized cam engines. However, for given cylinder wall and insulation materials there is an optimum auto-ignition moment which maximizes the net output work. The optimum auto

  5. 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)

  6. 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

  7. 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.

  8. Thermal Properties for the Thermal-Hydraulics Analyses of the BR2 Maximum Nominal Heat Flux

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Kim, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, G. L. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2015-02-01

    This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in 235U) to LEU (19.75% enriched in 235U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. Section 2 provides a summary of the thermal properties in the form of tables while the following sections and appendices present the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: i) aluminum, ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), iii) beryllium, and iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase’s volume fraction. Appendix B provides a revised methodology for determining the thermal conductivity as a function of burnup for HEU and LEU.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  10. 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

  11. 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

  12. 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

  13. Evolution and modulation of tropical heating from the last glacial maximum through the twenty-first century

    Energy Technology Data Exchange (ETDEWEB)

    Hoyos, Carlos D.; Webster, Peter J. [Georgia Institute of Technology, School of Earth and Atmospheric Sciences, Atlanta, GA (United States)

    2012-04-15

    Twentieth century observations show that during the last 50 years the sea-surface temperature (SST) of the tropical oceans has increased by {proportional_to}0.5 C and the area of SST >26.5 and 28 C (arbitrarily referred to as the oceanic warm pool: OWP) by 15 and 50% respectively in association with an increase in green house gas concentrations, with non-understood natural variability or a combination of both. Based on CMIP3 projections the OWP is projected to double during twenty-first century in a moderate CO{sub 2} forcing scenario (IPCC A1B scenario). However, during the observational period the area of positive atmospheric heating (referred to as the dynamic warm pool, DWP), has remained constant. The threshold SST (T{sub H}), which demarks the region of net heating and cooling, has increased from 26.6 C in the 1950s to 27.1 C in the last decade and it is projected to increase to {proportional_to}28.5 C by 2100. Based on climate model simulations, the area of the DWP is projected to remain constant during the twenty-first century. Analysis of the paleoclimate model intercomparison project (PMIP I and II) simulations for the Last Glacial maximum and the Mid-Holocene periods show a very similar behaviour, with a larger OWP in periods of elevated tropical SST, and an almost constant DWP associated with a varying T{sub H}. The constancy of the DWP area, despite shifts in the background SST, is shown to be the result of a near exact matching between increases in the integrated convective heating within the DWP and the integrated radiative cooling outside the DWP as SST changes. Although the area of the DWP remains constant, the total tropical atmospheric heating is a strong function of the SST. For example the net heating has increased by about 10% from 1950 to 2000 and it is projected to increase by a further 20% by 2100. Such changes must be compensated by a more vigorous atmospheric circulation, with growth in convective heating within the warm pool, and an

  14. 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...

  15. Studies in boiling heat transfer in two phase flow through tube arrays: nucleate boiling heat transfer coefficient and maximum heat flux as a function of velocity and quality of Freon-113

    International Nuclear Information System (INIS)

    Rahmani, R.

    1983-01-01

    The nucleate boiling heat-transfer coefficient and the maximum heat flux were studied experimentally as functions of velocity, quality and heater diameter for single-phase flow, and two-phase flow of Freon-113 (trichlorotrifluorethane). Results show: (1) peak heat flux: over 300 measured peak heat flux data from two 0.875-in. and four 0.625-in.-diameter heaters indicated that: (a) for pool boiling, single-phase and two-phase forced convection boiling the only parameter (among hysteresis, rate of power increase, aging, presence and proximity of unheated rods) that has a statistically significant effect on the peak heat flux is the velocity. (b) In the velocity range (0 0 position or the point of impact of the incident fluid) and the top (180 0 position) of the test element, respectively

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. Use of Maximum Likelihood-Mixed Models to select stable reference genes: a case of heat stress response in sheep

    Directory of Open Access Journals (Sweden)

    Salces Judit

    2011-08-01

    Full Text Available Abstract Background Reference genes with stable expression are required to normalize expression differences of target genes in qPCR experiments. Several procedures and companion software have been proposed to find the most stable genes. Model based procedures are attractive because they provide a solid statistical framework. NormFinder, a widely used software, uses a model based method. The pairwise comparison procedure implemented in GeNorm is a simpler procedure but one of the most extensively used. In the present work a statistical approach based in Maximum Likelihood estimation under mixed models was tested and compared with NormFinder and geNorm softwares. Sixteen candidate genes were tested in whole blood samples from control and heat stressed sheep. Results A model including gene and treatment as fixed effects, sample (animal, gene by treatment, gene by sample and treatment by sample interactions as random effects with heteroskedastic residual variance in gene by treatment levels was selected using goodness of fit and predictive ability criteria among a variety of models. Mean Square Error obtained under the selected model was used as indicator of gene expression stability. Genes top and bottom ranked by the three approaches were similar; however, notable differences for the best pair of genes selected for each method and the remaining genes of the rankings were shown. Differences among the expression values of normalized targets for each statistical approach were also found. Conclusions Optimal statistical properties of Maximum Likelihood estimation joined to mixed model flexibility allow for more accurate estimation of expression stability of genes under many different situations. Accurate selection of reference genes has a direct impact over the normalized expression values of a given target gene. This may be critical when the aim of the study is to compare expression rate differences among samples under different environmental

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. Study of 13Cr-4Ni-(Mo (F6NM Steel Grade Heat Treatment for Maximum Hardness Control in Industrial Heats

    Directory of Open Access Journals (Sweden)

    Massimo De Sanctis

    2017-09-01

    Full Text Available The standard NACE MR0175 (ISO 15156 requires a maximum hardness value of 23 HRC for 13Cr-4Ni-(Mo steel grade for sour service, requiring a double tempering heat treatment at temperature in the range 648–691 °C for the first tempering and 593–621 °C for the second tempering. Difficulties in limiting alloy hardness after the tempering of forged mechanical components (F6NM are often faced. Variables affecting the thermal behavior of 13Cr-4Ni-(Mo during single and double tempering treatments have been studied by means of transmission electron microscopy (TEM observations, X-ray diffraction measurements, dilatometry, and thermo-mechanical simulations. It has been found that relatively low Ac1 temperatures in this alloy induce the formation of austenite phase above 600 °C during tempering, and that the formed, reverted austenite tends to be unstable upon cooling, thus contributing to the increase of final hardness via transformation to virgin martensite. Therefore, it is necessary to increase the Ac1 temperature as much as possible to allow the tempering of martensite at the temperature range required by NACE without the detrimental formation of virgin martensite upon final cooling. Attempts to do so have been carried out by reducing both carbon (<0.02% C and nitrogen (<100 ppm levels. Results obtained herein show final hardness below NACE limits without an unacceptable loss of mechanical strength.

  11. 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)

  12. 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

  13. 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).

  14. 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)

  15. 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

  16. 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.

  17. Efficiency potentials of heat pumps with combined heat and power. For maximum reduction of CO2 emissions and for electricity generation from fossil fuels with CO2 reduction in Switzerland

    International Nuclear Information System (INIS)

    Rognon, F.

    2005-06-01

    This comprehensive report for the Swiss Federal Office of Energy (SFOE) takes a look at how the efficiency potential of heat pumps together with combined heat and power systems can help provide a maximum reduction of CO 2 emissions and provide electricity generation from fossil fuel in Switzerland together with reductions in CO 2 emissions. In Switzerland, approximately 80% of the low-temperature heat required for space-heating and for the heating-up of hot water is produced by burning combustibles. Around a million gas and oil boilers were in use in Switzerland in 2000, and these accounted for approximately half the country's 41.1 million tonnes of CO 2 emissions. The authors state that there is a more efficient solution with lower CO 2 emissions: the heat pump. With the enormous potential of our environment it would be possible to replace half the total number of boilers in use today with heat pumps. This would be equivalent to 90 PJ p.a. of useful heat, or 500,000 systems. The power source for heat pumps should come from the substitution of electric heating systems (electric resistor-based systems) and from the replacement of boilers. This should be done by using combined heat and power systems with full heat utilisation. This means, according to the authors, that the entire required power source can be provided without the need to construct new electricity production plants. The paper examines and discusses the theoretical, technical, market and realisable potentials

  18. 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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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).

  7. 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

  8. 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

  9. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  12. 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.

  13. 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.

  14. 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.

  15. Integrated solar thermal Brayton cycles with either one or two regenerative heat exchangers for maximum power output

    International Nuclear Information System (INIS)

    Jansen, E.; Bello-Ochende, T.; Meyer, J.P.

    2015-01-01

    The main objective of this paper is to optimise the open-air solar-thermal Brayton cycle by considering the implementation of the second law of thermodynamics and how it relates to the design of the heat exchanging components within it. These components included one or more regenerators (in the form of cross-flow heat exchangers) and the receiver of a parabolic dish concentrator where the system heat was absorbed. The generation of entropy was considered as it was associated with the destruction of exergy or available work. The dimensions of some components were used to optimise the cycles under investigation. EGM (Entropy Generation Minimisation) was employed to optimise the system parameters by considering their influence on the total generation of entropy (destruction of exergy). Various assumptions and constraints were considered and discussed. The total entropy generation rate and irreversibilities were determined by considering the individual components and ducts of the system, as well as their respective inlet and outlet conditions. The major system parameters were evaluated as functions of the mass flow rate to allow for a proper discussion of the system performance. The performances of both systems were investigated, and characteristics were listed for both. Finally, a comparison is made to shed light on the differences in performance. - Highlights: • Implementation of the second law of thermodynamics. • Design of heat exchanging and collecting equipment. • Utilisation of Entropy Generation Minimization. • Presentation of a multi-objective optimization. • Raise efficiency with more regeneration

  16. 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)

  17. 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.

  18. 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

  19. 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.

  20. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

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

  3. 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)

  4. 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)

  5. Modelling the occurrence of heat waves in maximum and minimum temperatures over Spain and projections for the period 2031-60

    Science.gov (United States)

    Abaurrea, J.; Asín, J.; Cebrián, A. C.

    2018-02-01

    The occurrence of extreme heat events in maximum and minimum daily temperatures is modelled using a non-homogeneous common Poisson shock process. It is applied to five Spanish locations, representative of the most common climates over the Iberian Peninsula. The model is based on an excess over threshold approach and distinguishes three types of extreme events: only in maximum temperature, only in minimum temperature and in both of them (simultaneous events). It takes into account the dependence between the occurrence of extreme events in both temperatures and its parameters are expressed as functions of time and temperature related covariates. The fitted models allow us to characterize the occurrence of extreme heat events and to compare their evolution in the different climates during the observed period. This model is also a useful tool for obtaining local projections of the occurrence rate of extreme heat events under climate change conditions, using the future downscaled temperature trajectories generated by Earth System Models. The projections for 2031-60 under scenarios RCP4.5, RCP6.0 and RCP8.5 are obtained and analysed using the trajectories from four earth system models which have successfully passed a preliminary control analysis. Different graphical tools and summary measures of the projected daily intensities are used to quantify the climate change on a local scale. A high increase in the occurrence of extreme heat events, mainly in July and August, is projected in all the locations, all types of event and in the three scenarios, although in 2051-60 the increase is higher under RCP8.5. However, relevant differences are found between the evolution in the different climates and the types of event, with a specially high increase in the simultaneous ones.

  6. Extreme warmth and heat-stressed plankton in the tropics during the Paleocene-Eocene Thermal Maximum.

    Science.gov (United States)

    Frieling, Joost; Gebhardt, Holger; Huber, Matthew; Adekeye, Olabisi A; Akande, Samuel O; Reichart, Gert-Jan; Middelburg, Jack J; Schouten, Stefan; Sluijs, Appy

    2017-03-01

    Global ocean temperatures rapidly warmed by ~5°C during the Paleocene-Eocene Thermal Maximum (PETM; ~56 million years ago). Extratropical sea surface temperatures (SSTs) met or exceeded modern subtropical values. With these warm extratropical temperatures, climate models predict tropical SSTs >35°C-near upper physiological temperature limits for many organisms. However, few data are available to test these projected extreme tropical temperatures or their potential lethality. We identify the PETM in a shallow marine sedimentary section deposited in Nigeria. On the basis of planktonic foraminiferal Mg/Ca and oxygen isotope ratios and the molecular proxy [Formula: see text], latest Paleocene equatorial SSTs were ~33°C, and [Formula: see text] indicates that SSTs rose to >36°C during the PETM. This confirms model predictions on the magnitude of polar amplification and refutes the tropical thermostat theory. We attribute a massive drop in dinoflagellate abundance and diversity at peak warmth to thermal stress, showing that the base of tropical food webs is vulnerable to rapid warming.

  7. Droplet condensation in rapidly decaying pressure fields

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. 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

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

    International Nuclear Information System (INIS)

    Nichols, A.L.; Nordborg, C.

    2009-02-01

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

  10. 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.

  11. Welding-induced local maximum residual stress in heat affected zone of low-carbon austenitic stainless steel with machined surface layer and its influential factors

    International Nuclear Information System (INIS)

    Okano, Shigetaka; Ihara, Ryohei; Kanamaru, Daisuke; Mochizuki, Masahito

    2015-01-01

    In this study, the effects of work-hardening and pre-existing stress in the machined surface layer of low-carbon austenitic stainless steel on the welding-induced residual stress were experimentally investigated through the use of weld specimens with three different surface layers; as-cutout, mechanically-polished and electrolytically-polished. The high tensile and compressive stresses exist in the work-hardened surface layer of the as-cutout and mechanically-polished specimens, respectively. Meanwhile, no stress and work-hardened surface layer exist in the electrolytically-polished specimen. TIG bead-on-plate welding under the same welding heat input conditions was performed to introduce the residual stress into these specimens. Using these welded specimens, the distributions of welding-induced residual stress were measured by the X-ray diffraction method. Similarly, the distributions of hardness in welds were estimated by the Vickers hardness test. And then, these distributions were compared with one another. Based on the results, the residual stress in the weld metal (WM) is completely unaffected by the machined surface layer because the work-hardened surface layer disappears through the processes of melting and solidification during welding. The local maximum longitudinal tensile residual stress in the heat affected zone (HAZ) depends on the work-hardening but not on the existing stress, regardless of whether tensile or compressive, in the machined surface layer before welding. At the base metal far from WM and HAZ, the residual stress is formed by the addition of the welding-induced residual stress to the pre-existing stress in the machined surface layer before welding. The features of the welding-induced residual stress in low-carbon austenitic stainless steel with the machined surface layer and their influential factors were thus clarified. (author)

  12. 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)

  13. 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

  14. 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

  15. MOnthly TEmperature DAtabase of Spain 1951-2010: MOTEDAS (2): The Correlation Decay Distance (CDD) and the spatial variability of maximum and minimum monthly temperature in Spain during (1981-2010).

    Science.gov (United States)

    Cortesi, Nicola; Peña-Angulo, Dhais; Simolo, Claudia; Stepanek, Peter; Brunetti, Michele; Gonzalez-Hidalgo, José Carlos

    2014-05-01

    One of the key point in the develop of the MOTEDAS dataset (see Poster 1 MOTEDAS) in the framework of the HIDROCAES Project (Impactos Hidrológicos del Calentamiento Global en España, Spanish Ministery of Research CGL2011-27574-C02-01) is the reference series for which no generalized metadata exist. In this poster we present an analysis of spatial variability of monthly minimum and maximum temperatures in the conterminous land of Spain (Iberian Peninsula, IP), by using the Correlation Decay Distance function (CDD), with the aim of evaluating, at sub-regional level, the optimal threshold distance between neighbouring stations for producing the set of reference series used in the quality control (see MOTEDAS Poster 1) and the reconstruction (see MOREDAS Poster 3). The CDD analysis for Tmax and Tmin was performed calculating a correlation matrix at monthly scale between 1981-2010 among monthly mean values of maximum (Tmax) and minimum (Tmin) temperature series (with at least 90% of data), free of anomalous data and homogenized (see MOTEDAS Poster 1), obtained from AEMEt archives (National Spanish Meteorological Agency). Monthly anomalies (difference between data and mean 1981-2010) were used to prevent the dominant effect of annual cycle in the CDD annual estimation. For each station, and time scale, the common variance r2 (using the square of Pearson's correlation coefficient) was calculated between all neighbouring temperature series and the relation between r2 and distance was modelled according to the following equation (1): Log (r2ij) = b*°dij (1) being Log(rij2) the common variance between target (i) and neighbouring series (j), dij the distance between them and b the slope of the ordinary least-squares linear regression model applied taking into account only the surrounding stations within a starting radius of 50 km and with a minimum of 5 stations required. Finally, monthly, seasonal and annual CDD values were interpolated using the Ordinary Kriging with a

  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. 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)

  18. 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

  19. Treponema pallidum 3-Phosphoglycerate Mutase Is a Heat-Labile Enzyme That May Limit the Maximum Growth Temperature for the Spirochete

    Science.gov (United States)

    Benoit, Stéphane; Posey, James E.; Chenoweth, Matthew R.; Gherardini, Frank C.

    2001-01-01

    In the causative agent of syphilis, Treponema pallidum, the gene encoding 3-phosphoglycerate mutase, gpm, is part of a six-gene operon (tro operon) that is regulated by the Mn-dependent repressor TroR. Since substrate-level phosphorylation via the Embden-Meyerhof pathway is the principal way to generate ATP in T. pallidum and Gpm is a key enzyme in this pathway, Mn could exert a regulatory effect on central metabolism in this bacterium. To study this, T. pallidum gpm was cloned, Gpm was purified from Escherichia coli, and antiserum against the recombinant protein was raised. Immunoblots indicated that Gpm was expressed in freshly extracted infective T. pallidum. Enzyme assays indicated that Gpm did not require Mn2+ while 2,3-diphosphoglycerate (DPG) was required for maximum activity. Consistent with these observations, Mn did not copurify with Gpm. The purified Gpm was stable for more than 4 h at 25°C, retained only 50% activity after incubation for 20 min at 34°C or 10 min at 37°C, and was completely inactive after 10 min at 42°C. The temperature effect was attenuated when 1 mM DPG was added to the assay mixture. The recombinant Gpm from pSLB2 complemented E. coli strain PL225 (gpm) and restored growth on minimal glucose medium in a temperature-dependent manner. Increasing the temperature of cultures of E. coli PL225 harboring pSLB2 from 34 to 42°C resulted in a 7- to 11-h period in which no growth occurred (compared to wild-type E. coli). These data suggest that biochemical properties of Gpm could be one contributing factor to the heat sensitivity of T. pallidum. PMID:11466272

  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. 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.

  2. 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)

  3. 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)

  4. 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.

  5. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  7. Tau decays

    International Nuclear Information System (INIS)

    Golutvin, A.

    1994-09-01

    The most recent experimental results of τ physics are reviewed. The covered topics include precision measurements of semihadronic τ decay and their impact on tau branching ratio budget, the current status of the tau consistency test, a determination of Michel parameters and τ neutrino helicity, and upper limits on lepton-number violating τ decays. (orig.)

  8. Decay tank

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Bang, In Cheol

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  11. B decays

    CERN Document Server

    Stone, Sheldon

    1992-01-01

    The study of b quarks has now reached a stage where it is useful to review what has been learned so far and also to look at the implications of future studies. The most important observations thus far - measurement of the "B" lifetime, B 0 - B 0 mixing, and the observation of b? u transitions, as well as more mundane results on hadronic and semileptonic transitions - are described in detail by experimentalists who have been closely involved with the measurements. Theoretical progress in understanding b quark decays, including the mechanisms of hadronic and semileptonic decays, are described. S

  12. B decays

    CERN Document Server

    Stone, Sheldon

    1994-01-01

    This book reviews the study of b quarks and also looks at the implications of future studies. The most important observations thus far - including measurement of the ""B"" lifetime and observations of b -> u transitions - as well as the more mundane results of hadronic and semileptonic transitions are described in detail by experimentalists who have been closely involved with the measurements. Theoretical progress in understanding b quark decays, including the mechanisms of hadronic and semileptonic decays, are described. Synthesizing the experimental and theoretical information, the authors d

  13. 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.

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

    International Nuclear Information System (INIS)

    Warnke, E.P.

    1990-02-01

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

  15. Proton decay theory

    International Nuclear Information System (INIS)

    Marciano, W.J.

    1983-01-01

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

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

    Science.gov (United States)

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

    2011-09-01

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

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

    International Nuclear Information System (INIS)

    Okano, Yasushi; Yamano, Hidemasa

    2015-01-01

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

  18. Monopole heat

    International Nuclear Information System (INIS)

    Turner, M.S.

    1983-01-01

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

  19. General Purpose Heat Source Simulator

    Science.gov (United States)

    Emrich, Bill

    2008-01-01

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

  20. Decay property of Timoshenko system in thermoelasticity

    KAUST Repository

    Said-Houari, Belkacem; Kasimov, Aslan R.

    2011-01-01

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

  1. Approximate maximum parsimony and ancestral maximum likelihood.

    Science.gov (United States)

    Alon, Noga; Chor, Benny; Pardi, Fabio; Rapoport, Anat

    2010-01-01

    We explore the maximum parsimony (MP) and ancestral maximum likelihood (AML) criteria in phylogenetic tree reconstruction. Both problems are NP-hard, so we seek approximate solutions. We formulate the two problems as Steiner tree problems under appropriate distances. The gist of our approach is the succinct characterization of Steiner trees for a small number of leaves for the two distances. This enables the use of known Steiner tree approximation algorithms. The approach leads to a 16/9 approximation ratio for AML and asymptotically to a 1.55 approximation ratio for MP.

  2. Rapid heating tensile tests of high-energy-rate-forged 316L stainless steel containing internal helium from radioactive decay of absorbed tritium

    International Nuclear Information System (INIS)

    Mosley, W.C.

    1990-01-01

    316L stainless steel is a candidate material for construction of equipment that will be exposed to tritium. This austenitic stainless steel is frequently used in the high-energy-rate-forged (HERF) metallurgical condition to take advantage of increased strength produced by cold work introduced by this process. Proper design of tritium-handling equipment will require an understanding of how helium-3, the product of radioactive decay of tritium, affects mechanical properties. This report describes results of elevated-temperature tensile testing of HERF 316L stainless steel specimens containing helium concentrations of 171 (calculated) atomic parts per million (appm). Results are compared with those reported previously for specimens containing 0 and 94 (measured) appm helium

  3. 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)

  4. Maximum permissible dose

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    This chapter presents a historic overview of the establishment of radiation guidelines by various national and international agencies. The use of maximum permissible dose and maximum permissible body burden limits to derive working standards is discussed

  5. Study of charmonium rare decays

    International Nuclear Information System (INIS)

    Brient, J.C.

    1986-09-01

    This thesis presents the study of rare decays of charmonium states formed in the interaction of an antiproton beam with an hydrogen gas jet target. Electromagnetic final states are used to sign the charmonium state formation (e + e - , e + e - + Χ, γγ). The selection of events used a two arms non magnetic spectrometer, with a charged track system, a threshold Cerenkov counter to tag the electron (positron), and an e.m. calorimeter. Energy scan technic have been used to observe the resonant formation through the excitation curves. Parameters of the states (mass, total and partial widths) are extracted from these curves using a statistical analysis. Two types of decays have been studied in this thesis: 1 P 1 charmonium state decay to the ψ (signed by its e + e - decay). In the energy scan around the center of gravity of the P charmonium states, we observe a cluster of 5 events, in a narrow mass range. This cluster correspond to a 2.7 σ signal. The most probable interpretation of this signal is given by a narrow resonance, with a mass of 3526. MeV. Due to the properties (mass, width and decay) of this signal, this could be interpreted as the 1 P 1 charmonium state. 2 photons decay of the η c and Χ 2 . 22 γγ events are observed, 15 in the η c region, and 7 in the Χ 2 region. This sample is interpreted as a direct observation of η c and Χ 2 decay into γγ. Parameters of these decays, (γγ partial width), are extracted using a maximum likekihood analysis. Theoretical models of charmonium explain correctly the properties of the charmonium, including the results presented in this thesis. 57 refs [fr

  6. Coupled ice sheet - climate simulations of the last glacial inception and last glacial maximum with a model of intermediate complexity that includes a dynamical downscaling of heat and moisture

    Science.gov (United States)

    Quiquet, Aurélien; Roche, Didier M.

    2017-04-01

    Comprehensive fully coupled ice sheet - climate models allowing for multi-millenia transient simulations are becoming available. They represent powerful tools to investigate ice sheet - climate interactions during the repeated retreats and advances of continental ice sheets of the Pleistocene. However, in such models, most of the time, the spatial resolution of the ice sheet model is one order of magnitude lower than the one of the atmospheric model. As such, orography-induced precipitation is only poorly represented. In this work, we briefly present the most recent improvements of the ice sheet - climate coupling within the model of intermediate complexity iLOVECLIM. On the one hand, from the native atmospheric resolution (T21), we have included a dynamical downscaling of heat and moisture at the ice sheet model resolution (40 km x 40 km). This downscaling accounts for feedbacks of sub-grid precipitation on large scale energy and water budgets. From the sub-grid atmospheric variables, we compute an ice sheet surface mass balance required by the ice sheet model. On the other hand, we also explicitly use oceanic temperatures to compute sub-shelf melting at a given depth. Based on palaeo evidences for rate of change of eustatic sea level, we discuss the capability of our new model to correctly simulate the last glacial inception ( 116 kaBP) and the ice volume of the last glacial maximum ( 21 kaBP). We show that the model performs well in certain areas (e.g. Canadian archipelago) but some model biases are consistent over time periods (e.g. Kara-Barents sector). We explore various model sensitivities (e.g. initial state, vegetation, albedo) and we discuss the importance of the downscaling of precipitation for ice nucleation over elevated area and for the surface mass balance of larger ice sheets.

  7. Parametric Decay during HHFW on NSTX

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  8. Irreversible and endoreversible behaviors of the LD-model for heat devices: the role of the time constraints and symmetries on the performance at maximum χ figure of merit

    Science.gov (United States)

    Gonzalez-Ayala, Julian; Calvo Hernández, A.; Roco, J. M. M.

    2016-07-01

    The main unified energetic properties of low dissipation heat engines and refrigerator engines allow for both endoreversible or irreversible configurations. This is accomplished by means of the constraints imposed on the characteristic global operation time or the contact times between the working system with the external heat baths and modulated by the dissipation symmetries. A suited unified figure of merit (which becomes power output for heat engines) is analyzed and the influence of the symmetries on the optimum performance discussed. The obtained results, independent on any heat transfer law, are faced with those obtained from Carnot-like heat models where specific heat transfer laws are needed. Thus, it is shown that only the inverse phenomenological law, often used in linear irreversible thermodynamics, correctly reproduces all optimized values for both the efficiency and coefficient of performance values.

  9. Maximum Acceleration Recording Circuit

    Science.gov (United States)

    Bozeman, Richard J., Jr.

    1995-01-01

    Coarsely digitized maximum levels recorded in blown fuses. Circuit feeds power to accelerometer and makes nonvolatile record of maximum level to which output of accelerometer rises during measurement interval. In comparison with inertia-type single-preset-trip-point mechanical maximum-acceleration-recording devices, circuit weighs less, occupies less space, and records accelerations within narrower bands of uncertainty. In comparison with prior electronic data-acquisition systems designed for same purpose, circuit simpler, less bulky, consumes less power, costs and analysis of data recorded in magnetic or electronic memory devices. Circuit used, for example, to record accelerations to which commodities subjected during transportation on trucks.

  10. Maximum Quantum Entropy Method

    OpenAIRE

    Sim, Jae-Hoon; Han, Myung Joon

    2018-01-01

    Maximum entropy method for analytic continuation is extended by introducing quantum relative entropy. This new method is formulated in terms of matrix-valued functions and therefore invariant under arbitrary unitary transformation of input matrix. As a result, the continuation of off-diagonal elements becomes straightforward. Without introducing any further ambiguity, the Bayesian probabilistic interpretation is maintained just as in the conventional maximum entropy method. The applications o...

  11. Maximum power demand cost

    International Nuclear Information System (INIS)

    Biondi, L.

    1998-01-01

    The charging for a service is a supplier's remuneration for the expenses incurred in providing it. There are currently two charges for electricity: consumption and maximum demand. While no problem arises about the former, the issue is more complicated for the latter and the analysis in this article tends to show that the annual charge for maximum demand arbitrarily discriminates among consumer groups, to the disadvantage of some [it

  12. Thermal decay of Lennard-Jones clusters

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  13. Bubble chamber: Omega production and decay

    CERN Document Server

    1973-01-01

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

  14. Measurements of Critical Heat Flux using Mass Transfer System

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seung Hyun; Chung Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    In a severe accident, the reactor vessel is heated by the decay heat from core melts and the outer surface of reactor vessel is cooled by the natural convection of water pool. When the heat flux increases, boiling will start. Further increase of the heat flux may result in the CHF, which is generated by the bubble combinations. The CHF means that the reactor vessel was separated with coolant and wall temperature is raised rapidly. It may damage the reactor vessel. Also the CHF indicates the maximum cooling capability of the system. Therefore, the CHF has been used as a criterion for the regulatory and licensing. Mechanism of hydrogen vapor bubbles generated and combined can be simulated water bubbles mechanism. And also the both heat and mass transfer mechanism of CHF can be identified in the same methods. Therefore, the CHF phenomena can be simulated enough by mass transfer.

  15. Bubble chamber: Omega production and decay

    CERN Multimedia

    1973-01-01

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

  16. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  17. 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

  18. IMPACT OF HEATING AND AIR CONDITIONING SYSTEM OPERATION AND LEAKAGE ON VENTILATION AND INTERCOMPARTMENT TRANSPORT: STUDIES IN UNOCCUPIED AND OCCUPIED TENNESSEE VALLEY HOMES

    Science.gov (United States)

    Forced-air heating and air conditioning (HAC) systems caused an average and maximum increase in air infiltration rates of 1.8- and 4.3-fold, respectively, during brief whole-house studies of tracer gas decay In 39 occupied houses. An average Increase in air infiltration rate of 0...

  19. Maximum likely scale estimation

    DEFF Research Database (Denmark)

    Loog, Marco; Pedersen, Kim Steenstrup; Markussen, Bo

    2005-01-01

    A maximum likelihood local scale estimation principle is presented. An actual implementation of the estimation principle uses second order moments of multiple measurements at a fixed location in the image. These measurements consist of Gaussian derivatives possibly taken at several scales and/or ...

  20. Robust Maximum Association Estimators

    NARCIS (Netherlands)

    A. Alfons (Andreas); C. Croux (Christophe); P. Filzmoser (Peter)

    2017-01-01

    textabstractThe maximum association between two multivariate variables X and Y is defined as the maximal value that a bivariate association measure between one-dimensional projections αX and αY can attain. Taking the Pearson correlation as projection index results in the first canonical correlation

  1. Number of detectable kaon decays at LAMPF II

    International Nuclear Information System (INIS)

    Sanford, T.W.L.

    1982-04-01

    The maximum number of kaon decays detectable at LAMPF II is estimated for both in-flight and stopping decays. Under reasonable assumptions, the momentum of the kaon beam that optimizes the decay yield occurs at about 6 GeV/c and 600 MeV/c for in-flight and stopping decays, respectively. K + decay yields are fo the order of 7 x 10 7 per 10 14 interacting with K - yields being typically 5 times less. By measuring decays from such beams, a statistical limit of 10 -15 on a branching ratio to a particular channel can be placed in a 100-day run. The large number of kaon decays available at LAMPF II thus provides a powerful tool for sensitively examining rare-decay processes of the kaon

  2. Dental Caries (Tooth Decay)

    Science.gov (United States)

    ... Materials Contact Us Home Research Data & Statistics Dental Caries (Tooth Decay) Dental caries (tooth decay) remains the most prevalent chronic disease ... adults, even though it is largely preventable. Although caries has significantly decreased for most Americans over the ...

  3. Dental Caries (Tooth Decay)

    Science.gov (United States)

    ... Contact Us Home Research Data & Statistics Share Dental Caries (Tooth Decay) Dental caries (tooth decay) remains the most prevalent chronic disease ... adults, even though it is largely preventable. Although caries has significantly decreased for most Americans over the ...

  4. Maximum power point tracking

    International Nuclear Information System (INIS)

    Enslin, J.H.R.

    1990-01-01

    A well engineered renewable remote energy system, utilizing the principal of Maximum Power Point Tracking can be m ore cost effective, has a higher reliability and can improve the quality of life in remote areas. This paper reports that a high-efficient power electronic converter, for converting the output voltage of a solar panel, or wind generator, to the required DC battery bus voltage has been realized. The converter is controlled to track the maximum power point of the input source under varying input and output parameters. Maximum power point tracking for relative small systems is achieved by maximization of the output current in a battery charging regulator, using an optimized hill-climbing, inexpensive microprocessor based algorithm. Through practical field measurements it is shown that a minimum input source saving of 15% on 3-5 kWh/day systems can easily be achieved. A total cost saving of at least 10-15% on the capital cost of these systems are achievable for relative small rating Remote Area Power Supply systems. The advantages at larger temperature variations and larger power rated systems are much higher. Other advantages include optimal sizing and system monitor and control

  5. MODEL RADIOACTIVE RADON DECAY

    Directory of Open Access Journals (Sweden)

    R.I. Parovik

    2012-06-01

    Full Text Available In a model of radioactive decay of radon in the sample (222Rn. The model assumes that the probability of the decay of radon and its half-life depends on the fractal properties of the geological environment. The dependencies of the decay parameters of the fractal dimension of the medium.

  6. Maximum entropy methods

    International Nuclear Information System (INIS)

    Ponman, T.J.

    1984-01-01

    For some years now two different expressions have been in use for maximum entropy image restoration and there has been some controversy over which one is appropriate for a given problem. Here two further entropies are presented and it is argued that there is no single correct algorithm. The properties of the four different methods are compared using simple 1D simulations with a view to showing how they can be used together to gain as much information as possible about the original object. (orig.)

  7. Heat transfer in a fuel pin shipping container

    International Nuclear Information System (INIS)

    Ingham, J.G.

    1980-01-01

    Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800 0 F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures

  8. Utilising heat from nuclear waste for space heating

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

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

  9. The last glacial maximum

    Science.gov (United States)

    Clark, P.U.; Dyke, A.S.; Shakun, J.D.; Carlson, A.E.; Clark, J.; Wohlfarth, B.; Mitrovica, J.X.; Hostetler, S.W.; McCabe, A.M.

    2009-01-01

    We used 5704 14C, 10Be, and 3He ages that span the interval from 10,000 to 50,000 years ago (10 to 50 ka) to constrain the timing of the Last Glacial Maximum (LGM) in terms of global ice-sheet and mountain-glacier extent. Growth of the ice sheets to their maximum positions occurred between 33.0 and 26.5 ka in response to climate forcing from decreases in northern summer insolation, tropical Pacific sea surface temperatures, and atmospheric CO2. Nearly all ice sheets were at their LGM positions from 26.5 ka to 19 to 20 ka, corresponding to minima in these forcings. The onset of Northern Hemisphere deglaciation 19 to 20 ka was induced by an increase in northern summer insolation, providing the source for an abrupt rise in sea level. The onset of deglaciation of the West Antarctic Ice Sheet occurred between 14 and 15 ka, consistent with evidence that this was the primary source for an abrupt rise in sea level ???14.5 ka.

  10. Maximum Entropy Fundamentals

    Directory of Open Access Journals (Sweden)

    F. Topsøe

    2001-09-01

    Full Text Available Abstract: In its modern formulation, the Maximum Entropy Principle was promoted by E.T. Jaynes, starting in the mid-fifties. The principle dictates that one should look for a distribution, consistent with available information, which maximizes the entropy. However, this principle focuses only on distributions and it appears advantageous to bring information theoretical thinking more prominently into play by also focusing on the "observer" and on coding. This view was brought forward by the second named author in the late seventies and is the view we will follow-up on here. It leads to the consideration of a certain game, the Code Length Game and, via standard game theoretical thinking, to a principle of Game Theoretical Equilibrium. This principle is more basic than the Maximum Entropy Principle in the sense that the search for one type of optimal strategies in the Code Length Game translates directly into the search for distributions with maximum entropy. In the present paper we offer a self-contained and comprehensive treatment of fundamentals of both principles mentioned, based on a study of the Code Length Game. Though new concepts and results are presented, the reading should be instructional and accessible to a rather wide audience, at least if certain mathematical details are left aside at a rst reading. The most frequently studied instance of entropy maximization pertains to the Mean Energy Model which involves a moment constraint related to a given function, here taken to represent "energy". This type of application is very well known from the literature with hundreds of applications pertaining to several different elds and will also here serve as important illustration of the theory. But our approach reaches further, especially regarding the study of continuity properties of the entropy function, and this leads to new results which allow a discussion of models with so-called entropy loss. These results have tempted us to speculate over

  11. Decay property of Timoshenko system in thermoelasticity

    KAUST Repository

    Said-Houari, Belkacem

    2011-12-30

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

  12. Probable maximum flood control

    International Nuclear Information System (INIS)

    DeGabriele, C.E.; Wu, C.L.

    1991-11-01

    This study proposes preliminary design concepts to protect the waste-handling facilities and all shaft and ramp entries to the underground from the probable maximum flood (PMF) in the current design configuration for the proposed Nevada Nuclear Waste Storage Investigation (NNWSI) repository protection provisions were furnished by the United States Bureau of Reclamation (USSR) or developed from USSR data. Proposed flood protection provisions include site grading, drainage channels, and diversion dikes. Figures are provided to show these proposed flood protection provisions at each area investigated. These areas are the central surface facilities (including the waste-handling building and waste treatment building), tuff ramp portal, waste ramp portal, men-and-materials shaft, emplacement exhaust shaft, and exploratory shafts facility

  13. Introduction to maximum entropy

    International Nuclear Information System (INIS)

    Sivia, D.S.

    1988-01-01

    The maximum entropy (MaxEnt) principle has been successfully used in image reconstruction in a wide variety of fields. We review the need for such methods in data analysis and show, by use of a very simple example, why MaxEnt is to be preferred over other regularizing functions. This leads to a more general interpretation of the MaxEnt method, and its use is illustrated with several different examples. Practical difficulties with non-linear problems still remain, this being highlighted by the notorious phase problem in crystallography. We conclude with an example from neutron scattering, using data from a filter difference spectrometer to contrast MaxEnt with a conventional deconvolution. 12 refs., 8 figs., 1 tab

  14. Solar maximum observatory

    International Nuclear Information System (INIS)

    Rust, D.M.

    1984-01-01

    The successful retrieval and repair of the Solar Maximum Mission (SMM) satellite by Shuttle astronauts in April 1984 permitted continuance of solar flare observations that began in 1980. The SMM carries a soft X ray polychromator, gamma ray, UV and hard X ray imaging spectrometers, a coronagraph/polarimeter and particle counters. The data gathered thus far indicated that electrical potentials of 25 MeV develop in flares within 2 sec of onset. X ray data show that flares are composed of compressed magnetic loops that have come too close together. Other data have been taken on mass ejection, impacts of electron beams and conduction fronts with the chromosphere and changes in the solar radiant flux due to sunspots. 13 references

  15. Introduction to maximum entropy

    International Nuclear Information System (INIS)

    Sivia, D.S.

    1989-01-01

    The maximum entropy (MaxEnt) principle has been successfully used in image reconstruction in a wide variety of fields. The author reviews the need for such methods in data analysis and shows, by use of a very simple example, why MaxEnt is to be preferred over other regularizing functions. This leads to a more general interpretation of the MaxEnt method, and its use is illustrated with several different examples. Practical difficulties with non-linear problems still remain, this being highlighted by the notorious phase problem in crystallography. He concludes with an example from neutron scattering, using data from a filter difference spectrometer to contrast MaxEnt with a conventional deconvolution. 12 refs., 8 figs., 1 tab

  16. Functional Maximum Autocorrelation Factors

    DEFF Research Database (Denmark)

    Larsen, Rasmus; Nielsen, Allan Aasbjerg

    2005-01-01

    MAF outperforms the functional PCA in concentrating the interesting' spectra/shape variation in one end of the eigenvalue spectrum and allows for easier interpretation of effects. Conclusions. Functional MAF analysis is a useful methods for extracting low dimensional models of temporally or spatially......Purpose. We aim at data where samples of an underlying function are observed in a spatial or temporal layout. Examples of underlying functions are reflectance spectra and biological shapes. We apply functional models based on smoothing splines and generalize the functional PCA in......\\verb+~+\\$\\backslash\\$cite{ramsay97} to functional maximum autocorrelation factors (MAF)\\verb+~+\\$\\backslash\\$cite{switzer85,larsen2001d}. We apply the method to biological shapes as well as reflectance spectra. {\\$\\backslash\\$bf Methods}. MAF seeks linear combination of the original variables that maximize autocorrelation between...

  17. Regularized maximum correntropy machine

    KAUST Repository

    Wang, Jim Jing-Yan; Wang, Yunji; Jing, Bing-Yi; Gao, Xin

    2015-01-01

    In this paper we investigate the usage of regularized correntropy framework for learning of classifiers from noisy labels. The class label predictors learned by minimizing transitional loss functions are sensitive to the noisy and outlying labels of training samples, because the transitional loss functions are equally applied to all the samples. To solve this problem, we propose to learn the class label predictors by maximizing the correntropy between the predicted labels and the true labels of the training samples, under the regularized Maximum Correntropy Criteria (MCC) framework. Moreover, we regularize the predictor parameter to control the complexity of the predictor. The learning problem is formulated by an objective function considering the parameter regularization and MCC simultaneously. By optimizing the objective function alternately, we develop a novel predictor learning algorithm. The experiments on two challenging pattern classification tasks show that it significantly outperforms the machines with transitional loss functions.

  18. Regularized maximum correntropy machine

    KAUST Repository

    Wang, Jim Jing-Yan

    2015-02-12

    In this paper we investigate the usage of regularized correntropy framework for learning of classifiers from noisy labels. The class label predictors learned by minimizing transitional loss functions are sensitive to the noisy and outlying labels of training samples, because the transitional loss functions are equally applied to all the samples. To solve this problem, we propose to learn the class label predictors by maximizing the correntropy between the predicted labels and the true labels of the training samples, under the regularized Maximum Correntropy Criteria (MCC) framework. Moreover, we regularize the predictor parameter to control the complexity of the predictor. The learning problem is formulated by an objective function considering the parameter regularization and MCC simultaneously. By optimizing the objective function alternately, we develop a novel predictor learning algorithm. The experiments on two challenging pattern classification tasks show that it significantly outperforms the machines with transitional loss functions.

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

    KAUST Repository

    Said-Houari, Belkacem

    2013-01-01

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

  20. SYMPOSIUM: Rare decays

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1989-04-15

    Late last year, a symposium entitled 'Rare Decays' attracted 115 participants to a hotel in Vancouver, Canada. These participants were particle physicists interested in checking conventional selection rules to look for clues of possible new behaviour outside today's accepted 'Standard Model'. For physicists, 'rare decays' include processes that have so far not been seen, explicitly forbidden by the rules of the Standard Model, or processes highly suppressed because the decay is dominated by an easier route, or includes processes resulting from multiple transitions.

  1. Effective Majorana neutrino decay

    Energy Technology Data Exchange (ETDEWEB)

    Duarte, Lucia [Instituto de Fisica, Facultad de Ingenieria,Universidad de la Republica, Montevideo (Uruguay); Romero, Ismael; Peressutti, Javier; Sampayo, Oscar A. [Universidad Nacional de Mar del Plata, Departamento de Fisica, Instituto de Investigaciones Fisicas de Mar del Plata (IFIMAR) CONICET, UNMDP, Mar del Plata (Argentina)

    2016-08-15

    We study the decay of heavy sterile Majorana neutrinos according to the interactions obtained from an effective general theory. We describe the two- and three-body decays for a wide range of neutrino masses. The results obtained and presented in this work could be useful for the study of the production and detection of these particles in a variety of high energy physics experiments and astrophysical observations. We show in different figures the dominant branching ratios and the total decay width. (orig.)

  2. Axigluon decays of toponium

    International Nuclear Information System (INIS)

    Faustov, R.N.; Vasilevskaya, I.G.

    1990-01-01

    Chiral-colour model predicts the existence of axigluons which is an octet of massive axial-vector gauge bosons. In this respect toponium decays into axigluons and gluons are of interest. The following toponium decays are considered: θ → Ag, θ → AAg, θ → ggg → AAg. The width of toponium S-state decays is calculated under various possible values of axigluon mass

  3. Decay of 143La

    International Nuclear Information System (INIS)

    Blachot, J.; Dousson, S.; Monnand, E.; Schussler, F.

    1976-01-01

    The decay of 143 La has been investigated. Sources have been obtained from 2 isotope separators (ISERE, OSIRIS). 12 gamma rays, with the most intense at 620keV representing only 1.4% of decay, have been attributed to the 143 La decay. A level scheme has been found and compared with the one deduced from (d,p) and (n,γ) reactions on 142 Ce [fr

  4. Intermittent heating of buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kohonen, K

    1983-02-01

    Conditions for intermittent heating of buildings are considered both theoretically and experimentally. Thermal behaviour of buildings adn rooms in intermittent heating is simulated by a program based on the convective heat balance equation and by simplified RC-models. The preheat times and the heating energy savings compared with continuous heating are presented for typical lightweight, mediumweight and heavyweight classroom and office modules. Formulaes for estimating the oversizing of the radiator network, the maximum heat output of heat exchangers in district heating and the efficiency of heating boilers in intermittent heating are presented. The preheat times and heating energy savings with different heating control systems are determined also experimentally in eight existing buildings. In addition some principles for the planning and application of intermittent heating systems are suggested.

  5. Solar maximum mission

    International Nuclear Information System (INIS)

    Ryan, J.

    1981-01-01

    By understanding the sun, astrophysicists hope to expand this knowledge to understanding other stars. To study the sun, NASA launched a satellite on February 14, 1980. The project is named the Solar Maximum Mission (SMM). The satellite conducted detailed observations of the sun in collaboration with other satellites and ground-based optical and radio observations until its failure 10 months into the mission. The main objective of the SMM was to investigate one aspect of solar activity: solar flares. A brief description of the flare mechanism is given. The SMM satellite was valuable in providing information on where and how a solar flare occurs. A sequence of photographs of a solar flare taken from SMM satellite shows how a solar flare develops in a particular layer of the solar atmosphere. Two flares especially suitable for detailed observations by a joint effort occurred on April 30 and May 21 of 1980. These flares and observations of the flares are discussed. Also discussed are significant discoveries made by individual experiments

  6. Decays of the tau lepton

    International Nuclear Information System (INIS)

    Burchat, P.R.

    1986-02-01

    Previous measurements of the branching fractions of the tau lepton result in a discrepancy between the inclusive branching fraction and the sum of the exclusive branching fractions to final states containing one charged particle. The sum of the exclusive branching fractions is significantly smaller than the inclusive branching fraction. In this analysis, the branching fractions for all the major decay modes are measured simultaneously with the sum of the branching fractions constrained to be one. The branching fractions are measured using an unbiased sample of tau decays, with little background, selected from 207 pb -1 of data accumulated with the Mark II detector at the PEP e + e - storage ring. The sample is selected using the decay products of one member of the γ + γ - pair produced in e + e - annihilation to identify the event and then including the opposite member of the pair in the sample. The sample is divided into subgroups according to charged and neutral particle multiplicity, and charged particle identification. The branching fractions are simultaneously measured using an unfold technique and a maximum likelihood fit. The results of this analysis indicate that the discrepancy found in previous experiments is possibly due to two sources. First, the leptonic branching fractions measured in this analysis are about one standard deviation higher than the world average. The measured leptonic branching fractions correspond to a tau lifetime of (3.0 +- 0.2) x 10 -13 s. Secondly, the total branching fraction to one charged hadron plus at least one neutral particle is measured to be (7 +- 3)% higher than the branching fraction expected from a combination of previous measurements and theoretical predictions. It is shown that decay modes involving the eta are not expected to contribute more than 3% to this excess

  7. The heating of plasma focus electrodes

    International Nuclear Information System (INIS)

    Angeli, E; Frignani, M; Mannucci, S; Rocchi, F; Sumini, M; Tartari, A

    2006-01-01

    Plasma focus (PF) technology development today is strictly related to the possibility of a high frequency repetitive working regime. One of the more relevant obstacles to this goal is the heating of structural components due to direct interaction with plasma. In this paper, temperature decay measurements of the inner electrode of a 7 kJ Mather type PF are presented. Data from several series of shots at different bank energies are analysed and compared with theoretical and numerical models. Two possible scale laws are derived from the experimental data to correlate thermal deposition with bank energy. It is found that a fraction of about 10% of total energy is released to the inner electrode. Finally, after some considerations about the cooling and heating mechanisms, an analysis on maximum temperature sustained by materials is presented

  8. Induced nuclear beta decay

    International Nuclear Information System (INIS)

    Reiss, H.R.

    1986-01-01

    Certain nuclear beta decay transitions normally inhibited by angular momentum or parity considerations can be induced to occur by the application of an electromagnetic field. Such decays can be useful in the controlled production of power, and in fission waste disposal

  9. B decays to baryons

    Indian Academy of Sciences (India)

    We note that two-body decays to baryons are suppressed relative to three- and four-body decays. In most of these analyses, the invariant baryon–antibaryon mass shows an enhancement near the threshold. We propose a phenomenological interpretation of this quite common feature of hadronization to baryons.

  10. Multiple preequilibrium decay processes

    International Nuclear Information System (INIS)

    Blann, M.

    1987-11-01

    Several treatments of multiple preequilibrium decay are reviewed with emphasis on the exciton and hybrid models. We show the expected behavior of this decay mode as a function of incident nucleon energy. The algorithms used in the hybrid model treatment are reviewed, and comparisons are made between predictions of the hybrid model and a broad range of experimental results. 24 refs., 20 figs

  11. Aspects of B decays

    International Nuclear Information System (INIS)

    Faller, Sven

    2011-01-01

    B-meson decays are a good probe for testing the flavour sector of the standard model of particle physics. The standard model describes at present all experimental data satisfactorily, although some ''tensions'' exist, i.e. two to three sigma deviations from the predictions, in particular in B decays. The arguments against the standard model are thus purely theoretical. These tensions between experimental data and theoretical predictions provide an extension of the standard model by new physics contributions. Within the flavour sector main theoretical uncertainties are related to the hadronic matrix elements. For exclusive semileptonic anti B → D (*) l anti ν decays QCD sum rule techniques, which are suitable for studying hadronic matrix elements, however, with substantial, but estimable hadronic uncertainties, are used. The exploration of new physics effects in B-meson decays is done in an twofold way. In exclusive semileptonic anti B → D (*) l anti ν decays the effect of additional right-handed vector as well as left- and right-handed scalar and tensor hadronic current structures in the decay rates and the form factors are studied at the non-recoil point. As a second approach one studied the non-leptonic B 0 s →J/ψφ and B 0 →J/ψK S,L decays discussing CP violating effects in the time-dependent decay amplitudes by considering new physics phase in the B 0 - anti B 0 mixing phase. (orig.)

  12. Decay of hypernuclei

    International Nuclear Information System (INIS)

    Bando, H.

    1985-01-01

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

  13. Rare Decays at LHCb

    CERN Document Server

    Belyaev, Ivan

    2006-01-01

    Rare loop-induced decays are sensitive to New Physics in many Standard Model extensions. In this paper we discuss the reconstruction of the radiative penguin decays $B^0_d \\to K^{*0} \\gamma, B^0_s \\to \\phi \\gamma , B^0_d \\to \\omega \\gamma, \\Lambda_b \\to \\Lambda \\gamma$, the electroweak penguin decays $B^0_d \\to K^{*0} \\mu^+ \\mu^-, B^+_u \\to K^+ \\mu^+ \\mu^-$, the gluonic penguin decays $B^0_d \\to \\phi K^0_S, B^0_s \\to \\phi \\phi$, and the decay $B^0_s \\to \\mu^+\\mu^-$ at LHCb. The selection criteria, evaluated efficiencies, expected annual yields and $B/S$ estimates are presented.

  14. Results of PMIP2 coupled simulations of the Mid-Holocene and Last Glacial Maximum – Part 2: feedbacks with emphasis on the location of the ITCZ and mid- and high latitudes heat budget

    Directory of Open Access Journals (Sweden)

    Y. Zhao

    2007-06-01

    Full Text Available A set of coupled ocean-atmosphere(-vegetation simulations using state of the art climate models is now available for the Last Glacial Maximum (LGM and the Mid-Holocene (MH through the second phase of the Paleoclimate Modeling Intercomparison Project (PMIP2. Here we quantify the latitudinal shift of the location of the Intertropical Convergence Zone (ITCZ in the tropical regions during boreal summer and the change in precipitation in the northern part of the ITCZ. For both periods the shift is more pronounced over the continents and East Asia. The maritime continent is the region where the largest spread is found between models. We also clearly establish that the larger the increase in the meridional temperature gradient in the tropical Atlantic during summer at the MH, the larger the change in precipitation over West Africa. The vegetation feedback is however not as large as found in previous studies, probably due to model differences in the control simulation. Finally, we show that the feedback from snow and sea-ice at mid and high latitudes contributes for half of the cooling in the Northern Hemisphere for the LGM, with the remaining being achieved by the reduced CO2 and water vapour in the atmosphere. For the MH the snow and albedo feedbacks strengthen the spring cooling and enhance the boreal summer warming, whereas water vapour reinforces the late summer warming. These feedbacks are modest in the Southern Hemisphere. For the LGM most of the surface cooling is due to CO2 and water vapour.

  15. β decay studies of n-rich Cs isotopes with the ISOLDE Decay Station

    Science.gov (United States)

    Lică, R.; Benzoni, G.; Morales, A. I.; Borge, M. J. G.; Fraile, L. M.; Mach, H.; Madurga, M.; Sotty, C.; Vedia, V.; De Witte, H.; Benito, J.; Berry, T.; Blasi, N.; Bracco, A.; Camera, F.; Ceruti, S.; Charviakova, V.; Cieplicka-Oryńczak, N.; Costache, C.; Crespi, F. C. L.; Creswell, J.; Fernández-Martínez, G.; Fynbo, H.; Greenlees, P.; Homm, I.; Huyse, M.; Jolie, J.; Karayonchev, V.; Köster, U.; Konki, J.; Kröll, T.; Kurcewicz, J.; Kurtukian-Nieto, T.; Lazarus, I.; Leoni, S.; Lund, M.; Marginean, N.; Marginean, R.; Mihai, C.; Mihai, R.; Negret, A.; Orduz, A.; Patyk, Z.; Pascu, S.; Pucknell, V.; Rahkila, P.; Regis, J. M.; Rotaru, F.; Saed-Sami, N.; Sánchez-Tembleque, V.; Stanoiu, M.; Tengblad, O.; Thuerauf, M.; Turturica, A.; Van Duppen, P.; Warr, N.

    2017-05-01

    Neutron-rich Ba isotopes are expected to exhibit octupolar correlations, reaching their maximum in isotopes around mass A = 146. The odd-A neutron-rich members of this isotopic chain show typical patterns related to non-axially symmetric shapes, which are however less marked compared to even-A ones, pointing to a major contribution from vibrations. In the present paper we present results from a recent study focused on 148-150Cs β-decay performed at the ISOLDE Decay Station equipped with fast-timing detectors. A detailed analysis of the measured decay half-lives and decay scheme of 149Ba is presented, giving a first insight in the structure of this neutron-rich nucleus.

  16. Effects of Angle of Rotation on Pool Boiling Heat Transfer of V-shape Tube Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myeong-Gie [Andong National University, Andong (Korea, Republic of)

    2016-10-15

    The most important facility for the systems is a passive heat exchanger that transfers core decay heat to the cold water in a water storage tank under atmospheric pressure. Since the space for the installation of the heat exchanger is usually limited, developing more efficient heat exchangers is important. In general, pool boiling is generated on the surface of the heat exchanging tube. The major design parameter of the heat exchanger is a tube arrangement. The upper tube is affected by the lower tube and the enhancement of the heat transfer on the upper tube is estimated by the bundle effect. Since heat transfer is related to the conditions of a tube surface, bundle geometries, and a liquid type, lots of studies have been carried out for the combinations of those parameters. An experimental study was performed to investigate the effects of the angle of rotation on pool boiling heat transfer of a V-shape tube bundle. For the test, two smooth stainless steel tubes of 19 mm outside diameter and the water at atmospheric pressure were used. The enhancement of the heat transfer is clearly observed when the angle becomes to 90° where the upper tube has the maximum region of influence by the lower tube. The convective flow and liquid agitation enhance heat transfer while the coalescence of the bubbles deteriorates heat transfer.

  17. Charm Decays at BABAR

    International Nuclear Information System (INIS)

    Charles, M.

    2004-01-01

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

  18. Iconic decay in schizophrenia.

    Science.gov (United States)

    Hahn, Britta; Kappenman, Emily S; Robinson, Benjamin M; Fuller, Rebecca L; Luck, Steven J; Gold, James M

    2011-09-01

    Working memory impairment is considered a core deficit in schizophrenia, but the precise nature of this deficit has not been determined. Multiple lines of evidence implicate deficits at the encoding stage. During encoding, information is held in a precategorical sensory store termed iconic memory, a literal image of the stimulus with high capacity but rapid decay. Pathologically increased iconic decay could reduce the number of items that can be transferred into working memory before the information is lost and could thus contribute to the working memory deficit seen in the illness. The current study used a partial report procedure to test the hypothesis that patients with schizophrenia (n = 37) display faster iconic memory decay than matched healthy control participants (n = 28). Six letters, arranged in a circle, were presented for 50 ms. Following a variable delay of 0-1000 ms, a central arrow cue indicated the item to be reported. In both patients and control subjects, recall accuracy decreased with increasing cue delay, reflecting decay of the iconic representation of the stimulus array. Patients displayed impaired memory performance across all cue delays, consistent with an impairment in working memory, but the rate of iconic memory decay did not differ between patients and controls. This provides clear evidence against faster loss of iconic memory representations in schizophrenia, ruling out iconic decay as an underlying source of the working memory impairment in this population. Thus, iconic decay rate can be added to a growing list of unimpaired cognitive building blocks in schizophrenia.

  19. Weak radiative hyperon decays

    International Nuclear Information System (INIS)

    Roberts, B.L.; Booth, E.C.; Gall, K.P.; McIntyre, E.K.; Miller, J.P.; Whitehouse, D.A.; Bassalleck, B.; Hall, J.R.; Larson, K.D.; Wolfe, D.M.; Fickinger, W.J.; Robinson, D.K.; Hallin, A.L.; Hasinoff, M.D.; Measday, D.F.; Noble, A.J.; Waltham, C.E.; Hessey, N.P.; Lowe, J.; Horvath, D.; Salomon, M.

    1990-01-01

    New measurements of the Σ + and Λ weak radiative decays are discussed. The hyperons were produced at rest by the reaction K - p → Yπ where Y = Σ + or Λ. The monoenergetic pion was used to tag the hyperon production, and the branching ratios were determined from the relative amplitudes of Σ + → pγ to Σ + → pπ 0 and Λ → nγ to Λ → nπ 0 . The photons from weak radiative decays and from π 0 decays were detected with modular NaI arrays. (orig.)

  20. SYMPOSIUM: Rare decays

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Late last year, a symposium entitled 'Rare Decays' attracted 115 participants to a hotel in Vancouver, Canada. These participants were particle physicists interested in checking conventional selection rules to look for clues of possible new behaviour outside today's accepted 'Standard Model'. For physicists, 'rare decays' include processes that have so far not been seen, explicitly forbidden by the rules of the Standard Model, or processes highly suppressed because the decay is dominated by an easier route, or includes processes resulting from multiple transitions

  1. New results for rare muon decays

    International Nuclear Information System (INIS)

    Mischke, R.E.; Bolton, R.D.; Bowman, J.D.

    1986-01-01

    Branching-ratio limits obtained with the Crystal Box detector are presented for the rare muon decays μ → eee, μ → eγ, and μ → eγγ. These decays, which violate the conservation of separate lepton-family numbers, are expected to occur in many extensions to the standard model. We found no candidates for the decay μ → eee, yielding an upper limit for the branching ratio of B/sub μ3e/ -11 (90% C.L.). A maximum-likelihood analysis of the μ → eγ candidates yields an upper limit of B/sub μeγ/ -11 and an analogous analysis of μ → eγγ candidates gives an upper limit of B/sub μeγγ/ -11 . These results strengthen the constraints on models that allow transitions between lepton families

  2. New results for rare muon decays

    International Nuclear Information System (INIS)

    Mischke, R.E.; Bolton, R.D.; Bowman, J.D.; Cooper, M.D.; Frank, J.S.; Hallin, A.L.; Heusi, P.A.; Hoffman, C.M.; Hogan, G.E.; Mariam, F.G.; Matis, H.S.; Nagle, D.E.; Piilonen, L.E.; Sandberg, V.D.; Sanders, G.H.; Sennhauser, U.; Werbeck, R.; Williams, R.A.; Wilson, S.L.; Hofstadter, R.; Hughes, E.B.; Ritter, M.W.; Grosnick, D.; Wright, S.C.; Highland, V.L.; McDonough, J.

    1986-01-01

    Branching-ratio limits obtained with the Crystal Box detector are presented for the rare muon decays μ→eee, μ→eγ, and μ→eγγ. These decays, which violate the conservation of separate lepton-family numbers, are expected to occur in many extensions to the standard model. We found no candidates for the decay μ→eee, yielding an upper limit for the branching ratio of B μ3e -11 (90% C.L.). A maximum-likelihood analysis of the μ→eγ candidates yields an upper limit of B μeγ -11 and an analogous analysis of μ→eγγ candidates gives an upper limit of B μeγγ -11 . These results strengthen the constraints on models that allow transitions between lepton families. (orig.)

  3. Wave propagation model of heat conduction and group speed

    Science.gov (United States)

    Zhang, Long; Zhang, Xiaomin; Peng, Song

    2018-03-01

    In view of the finite relaxation model of non-Fourier's law, the Cattaneo and Vernotte (CV) model and Fourier's law are presented in this work for comparing wave propagation modes. Independent variable translation is applied to solve the partial differential equation. Results show that the general form of the time spatial distribution of temperature for the three media comprises two solutions: those corresponding to the positive and negative logarithmic heating rates. The former shows that a group of heat waves whose spatial distribution follows the exponential function law propagates at a group speed; the speed of propagation is related to the logarithmic heating rate. The total speed of all the possible heat waves can be combined to form the group speed of the wave propagation. The latter indicates that the spatial distribution of temperature, which follows the exponential function law, decays with time. These features show that propagation accelerates when heated and decelerates when cooled. For the model media that follow Fourier's law and correspond to the positive heat rate of heat conduction, the propagation mode is also considered the propagation of a group of heat waves because the group speed has no upper bound. For the finite relaxation model with non-Fourier media, the interval of group speed is bounded and the maximum speed can be obtained when the logarithmic heating rate is exactly the reciprocal of relaxation time. And for the CV model with a non-Fourier medium, the interval of group speed is also bounded and the maximum value can be obtained when the logarithmic heating rate is infinite.

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

    International Nuclear Information System (INIS)

    Vossebrecker, H.; Groenefeld, G.

    1976-12-01

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

  5. Teleportation via decay

    Indian Academy of Sciences (India)

    therefore normally plays a negative role in quantum information processing [1]. ... of a decay be used in a fruitful way for quantum information process- ing? ..... The model independent portions of the analysis of communication through a noisy.

  6. Decay of Hoyle state

    Indian Academy of Sciences (India)

    2014-11-02

    Nov 2, 2014 ... T K RANA, C BHATTACHARYA, S KUNDU, ... of various direct 3α decay mechanisms of the Hoyle state. ... Pramana – J. Phys., Vol. ... FMD predicts a compact triangle shape and LEFT predicts a bent arm chain structure,.

  7. RARE KAON DECAYS

    International Nuclear Information System (INIS)

    LITTENBERG, L.

    2005-01-01

    Lepton flavor violation (LFV) experiments have probed sensitivities corresponding to mass scales of well over 100 TeV, making life difficult for models predicting accessible LFV in kaon decay and discouraging new dedicated experiments of this type

  8. Neutrinoless double beta decay

    Indian Academy of Sciences (India)

    2012-10-06

    Oct 6, 2012 ... Anyhow, the 'multi-isotope' ansatz is needed to compensate for matrix element ... The neccessary half-life requirement to touch this ... site energy depositions (like double beta decay) and multiple site interactions (most of.

  9. Cavities/Tooth Decay

    Science.gov (United States)

    ... milk, ice cream, honey, sugar, soda, dried fruit, cake, cookies, hard candy and mints, dry cereal, and ... teeth can wear down and gums may recede, making teeth more vulnerable to root decay. Older adults ...

  10. Inflaton decay in supergravity

    Energy Technology Data Exchange (ETDEWEB)

    Endo, M.; Takahashi, F. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Yanagida, T.T. [Tokyo Univ. (Japan). Dept. of Physics]|[Tokyo Univ. (Japan). Research Center for the Early Universe

    2007-06-15

    We discuss inflaton decay in supergravity, taking account of the gravitational effects. It is shown that, if the inflaton has a nonzero vacuum expectation value, it generically couples to any matter fields that appear in the superpotential at the tree level, and to any gauge sectors through anomalies in the supergravity. Through these processes, the inflaton generically decays into the supersymmetry breaking sector, producing many gravitinos. The inflaton also directly decays into a pair of the gravitinos. We derive constraints on both inflation models and supersymmetry breaking scenarios for avoiding overproduction of the gravitinos. Furthermore, the inflaton naturally decays into the visible sector via the top Yukawa coupling and SU(3){sub C} gauge interactions. (orig.)

  11. Inflaton decay in supergravity

    International Nuclear Information System (INIS)

    Endo, M.; Takahashi, F.; Yanagida, T.T.; Tokyo Univ.

    2007-06-01

    We discuss inflaton decay in supergravity, taking account of the gravitational effects. It is shown that, if the inflaton has a nonzero vacuum expectation value, it generically couples to any matter fields that appear in the superpotential at the tree level, and to any gauge sectors through anomalies in the supergravity. Through these processes, the inflaton generically decays into the supersymmetry breaking sector, producing many gravitinos. The inflaton also directly decays into a pair of the gravitinos. We derive constraints on both inflation models and supersymmetry breaking scenarios for avoiding overproduction of the gravitinos. Furthermore, the inflaton naturally decays into the visible sector via the top Yukawa coupling and SU(3) C gauge interactions. (orig.)

  12. Double beta decay: experiments

    International Nuclear Information System (INIS)

    Fiorini, Ettore

    2006-01-01

    The results obtained so far and those of the running experiments on neutrinoless double beta decay are reviewed. The plans for second generation experiments, the techniques to be adopted and the expected sensitivities are compared and discussed

  13. Streamer chamber: pion decay

    CERN Multimedia

    1992-01-01

    The real particles produced in the decay of a positive pion can be seen in this image from a streamer chamber. Streamer chambers consist of a gas chamber through which a strong pulsed electric field is passed, creating sparks as a charged particle passes through it. A magnetic field is added to cause the decay products to follow curved paths so that their charge and momentum can be measured.

  14. Aspects of B decays

    Energy Technology Data Exchange (ETDEWEB)

    Faller, Sven

    2011-03-04

    B-meson decays are a good probe for testing the flavour sector of the standard model of particle physics. The standard model describes at present all experimental data satisfactorily, although some ''tensions'' exist, i.e. two to three sigma deviations from the predictions, in particular in B decays. The arguments against the standard model are thus purely theoretical. These tensions between experimental data and theoretical predictions provide an extension of the standard model by new physics contributions. Within the flavour sector main theoretical uncertainties are related to the hadronic matrix elements. For exclusive semileptonic anti B {yields} D{sup (*)}l anti {nu} decays QCD sum rule techniques, which are suitable for studying hadronic matrix elements, however, with substantial, but estimable hadronic uncertainties, are used. The exploration of new physics effects in B-meson decays is done in an twofold way. In exclusive semileptonic anti B {yields} D{sup (*)}l anti {nu} decays the effect of additional right-handed vector as well as left- and right-handed scalar and tensor hadronic current structures in the decay rates and the form factors are studied at the non-recoil point. As a second approach one studied the non-leptonic B{sup 0}{sub s}{yields}J/{psi}{phi} and B{sup 0}{yields}J/{psi}K{sub S,L} decays discussing CP violating effects in the time-dependent decay amplitudes by considering new physics phase in the B{sup 0}- anti B{sup 0} mixing phase. (orig.)

  15. Tau decays into kaons

    International Nuclear Information System (INIS)

    Finkemeier, M.; Mirkes, E.

    1995-04-01

    Predictions for semi-leptonic decay rates of the τ lepton into two meson final states and three meson final states are derived. The hadronic matrix elements are expressed in terms of form factors, which can be predicted by chiral Lagrangians supplemented by informations about all possible low-lying resonances in the different channels. Isospin symmetry relations among the different final states are carefully taken into account. The calculated brancing ratios are compared with measured decay rates where data are available

  16. Iconic Decay in Schizophrenia

    OpenAIRE

    Hahn, Britta; Kappenman, Emily S.; Robinson, Benjamin M.; Fuller, Rebecca L.; Luck, Steven J.; Gold, James M.

    2010-01-01

    Working memory impairment is considered a core deficit in schizophrenia, but the precise nature of this deficit has not been determined. Multiple lines of evidence implicate deficits at the encoding stage. During encoding, information is held in a precategorical sensory store termed iconic memory, a literal image of the stimulus with high capacity but rapid decay. Pathologically increased iconic decay could reduce the number of items that can be transferred into working memory before the info...

  17. Annihilation decays of bottomonium

    International Nuclear Information System (INIS)

    Monteiro, Antony Prakash; Bhat, Manjunath; D'Souza, Praveen P.; Vijaya Kumar, K.B.

    2016-01-01

    The bound state of a bottom quark b and its anti quark b-bar known as bottomonium was first seen in the spectrum of μμ"- pairs produced in 400 GeV proton-nucleus collisions at Fermilab. It was discovered as spin triplet states ϒ(1S), ϒ(2S) and ϒ(3S) by E288 collaboration at Fermilab. We have calculated annihilation decay widths of bottomonium states. The calculated decay widths are presented

  18. Rare psi decays

    International Nuclear Information System (INIS)

    Partridge, R.

    1986-01-01

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

  19. Decays of supernova neutrinos

    International Nuclear Information System (INIS)

    Lindner, Manfred; Ohlsson, Tommy; Winter, Walter

    2002-01-01

    Supernova neutrinos could be well-suited for probing neutrino decay, since decay may be observed even for very small decay rates or coupling constants. We will introduce an effective operator framework for the combined description of neutrino decay and neutrino oscillations for supernova neutrinos, which can especially take into account two properties: one is the radially symmetric neutrino flux, allowing a decay product to be re-directed towards the observer even if the parent neutrino had a different original direction of propagation. The other is decoherence because of the long baselines for coherently produced neutrinos. We will demonstrate how to use this effective theory to calculate the time-dependent fluxes at the detector. In addition, we will show the implications of a Majoron-like decay model. As a result, we will demonstrate that for certain parameter values one may observe some effects which could also mimic signals similar to the ones expected from supernova models, making it in general harder to separate neutrino and supernova properties

  20. Rare and forbidden decays

    CERN Document Server

    Trampetic, Josip

    2002-01-01

    In these lectures I first cover radiative and semileptonic B decays, including the QCD corrections for the quark subprocesses. The exclusive modes and the evaluation of the hadronic matrix elements, i.e. the relevant hadronic form factors, are the second step. Small effects due to the long-distance, spectator contributions, etc. are discussed next. The second section we started with non-leptonic decays, typically $B \\to \\pi\\pi, K\\pi, \\rho\\pi,...$ We describe in more detail our prediction for decays dominated by the $b\\to s \\eta_c$ transition. Reports on the most recent experimental results are given at the end of each subsection. In the second part of the lectures I discuss decays forbidden by the Lorentz and gauge invariance, and due to the violation of the angular moment conservation, generally called the Standard Model-forbiden decays. However, the non-commutative QED and/or non-commutative Standard Model (NCSM), developed in a series of works in the last few years allow some of those decay modes. These ar...

  1. Credal Networks under Maximum Entropy

    OpenAIRE

    Lukasiewicz, Thomas

    2013-01-01

    We apply the principle of maximum entropy to select a unique joint probability distribution from the set of all joint probability distributions specified by a credal network. In detail, we start by showing that the unique joint distribution of a Bayesian tree coincides with the maximum entropy model of its conditional distributions. This result, however, does not hold anymore for general Bayesian networks. We thus present a new kind of maximum entropy models, which are computed sequentially. ...

  2. 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)

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

    International Nuclear Information System (INIS)

    Siebert, H.W.

    1989-01-01

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

  4. Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

  5. Thulium-170 heat source

    Science.gov (United States)

    Walter, Carl E.; Van Konynenburg, Richard; VanSant, James H.

    1992-01-01

    An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.

  6. CP violation in B decay

    OpenAIRE

    Yamamoto, Hitoshi

    2001-01-01

    We review the physics of CP violation in B decays. After introducing the CKM matrix and how it causes CP violation, we cover three types of CP violation that can occur in B decays: CP violation in mixing, CP violation by mixing-decay interference, and CP violation in decay.

  7. Radioactive decay and labeled compounds

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

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

  8. Strength loss in decayed wood

    Science.gov (United States)

    Rebecca E. Ibach; Patricia K. Lebow

    2014-01-01

    Wood is a durable engineering material when used in an appropriate manner, but it is susceptible to biological decay when a log, sawn product, or final product is not stored, handled, or designed properly. Even before the biological decay of wood becomes visually apparent, the decay can cause the wood to become structurally unsound. The progression of decay to that...

  9. Sigma beta decay

    International Nuclear Information System (INIS)

    Newman, D.E.

    1975-01-01

    Describes an experiment to measure beta decays of the sigma particle. Sigmas produced by stopping a K - beam in a liquid hydrogen target decayed in the following reactions: Kp → Σπ; Σ → Neν. The electron and pion were detected by wire spark chambers in a magnetic spectrometer and by plastic scintillators, and were differentiated by a threshold gas Cherenkov counter. The neutron was detected by liquid scintillation counters. The data (n = 3) shell electrons or the highly excited electrons decay first. Instead, it is suggested that when there are two to five electrons in highly excited states immediately after a heavy ion--atom collision the first transitions to occur will be among highly excited Rydberg states in a cascade down to the 4s, 4p, and 3d-subshells. If one of the long lived states becomes occupied by electrons promoted during the collision or by electrons falling from higher levels, it will not decay until after the valence shell decays. LMM rates calculated to test the methods used are compared to previous works. The mixing coefficients are given in terms of the states 4s4p, 45sp+-, and 5s5p. The applicability of Cooper, Fano, and Prats' discussion of the energies and transition rates of doubly excited states is considered

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

    International Nuclear Information System (INIS)

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

    1986-08-01

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

  11. Beta and muon decays

    International Nuclear Information System (INIS)

    Galindo, A.; Pascual, P.

    1967-01-01

    These notes represent a series of lectures delivered by the authors in the Junta de Energia Nuclear, during the Spring term of 1965. They were devoted to graduate students interested in the Theory of Elementary Particles. Special emphasis was focussed into the computational problems. Chapter I is a review of basic principles (Dirac equation, transition probabilities, final state interactions.) which will be needed later. In Chapter II the four-fermion punctual Interaction is discussed, Chapter III is devoted to the study of beta-decay; the main emphasis is given to the deduction of the formulae corresponding to electron-antineutrino correlation, electron energy spectrum, lifetimes, asymmetry of electrons emitted from polarized nuclei, electron and neutrino polarization and time reversal invariance in beta decay. In Chapter IV we deal with the decay of polarized muons with radiative corrections. Chapter V is devoted to an introduction to C.V.C. theory. (Author)

  12. Decay of superdeformed bands

    International Nuclear Information System (INIS)

    Carpenter, M.P.; Khoo, T.L.; Lauritsen, T.

    1995-01-01

    One of the major challenges in the study of superdeformation is to directly connect the large number of superdeformed bands now known to the yrast states. In this way, excitation energies, spins and parities can be assigned to the levels in the second well which is essential to establish the collective and single-particle components of these bands. This paper will review some of the progress which has been made to understand the decay of superdeformed bands using the new arrays including the measurement of the total decay spectrum and the establishment of direct one-step decays from the superdeformed band to the yrast line in 194 Hg. 42 refs., 5 figs

  13. Beta and muon decays

    Energy Technology Data Exchange (ETDEWEB)

    Galindo, A; Pascual, P

    1967-07-01

    These notes represent a series of lectures delivered by the authors in the Junta de Energia Nuclear, during the Spring term of 1965. They were devoted to graduate students interested in the Theory of Elementary Particles. Special emphasis was focussed into the computational problems. Chapter I is a review of basic principles (Dirac equation, transition probabilities, final state interactions.) which will be needed later. In Chapter II the four-fermion punctual Interaction is discussed, Chapter III is devoted to the study of beta-decay; the main emphasis is given to the deduction of the formulae corresponding to electron-antineutrino correlation, electron energy spectrum, lifetimes, asymmetry of electrons emitted from polarized nuclei, electron and neutrino polarization and time reversal invariance in beta decay. In Chapter IV we deal with the decay of polarized muons with radiative corrections. Chapter V is devoted to an introduction to C.V.C. theory. (Author)

  14. Suppressed Charmed B Decay

    Energy Technology Data Exchange (ETDEWEB)

    Snoek, Hella Leonie [Vrije Univ., Amsterdam (Netherlands)

    2009-06-02

    This thesis describes the measurement of the branching fractions of the suppressed charmed B0 → D*- a0+ decays and the non-resonant B0 → D*- ηπ+ decays in approximately 230 million Υ(4S) → B$\\bar{B}$ events. The data have been collected with the BABAR detector at the PEP-II B factory at the Stanford Linear Accelerator Center in California. Theoretical predictions of the branching fraction of the B0 → D*- a{sub 0}+ decays show large QCD model dependent uncertainties. Non-factorizing terms, in the naive factorization model, that can be calculated by QCD factorizing models have a large impact on the branching fraction of these decay modes. The predictions of the branching fractions are of the order of 10-6. The measurement of the branching fraction gives more insight into the theoretical models. In general a better understanding of QCD models will be necessary to conduct weak interaction physics at the next level. The presence of CP violation in electroweak interactions allows the differentiation between matter and antimatter in the laws of physics. In the Standard Model, CP violation is incorporated in the CKM matrix that describes the weak interaction between quarks. Relations amongst the CKM matrix elements are used to present the two relevant parameters as the apex of a triangle (Unitarity Triangle) in a complex plane. The over-constraining of the CKM triangle by experimental measurements is an important test of the Standard Model. At this moment no stringent direct measurements of the CKM angle γ, one of the interior angles of the Unitarity Triangle, are available. The measurement of the angle γ can be performed using the decays of neutral B mesons. The B0 → D*- a0+ decay is sensitive to the angle γ and, in comparison to the current decays that are being employed, could significantly

  15. Weak interactions: muon decay

    International Nuclear Information System (INIS)

    Sachs, A.M.; Sirlin, A.

    1975-01-01

    The traditional theory of the dominant mode of muon decay is presented, a survey of the experiments which have measured the observable features of the decay is given, and those things which can be learned about the parameters and nature of the theory from the experimental results are indicated. The following aspects of the theory of muon decay are presented first: general four-fermion theory, two-component theory of the neutrino, V--A theory, two-component and V--A theories vs general four-fermion theory, intermediate-boson hypothesis, radiative corrections, radiative corrections in the intermediate-boson theory, and endpoint singularities and corrections of order α 2 . Experiments on muon lifetime, isotropic electron spectrum, total asymmetry and energy dependence of asymmetry of electrons from polarized muons, and electron polarization are described, and a summary of experimental results is given. 7 figures, 2 tables, 109 references

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-10-15

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

  17. 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)

  18. 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)

  19. Performance of heat engines with non-zero heat capacity

    International Nuclear Information System (INIS)

    Odes, Ron; Kribus, Abraham

    2013-01-01

    Highlights: ► Finite heat capacity is a second irreversibility mechanism in addition to thermal resistance. ► Heat capacity introduces thermal transients and reverse heat flow. ► Engine maximum power and efficiency are lower for finite heat capacity. ► Implementing the optimal engine cycle requires active control. - Abstract: The performance of a heat engine is analyzed subject to two types of irreversibility: a non-zero heat capacity, together with the more common finite heat transfer rate between the engine and the external heat reservoirs. The heat capacity represents an engine body that undergoes significant temperature variations during the engine cycle. An option to cut off the heat exchange between the engine and the external surrounding for part of the engine cycle is also explored. A variational approach was taken to find the engine’s internal temperature profile (which defines the internal thermodynamic cycle) that would produce maximum power. The maximum power is shown to be lower than the case of zero heat capacity, due to a loss of heat that is stored in the engine body and then lost, bypassing the thermodynamic cycle. The maximum efficiency and the efficiency at maximum power are also lower than the zero heat capacity case. Similar to the Curzon–Ahlborn analysis, power can be traded for increased efficiency, but for high heat capacity, the range of efficiency that is available for such a trade is diminished. Isolating the engine during part of the cycle reduces maximum power, but the efficiency at maximum power and the maximum efficiency are improved, due to better exploitation of heat stored in the engine body. This might be useful for real engines that are limited by the internal energy change during a single engine cycle or by the operating frequency, leading to a broader power–efficiency curve.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  1. Sequential decay of Reggeons

    International Nuclear Information System (INIS)

    Yoshida, Toshihiro

    1981-01-01

    Probabilities of meson production in the sequential decay of Reggeons, which are formed from the projectile and the target in the hadron-hadron to Reggeon-Reggeon processes, are investigated. It is assumed that pair creation of heavy quarks and simultaneous creation of two antiquark-quark pairs are negligible. The leading-order terms with respect to ratio of creation probabilities of anti s s to anti u u (anti d d) are calculated. The production cross sections in the target fragmentation region are given in terms of probabilities in the initial decay of the Reggeons and an effect of manyparticle production. (author)

  2. Do protons decay

    International Nuclear Information System (INIS)

    Litchfield, P.J.

    1984-09-01

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

  3. 103Pd decay

    International Nuclear Information System (INIS)

    Belyavenko, V.S.; Borozenets, G.P.; Vishnevskij, I.N.; Zheltonozhskij, V.A.

    1986-01-01

    103 Pd decay in different chemical states has been investigated. The change of the partial half-life period equal to 0.67±0.15% has been detected. The γ-spectrum has been measured to a high precision. The new data have been obtained on population probabilities of 103 Rh excited states and the total energy of decay for 103 Pd has been determined to a high precision (543.0±0.8). The values of log ft have been determined

  4. Decay of 99Mo

    International Nuclear Information System (INIS)

    Dickens, J.K.; Love, T.A.

    1976-01-01

    Relative intensities for K x-rays and gamma rays emanating from 99 Mo in equilibrium with its 99 Tc* daughter have been measured using several Ge photon detectors. Combining these intensities with an evaluated set of electron-conversion coefficients has provided a set of absolute intensities for the observed gamma rays. The absolute intensity for the dominant 140.5-keV gamma ray in 99 Tc was determined to be 90.7 +- 0.6/100 99 Mo disintegrations for 99 Mo decay in equilibrium with decay of the 99 Tc* daughter

  5. Supersymmetry in Z' decays

    International Nuclear Information System (INIS)

    Corcella, G.

    2014-01-01

    I study the phenomenology of new heavy neutral gauge bosons Z', predicted by Grand Unification Theories-driven U(1)' gauge groups and by the sequential standard model. BSM (Beyond Standard Model) decays into supersymmetric final states are accounted for, besides the SM (Standard Model) modes usually investigated. I give an estimate of the number of supersymmetric events in Z' decays possibly expected at LHC, as well as of the product of the Z' cross section times the branching fraction into electron and muon pairs. (author)

  6. Effect of geometric factors on performance of a sodium to air heat exchanger in a fast breeder reactor

    International Nuclear Information System (INIS)

    Kannan, K.; Vinod, V.; Padmakumar, G.; Rudramoorthy, R.; Rajan, K.K.

    2015-01-01

    Highlights: • A heat exchanger analysis (HE) before scale up reduces excess heat transfer area. • Representative Elementary Volume analysis of a HE speeds up the solution. • The error in air temperature rise prediction by numerical across HE is within 5%. • When both pitches are reduced, the maximum increase in heat flux is experienced. • The experience has resulted in better design of next level heat exchangers. - Abstract: Prototype fast breeder reactor (PFBR) has a safety grade decay heat removal system whose performance depends on the effective functioning of natural convection heat exchangers called sodium to air heat exchangers. The development of Representative Elementary Volume (REV) model for the sodium to air heat exchanger is necessary to envisage its design and to study the effect of various factors for continuous improvement in design. With a Representative Elementary Volume, the hydrodynamic and heat transfer characteristics of the heat exchanger was studied and the results agree well with experimental data. The effect of longitudinal pitch and transverse pitch on the heat exchanger performance has been studied and an improvement of 22% in heat transfer is predicted

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Triton beta decay

    International Nuclear Information System (INIS)

    Saito, T.Y.; Wu, Y.; Ishikawa, S.; Sasakawa, T.

    1990-01-01

    Triton β-decay has been calculated using wave functions for 3 He and 3 H obtained from (Coulomb-modified) Faddeev equations for various interactions. We get a value for the Gamow-Teller matrix element of √3 (0.962±0.002) without regards to two- or three-nucleon inteactions. This value agrees with the experimental value. (orig.)

  9. Unparticles and muon decay

    International Nuclear Information System (INIS)

    Choudhury, Debajyoti; Ghosh, Dilip Kumar; Mamta

    2008-01-01

    Recently Georgi has discussed the possible existence of 'Unparticles' describable by operators having non-integral scaling dimensions. With the interaction of these with the Standard Model particles being constrained only by gauge and Lorentz symmetries, it affords a new source for lepton flavour violation. Current and future muon decay experiments are shown to be very sensitive to such scenarios

  10. Unparticles and muon decay

    Energy Technology Data Exchange (ETDEWEB)

    Choudhury, Debajyoti [Department of Physics and Astrophysics, University of Delhi, Delhi 110 007 (India); Ghosh, Dilip Kumar [Department of Physics and Astrophysics, University of Delhi, Delhi 110 007 (India)], E-mail: dkghosh@physics.du.ac.in; Mamta [Department of Physics, S.G.T.B. Khalsa College, University of Delhi, Delhi 110 007 (India)

    2008-01-03

    Recently Georgi has discussed the possible existence of 'Unparticles' describable by operators having non-integral scaling dimensions. With the interaction of these with the Standard Model particles being constrained only by gauge and Lorentz symmetries, it affords a new source for lepton flavour violation. Current and future muon decay experiments are shown to be very sensitive to such scenarios.

  11. Gluons in quarkonium decay

    International Nuclear Information System (INIS)

    Koller, K.; Walsh, T.

    1978-03-01

    We discuss what can be learned of the 3 S 1 quarkonium decay QantiQ → 3 gluoans QantiQ → γ + 2 gluons. The former is a way to find gluon jets and test QCD. The latter also allows us to measure gluoan + gluon → hadrons and look for pure gluonic resonances (glueballs). (orig.) [de

  12. Symmetry violating kaon decays

    International Nuclear Information System (INIS)

    Herczeg, P.

    1979-01-01

    An analysis of the muon number violating decay modes of the K-mesons is given. Subsequently, some new developments in the field of CP-violation are reviewed and the question of time-reversal invariance and the status of CPT-invariance are briefly considered. 42 references

  13. Double Beta Decay Experiments

    International Nuclear Information System (INIS)

    Piepke, A.

    2005-01-01

    The experimental observation of neutrino oscillations and thus neutrino mass and mixing gives a first hint at new particle physics. The absolute values of the neutrino mass and the properties of neutrinos under CP-conjugation remain unknown. The experimental investigation of the nuclear double beta decay is one of the key techniques for solving these open problems

  14. On the proton decay

    International Nuclear Information System (INIS)

    Fonda, L.; Ghirardi, G.C.; Weber, T.

    1983-07-01

    The problem of the proton decay is considered taking into account that in actual experiments there is an interaction of the proton with its environment which could imply an increase of its theoretical lifetime. It is seen that, by application of the time-energy uncertainty relation, no prolongation of the lifetime is obtained in this case. (author)

  15. Cosmology with decaying particles

    International Nuclear Information System (INIS)

    Turner, M.S.

    1984-09-01

    We consider a cosmological model in which an unstable massive relic particle species (denoted by X) has an initial mass density relative to baryons β -1 identically equal rho/sub X//rho/sub B/ >> 1, and then decays recently (redshift z less than or equal to 1000) into particles which are still relativistic today (denoted by R). We write down and solve the coupled equations for the cosmic scale factor a(t), the energy density in the various components (rho/sub X/, rho/sub R/, rho/sub B/), and the growth of linear density perturbations (delta rho/rho). The solutions form a one parameter (β) family of solutions; physically β -1 approx. = (Ω/sub R//Ω/sub NR/) x (1 + z/sub D/) = (ratio today of energy density of relativistic to nonrelativistic particles) x (1 + redshift of (decay)). We discuss the observational implications of such a cosmological model and compare our results to earlier results computed in the simultaneous decay approximation. In an appendix we briefly consider the case where one of the decay products of the X is massive and becomes nonrelativistic by the present epoch. 21 references

  16. Cosmology with decaying particles

    Energy Technology Data Exchange (ETDEWEB)

    Turner, M.S.

    1984-09-01

    We consider a cosmological model in which an unstable massive relic particle species (denoted by X) has an initial mass density relative to baryons ..beta../sup -1/ identically equal rho/sub X//rho/sub B/ >> 1, and then decays recently (redshift z less than or equal to 1000) into particles which are still relativistic today (denoted by R). We write down and solve the coupled equations for the cosmic scale factor a(t), the energy density in the various components (rho/sub X/, rho/sub R/, rho/sub B/), and the growth of linear density perturbations (delta rho/rho). The solutions form a one parameter (..beta..) family of solutions; physically ..beta../sup -1/ approx. = (..cap omega../sub R//..cap omega../sub NR/) x (1 + z/sub D/) = (ratio today of energy density of relativistic to nonrelativistic particles) x (1 + redshift of (decay)). We discuss the observational implications of such a cosmological model and compare our results to earlier results computed in the simultaneous decay approximation. In an appendix we briefly consider the case where one of the decay products of the X is massive and becomes nonrelativistic by the present epoch. 21 references.

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

    CERN Multimedia

    CERN. Geneva

    2018-01-01

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

  18. Parametric optimization of thermoelectric elements footprint for maximum power generation

    DEFF Research Database (Denmark)

    Rezania, A.; Rosendahl, Lasse; Yin, Hao

    2014-01-01

    The development studies in thermoelectric generator (TEG) systems are mostly disconnected to parametric optimization of the module components. In this study, optimum footprint ratio of n- and p-type thermoelectric (TE) elements is explored to achieve maximum power generation, maximum cost......-performance, and variation of efficiency in the uni-couple over a wide range of the heat transfer coefficient on the cold junction. The three-dimensional (3D) governing equations of the thermoelectricity and the heat transfer are solved using the finite element method (FEM) for temperature dependent properties of TE...... materials. The results, which are in good agreement with the previous computational studies, show that the maximum power generation and the maximum cost-performance in the module occur at An/Ap

  19. Nuclear decay data measurements at the INEL ISOL facility

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. 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)

  1. 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

  2. Ion cyclotron resonance heating

    International Nuclear Information System (INIS)

    Tajima, T.

    1982-01-01

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

  3. Rare B decays at LEP

    CERN Document Server

    Kluit, P M

    2001-01-01

    The results of the LEP experiments for rare B decays will be reviewed, covering hadronic final states, radiative and other rare decays and results for the inclusive charmless branching ratio. (8 refs).

  4. Maximum Entropy in Drug Discovery

    Directory of Open Access Journals (Sweden)

    Chih-Yuan Tseng

    2014-07-01

    Full Text Available Drug discovery applies multidisciplinary approaches either experimentally, computationally or both ways to identify lead compounds to treat various diseases. While conventional approaches have yielded many US Food and Drug Administration (FDA-approved drugs, researchers continue investigating and designing better approaches to increase the success rate in the discovery process. In this article, we provide an overview of the current strategies and point out where and how the method of maximum entropy has been introduced in this area. The maximum entropy principle has its root in thermodynamics, yet since Jaynes’ pioneering work in the 1950s, the maximum entropy principle has not only been used as a physics law, but also as a reasoning tool that allows us to process information in hand with the least bias. Its applicability in various disciplines has been abundantly demonstrated. We give several examples of applications of maximum entropy in different stages of drug discovery. Finally, we discuss a promising new direction in drug discovery that is likely to hinge on the ways of utilizing maximum entropy.

  5. Visible neutrino decay at DUNE

    Energy Technology Data Exchange (ETDEWEB)

    Coloma, Pilar [Fermilab; Peres, Orlando G. [ICTP, Trieste

    2017-05-09

    If the heaviest neutrino mass eigenstate is unstable, its decay modes could include lighter neutrino eigenstates. In this case part of the decay products could be visible, as they would interact at neutrino detectors via mixing. At neutrino oscillation experiments, a characteristic signature of such \\emph{visible neutrino decay} would be an apparent excess of events at low energies. We focus on a simple phenomenological model in which the heaviest neutrino decays as $\

  6. Paleoclassical electron heat transport

    International Nuclear Information System (INIS)

    Callen, J.D.

    2005-01-01

    Radial electron heat transport in low collisionality, magnetically-confined toroidal plasmas is shown to result from paleoclassical Coulomb collision processes (parallel electron heat conduction and magnetic field diffusion). In such plasmas the electron temperature equilibrates along magnetic field lines a long length L, which is the minimum of the electron collision length and a maximum effective half length of helical field lines. Thus, the diffusing field lines induce a radial electron heat diffusivity M ≅ L/(πR 0q ) ∼ 10 >> 1 times the magnetic field diffusivity η/μ 0 ≅ ν e (c/ω p ) 2 . The paleoclassical electron heat flux model provides interpretations for many features of 'anomalous' electron heat transport: magnitude and radial profile of electron heat diffusivity (in tokamaks, STs, and RFPs), Alcator scaling in high density plasmas, transport barriers around low order rational surfaces and near a separatrix, and a natural heat pinch (or minimum temperature gradient) heat flux form. (author)

  7. Character of decay instability

    International Nuclear Information System (INIS)

    Polovin, R.V.; Demutskii, V.P.

    1981-01-01

    If the initial wave is unstable in the upper half plane Im ω>0 and there are no branch points of the quasiwave number, or if waves traveling in the same direction coalesce at a branch point, the instability is convective. On the other hand, if a branch point k(ω) does exist in the upper half-plane Im ω>0, and not all the waves that merge at this point travel in the same direction, the instability is absolute. A Green's function that describes the evolution of the perturbations of the initial wave in space and in time is constructed. The growth rates of the decay instability of the harmonics are determined. The produced waves are richer in harmonics than the initial waves. It is shown that the decay instability of an Alfven wave is absolute

  8. Decay of 57Ni

    International Nuclear Information System (INIS)

    Santos Scardino, A.M. dos.

    1987-01-01

    The decay of 57 Ni to 57 Co was studied by gamma ray spectroscopy using both singles and coincidence spectra. The sources were obtained with the 58 Ni (Y,n) 57 Ni reaction. Natural metallic nickel was irradiated in the bremsstrahluhng beam of the linear accelerator of the Instituto de Fisica da Universidade de Sao Paulo with 30 MeV electrons. The singles espectra were taken with 104 cc HPGe detector and the coincidences espectra with 27 and 53cc Ge(Li) and 104 cc. HPGe detectors. The energies of transitions that follow the 57 Ni decay were measured using 56 Co as standard (which was obtained by (Y,np) reaction in 58 Ni) and taking into account the cascade cross-over relations. (author) [pt

  9. Electroweak penguin B decays

    CERN Document Server

    Nikodem, Thomas

    2016-01-01

    Flavour Changing Neutral Currents (FCNC) are sensitive probes for physics beyond the Standard Model (SM), so-called New Physics. An example of a FCNC is the $b \\to s$ quark transition described by the electroweak penguin Feynman diagram shown in Figure 1. In the SM such FCNC are only allowed with a loop structure (as e:g: shown in the figure) and not by tree level processes. In the loops heavy particles appear virtually and do not need to be on shell. Therefore also not yet discovered heavy particles with up to a mass $\\mathcal{O}$(TeV) could virtually contribute significantly to observables. Several recent measurements of electroweak penguin B decays exhibit interesting tensions with SM predictions, most prominently in the angular observable $P'_5$ 5 of the decay $B^0 \\to K^{*0} \\mu^+ \\mu^1$[1], which triggered a lot of discussion in the theory community [2]-[14].

  10. Decay /sup 133/Ba

    Energy Technology Data Exchange (ETDEWEB)

    Singh, K; Hasiza, M L; Grewal, B S; Sahota, H S

    1982-07-01

    The relative gamma ray intensities of transitions in the decay of /sup 133/Ba have been measured using an intrinsic Ge detector. The electron capture branching ratios have been determined for 81, 161, 384 and 437 keV levels. The attenuation effect of long half-life of 81 keV levels has been studied in solid and liquid media. The electron capture decay has been investigated by changing the concentration of ethylene-diamine-tetraacetic acid (EDTA) environment. The 5/2/sup +/ yields 5/2/sup +/ 79.67 keV transition has an E0 to E2 intensity qsub(k)sup(2) <= 0.31. 10 refs., 4 figures.

  11. Hypernuclear weak decay puzzle

    International Nuclear Information System (INIS)

    Barbero, C.; Horvat, D.; Narancic, Z.; Krmpotic, F.; Kuo, T.T.S.; Tadic, D.

    2002-01-01

    A general shell model formalism for the nonmesonic weak decay of the hypernuclei has been developed. It involves a partial wave expansion of the emitted nucleon waves, preserves naturally the antisymmetrization between the escaping particles and the residual core, and contains as a particular case the weak Λ-core coupling formalism. The extreme particle-hole model and the quasiparticle Tamm-Dancoff approximation are explicitly worked out. It is shown that the nuclear structure manifests itself basically through the Pauli principle, and a very simple expression is derived for the neutron- and proton-induced decays rates Γ n and Γ p , which does not involve the spectroscopic factors. We use the standard strangeness-changing weak ΛN→NN transition potential which comprises the exchange of the complete pseudoscalar and vector meson octets (π,η,K,ρ,ω,K * ), taking into account some important parity-violating transition operators that are systematically omitted in the literature. The interplay between different mesons in the decay of Λ 12 C is carefully analyzed. With the commonly used parametrization in the one-meson-exchange model (OMEM), the calculated rate Γ NM =Γ n +Γ p is of the order of the free Λ decay rate Γ 0 (Γ NM th congruent with Γ 0 ) and is consistent with experiments. Yet the measurements of Γ n/p =Γ n /Γ p and of Γ p are not well accounted for by the theory (Γ n/p th p th > or approx. 0.60Γ 0 ). It is suggested that, unless additional degrees of freedom are incorporated, the OMEM parameters should be radically modified

  12. Meson radiative decays

    International Nuclear Information System (INIS)

    Edwards, B.J.; Kamal, A.N.

    1979-04-01

    The status of decays of the kind V → Pγ and P → Vγviewed with special emphasis on the work done by the authors in this field. The low experimental value of GAMMA(rho → πγ) remains the outstanding problem. The lastest preliminary numbers from a Fermi Laboratory experiment go in the right direction but not far enough. 15 references

  13. Decay of 83Sr

    International Nuclear Information System (INIS)

    Yu Xiaohan; Shi Shuanghui; Gu Jiahui

    1997-01-01

    The decay of 83 Sr was reinvestigated using γ singles and γ-γ-t coincidence measurement. A new level scheme of Rb, which contains 41 excited levels and about 180 transitions, is constructed. 19 new levels were added to the old level scheme and 8 formerly adopted levels were denied. A new data set of branching ratio, log(ft) value and spin parity was obtained

  14. MULTIFLUID MAGNETOHYDRODYNAMIC TURBULENT DECAY

    International Nuclear Information System (INIS)

    Downes, T. P.; O'Sullivan, S.

    2011-01-01

    It is generally believed that turbulence has a significant impact on the dynamics and evolution of molecular clouds and the star formation that occurs within them. Non-ideal magnetohydrodynamic (MHD) effects are known to influence the nature of this turbulence. We present the results of a suite of 512 3 resolution simulations of the decay of initially super-Alfvenic and supersonic fully multifluid MHD turbulence. We find that ambipolar diffusion increases the rate of decay of the turbulence while the Hall effect has virtually no impact. The decay of the kinetic energy can be fitted as a power law in time and the exponent is found to be -1.34 for fully multifluid MHD turbulence. The power spectra of density, velocity, and magnetic field are all steepened significantly by the inclusion of non-ideal terms. The dominant reason for this steepening is ambipolar diffusion with the Hall effect again playing a minimal role except at short length scales where it creates extra structure in the magnetic field. Interestingly we find that, at least at these resolutions, the majority of the physics of multifluid turbulence can be captured by simply introducing fixed (in time and space) resistive terms into the induction equation without the need for a full multifluid MHD treatment. The velocity dispersion is also examined and, in common with previously published results, it is found not to be power law in nature.

  15. 152Eu decay

    International Nuclear Information System (INIS)

    Artamonova, K.P.; Vinogradov, V.M.; Grigor'ev, E.P.; Zolotavin, A.V.; Makarov, V.M.; Sergeev, V.O.; Usynko, T.M.

    1978-01-01

    The purpose of this paper is the measurement of the relative intensities of the most intensive conversion lines of 152 Eu, the determination of as reliable as possible magnitudes of the intensities of γ-quanta using all the available data on γ-radiation of 152 Eu, the measurement of the interval conversion coefficients (ICC) for the most intensive γ-transitions, the determination of the probabilities of the 152 Eu β-decays to the 152 Sm and 152 Gd levels. The conversion lines of the most intensive γ-transitions in the 152 Eu decay are studied and the corresponding ICC are measured on the beta-spectrometers of π√2 and UMB type. The balance for the γ-transitions in the 152 Sm and 152 Gd daughter nuclei are presented. This balance is used to determine the absolute intensities of γ-rays (in terms of the percentage of the 152 Eu decays) and the probabilities of β-transitions to the levels of daughter nuclei. More accurate data on γ-rays and conversion electrons obtained can be used for the calibration of gamma and beta spectrometers

  16. Maximum stellar iron core mass

    Indian Academy of Sciences (India)

    60, No. 3. — journal of. March 2003 physics pp. 415–422. Maximum stellar iron core mass. F W GIACOBBE. Chicago Research Center/American Air Liquide ... iron core compression due to the weight of non-ferrous matter overlying the iron cores within large .... thermal equilibrium velocities will tend to be non-relativistic.

  17. Maximum entropy beam diagnostic tomography

    International Nuclear Information System (INIS)

    Mottershead, C.T.

    1985-01-01

    This paper reviews the formalism of maximum entropy beam diagnostic tomography as applied to the Fusion Materials Irradiation Test (FMIT) prototype accelerator. The same formalism has also been used with streak camera data to produce an ultrahigh speed movie of the beam profile of the Experimental Test Accelerator (ETA) at Livermore. 11 refs., 4 figs

  18. Maximum entropy beam diagnostic tomography

    International Nuclear Information System (INIS)

    Mottershead, C.T.

    1985-01-01

    This paper reviews the formalism of maximum entropy beam diagnostic tomography as applied to the Fusion Materials Irradiation Test (FMIT) prototype accelerator. The same formalism has also been used with streak camera data to produce an ultrahigh speed movie of the beam profile of the Experimental Test Accelerator (ETA) at Livermore

  19. A portable storage maximum thermometer

    International Nuclear Information System (INIS)

    Fayart, Gerard.

    1976-01-01

    A clinical thermometer storing the voltage corresponding to the maximum temperature in an analog memory is described. End of the measurement is shown by a lamp switch out. The measurement time is shortened by means of a low thermal inertia platinum probe. This portable thermometer is fitted with cell test and calibration system [fr

  20. Fine Structure of 211 Po Alpha Decay

    International Nuclear Information System (INIS)

    Mirea, M.

    2000-01-01

    Recently, a theory based on the Landau-Zener effect was developed intending to describe quantitatively the cluster decay fine structure phenomenon. It was claimed that the same promotion effect can also govern the fine structure in the case of α-decay. This formalism intends to explain the fine structure of α-decay by considering single-particle transitions due to the radial and the rotational couplings. The levels with the same good quantum numbers associated to some symmetries of the system cannot in general intersect, but exhibit quasi-crossings, or pseudo-crossings, or avoided level crossings. The system is characterised by an axial symmetry, therefore the good quantum numbers are the projections of the nucleon spin Ω. The radial coupling causes transitions of the unpaired nucleon near the avoided level crossings. True crossings can also be obtained between levels characterized by different quantum numbers. Generally, the rotational coupling has a maximum strength in the vicinity of the true crossings. Transitions due to both couplings are taken into account in order to explain the excitations of the unpaired nucleon. For a tunnelling velocity of 9 x 10 6 fm/fs, the ratio between the intensity for transitions to the first excited state and to the ground state was found to be 0.0071 and the obtained ratio of the same parameter between the second excited state and the ground state was 0.0062, in good agreement with experimental data. These calculations suggest that the α-decay fine structure phenomenon can be explained quantitatively by describing the decaying system with molecular models and it can be stated that the quantitative characteristics of this phenomenon are ruled by dynamical effects. (author)

  1. Neutron spectra unfolding with maximum entropy and maximum likelihood

    International Nuclear Information System (INIS)

    Itoh, Shikoh; Tsunoda, Toshiharu

    1989-01-01

    A new unfolding theory has been established on the basis of the maximum entropy principle and the maximum likelihood method. This theory correctly embodies the Poisson statistics of neutron detection, and always brings a positive solution over the whole energy range. Moreover, the theory unifies both problems of overdetermined and of underdetermined. For the latter, the ambiguity in assigning a prior probability, i.e. the initial guess in the Bayesian sense, has become extinct by virtue of the principle. An approximate expression of the covariance matrix for the resultant spectra is also presented. An efficient algorithm to solve the nonlinear system, which appears in the present study, has been established. Results of computer simulation showed the effectiveness of the present theory. (author)

  2. RADIATIVE PENGUIN DECAYS FROM BABAR

    Energy Technology Data Exchange (ETDEWEB)

    Eigen, Gerald

    2003-08-28

    Electroweak penguin decays provide a promising hunting ground for Physics beyond the Standard Model (SM). The decay B {yields} X{sub s}{gamma}, which proceeds through an electromagnetic penguin loop, already provides stringent constraints on the supersymmetric (SUSY) parameter space. The present data samples of {approx}1 x 10{sup 8} B{bar B} events allow to explore radiative penguin decays with branching fractions of the order of 10{sup -6} or less. In this brief report they discuss a study of B {yields} K*{ell}{sup +}{ell}{sup -} decay modes and a search for B {yields} {rho}({omega}){gamma} decays.

  3. Shannon entropy and particle decays

    Science.gov (United States)

    Carrasco Millán, Pedro; García-Ferrero, M. Ángeles; Llanes-Estrada, Felipe J.; Porras Riojano, Ana; Sánchez García, Esteban M.

    2018-05-01

    We deploy Shannon's information entropy to the distribution of branching fractions in a particle decay. This serves to quantify how important a given new reported decay channel is, from the point of view of the information that it adds to the already known ones. Because the entropy is additive, one can subdivide the set of channels and discuss, for example, how much information the discovery of a new decay branching would add; or subdivide the decay distribution down to the level of individual quantum states (which can be quickly counted by the phase space). We illustrate the concept with some examples of experimentally known particle decay distributions.

  4. Column: Factors Affecting Data Decay

    Directory of Open Access Journals (Sweden)

    Kevin Fairbanks

    2012-06-01

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

  5. Rare B decays at LHCb

    CERN Document Server

    Puig Navarro, Albert

    2017-01-01

    Rare decays are flavour changing neutral current processes that allow sensitive searches for phenomena beyond the Standard Model (SM). In the SM, rare decays are loop-suppressed and new particles in SM extensions can give significant contributions. The very rare decay $B^0_s\\to\\mu^+\\mu^-$ in addition helicity suppressed and constitutes a powerful probe for new (pseudo) scalar particles. Of particular interest are furthermore tests of lepton universality in rare $b\\to s\\ell^+\\ell^-$ decays. The LHCb experiment is designed for the study of b-hadron decays and ideally suited for the analysis of rare decays due to its high trigger efficiency, as well as excellent tracking and particle identification performance. Recent results from the LHCb experiment in the area of rare decays are presented, including tests of lepton universality and searches for lepton flavour violation.

  6. B decays to open charm

    CERN Document Server

    AUTHOR|(CDS)2073670

    2016-01-01

    Studies of $B$ meson decays to states involving open charm mesons in data recorded by the LHCb experiment have resulted in first observations of several new decay modes, including $B_s^{0} \\rightarrow D_s^{*\\mp} K^{\\pm}$, $B_s^{0} \\rightarrow \\overline{D}^{0} K_S^{0}$ and $B^{+} \\rightarrow D^{+} K^{+} \\pi^{-}$ decays. An upper limit has been placed on the branching fraction of $B_s^{0} \\rightarrow \\overline{D}^{0} f_0(980)$ decays. Measurements of other branching fractions, such as those of $B_s^{0} \\rightarrow D_s^{(*)+} D_s^{(*)-}$ decays, are the most precise to date. Additionally, amplitude analyses of $B^{0} \\rightarrow \\overline{D}^{0} \\pi^{+} \\pi^{-}$ and $B^{0} \\rightarrow \\overline{D}^{0} K^{+} \\pi^{-}$ decays have been performed, alongside the first $CP$ violation analysis using the Dalitz plot of $B^{0} \\rightarrow D K^{+} \\pi^{-}$ decays.

  7. Maximum permissible concentration (MPC) values for spontaneously fissioning radionuclides

    International Nuclear Information System (INIS)

    Ford, M.R.; Snyder, W.S.; Dillman, L.T.; Watson, S.B.

    1976-01-01

    The radiation hazards involved in handling certain of the transuranic nuclides that exhibit spontaneous fission as a mode of decay were reaccessed using recent advances in dosimetry and metabolic modeling. Maximum permissible concentration (MPC) values in air and water for occupational exposure (168 hr/week) were calculated for 244 Pu, 246 Cm, 248 Cm, 250 Cf, 252 Cf, 254 Cf, /sup 254m/Es, 255 Es, 254 Fm, and 256 Fm. The half-lives, branching ratios, and principal modes of decay of the parent-daughter members down to a member that makes a negligible contribution to the dose are given, and all daughters that make a significant contribution to the dose to body organs following inhalation or ingestion are included in the calculations. Dose commitments for body organs are also given

  8. Inter-decadal modulation of ENSO teleconnections to the Indian Ocean in a coupled model: Special emphasis on decay phase of El Niño

    Science.gov (United States)

    Chowdary, J. S.; Parekh, Anant; Gnanaseelan, C.; Sreenivas, P.

    2014-01-01

    Inter-decadal modulation of El Niño-Southern Oscillation (ENSO) teleconnections to tropical Indian Ocean (TIO) is investigated in the coupled general circulation model Climate Forecast System (CFS) using a hundred year integration. The model is able to capture the periodicity of El Niño variability, which is similar to that of the observations. The maximum TIO/north Indian Ocean (NIO) SST warming (during spring following the decay phase of El Niño) associated with El Niño is well captured by the model. Detailed analysis reveals that the surface heat flux variations mainly contribute to the El Niño forced TIO SST variations both in observations and model. However, spring warming is nearly stationary throughout the model integration period, indicating poor inter-decadal El Niño teleconnections. The observations on the other hand displayed maximum SST warming with strong seasonality from epoch to epoch. The model El Niño decay delayed by more than two seasons, results in persistent TIO/NIO SST warming through the following December unlike in the observations. The ocean wave adjustments and persistent westerly wind anomalies over the equatorial Pacific are responsible for late decay of El Niño in the model. Consistent late decay of El Niño, throughout the model integration period (low variance), is mainly responsible for the poor inter-decadal ENSO teleconnections to TIO/NIO. This study deciphers that the model needs to produce El Niño decay phase variability correctly to obtain decadal-modulations in ENSO teleconnection.

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

    KAUST Repository

    Said-Houari, Belkacem

    2013-02-01

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

  10. On Maximum Entropy and Inference

    Directory of Open Access Journals (Sweden)

    Luigi Gresele

    2017-11-01

    Full Text Available Maximum entropy is a powerful concept that entails a sharp separation between relevant and irrelevant variables. It is typically invoked in inference, once an assumption is made on what the relevant variables are, in order to estimate a model from data, that affords predictions on all other (dependent variables. Conversely, maximum entropy can be invoked to retrieve the relevant variables (sufficient statistics directly from the data, once a model is identified by Bayesian model selection. We explore this approach in the case of spin models with interactions of arbitrary order, and we discuss how relevant interactions can be inferred. In this perspective, the dimensionality of the inference problem is not set by the number of parameters in the model, but by the frequency distribution of the data. We illustrate the method showing its ability to recover the correct model in a few prototype cases and discuss its application on a real dataset.

  11. Maximum Water Hammer Sensitivity Analysis

    OpenAIRE

    Jalil Emadi; Abbas Solemani

    2011-01-01

    Pressure waves and Water Hammer occur in a pumping system when valves are closed or opened suddenly or in the case of sudden failure of pumps. Determination of maximum water hammer is considered one of the most important technical and economical items of which engineers and designers of pumping stations and conveyance pipelines should take care. Hammer Software is a recent application used to simulate water hammer. The present study focuses on determining significance of ...

  12. Maximum Gene-Support Tree

    Directory of Open Access Journals (Sweden)

    Yunfeng Shan

    2008-01-01

    Full Text Available Genomes and genes diversify during evolution; however, it is unclear to what extent genes still retain the relationship among species. Model species for molecular phylogenetic studies include yeasts and viruses whose genomes were sequenced as well as plants that have the fossil-supported true phylogenetic trees available. In this study, we generated single gene trees of seven yeast species as well as single gene trees of nine baculovirus species using all the orthologous genes among the species compared. Homologous genes among seven known plants were used for validation of the finding. Four algorithms—maximum parsimony (MP, minimum evolution (ME, maximum likelihood (ML, and neighbor-joining (NJ—were used. Trees were reconstructed before and after weighting the DNA and protein sequence lengths among genes. Rarely a gene can always generate the “true tree” by all the four algorithms. However, the most frequent gene tree, termed “maximum gene-support tree” (MGS tree, or WMGS tree for the weighted one, in yeasts, baculoviruses, or plants was consistently found to be the “true tree” among the species. The results provide insights into the overall degree of divergence of orthologous genes of the genomes analyzed and suggest the following: 1 The true tree relationship among the species studied is still maintained by the largest group of orthologous genes; 2 There are usually more orthologous genes with higher similarities between genetically closer species than between genetically more distant ones; and 3 The maximum gene-support tree reflects the phylogenetic relationship among species in comparison.

  13. Maximum nondiffracting propagation distance of aperture-truncated Airy beams

    Science.gov (United States)

    Chu, Xingchun; Zhao, Shanghong; Fang, Yingwu

    2018-05-01

    Airy beams have called attention of many researchers due to their non-diffracting, self-healing and transverse accelerating properties. A key issue in research of Airy beams and its applications is how to evaluate their nondiffracting propagation distance. In this paper, the critical transverse extent of physically realizable Airy beams is analyzed under the local spatial frequency methodology. The maximum nondiffracting propagation distance of aperture-truncated Airy beams is formulated and analyzed based on their local spatial frequency. The validity of the formula is verified by comparing the maximum nondiffracting propagation distance of an aperture-truncated ideal Airy beam, aperture-truncated exponentially decaying Airy beam and exponentially decaying Airy beam. Results show that the formula can be used to evaluate accurately the maximum nondiffracting propagation distance of an aperture-truncated ideal Airy beam. Therefore, it can guide us to select appropriate parameters to generate Airy beams with long nondiffracting propagation distance that have potential application in the fields of laser weapons or optical communications.

  14. LCLS Maximum Credible Beam Power

    International Nuclear Information System (INIS)

    Clendenin, J.

    2005-01-01

    The maximum credible beam power is defined as the highest credible average beam power that the accelerator can deliver to the point in question, given the laws of physics, the beam line design, and assuming all protection devices have failed. For a new accelerator project, the official maximum credible beam power is determined by project staff in consultation with the Radiation Physics Department, after examining the arguments and evidence presented by the appropriate accelerator physicist(s) and beam line engineers. The definitive parameter becomes part of the project's safety envelope. This technical note will first review the studies that were done for the Gun Test Facility (GTF) at SSRL, where a photoinjector similar to the one proposed for the LCLS is being tested. In Section 3 the maximum charge out of the gun for a single rf pulse is calculated. In Section 4, PARMELA simulations are used to track the beam from the gun to the end of the photoinjector. Finally in Section 5 the beam through the matching section and injected into Linac-1 is discussed

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

    Indian Academy of Sciences (India)

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

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

    NARCIS (Netherlands)

    Mapelli, M.; Ripamonti, E.

    2007-01-01

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

  17. Decay of 36K

    International Nuclear Information System (INIS)

    Fritts, M.J.

    1976-01-01

    36 K was produced via the 36 Ar(p, n) 36 K reaction. Measurement of 27 β + delayed γ rays associated with the decay of 36 K implied 10 new β + branches to energy levels in 36 Ar. Branching ratios and logft values are calculated for the β + branches. Restrictions on spin and parity assignments for the 36 Ar levels are given, as well as branching ratios for γ transitions from these levels. The half-life of 36 K is determined to be 344 +- 3 msec

  18. η decays at Saclay

    International Nuclear Information System (INIS)

    Mayer, B.

    1991-01-01

    A facility dedicated to the production of η mesons has been installed at the Saturne synchrotron with the purpose of investigating rare decays of this meson. The η are produced by the pd → 3 Heη reaction near threshold and tagged by the detection of 3 He in a magnetic spectrometer (SPES2). A rate of 10 5 /s tagged η can be achieved. In the first experiment, η → μ + μ - , the μ will be detected in range telescopes. Magnetic spectrometers for lepton detection are considered for future experiments

  19. FTR europia gamma heating

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1975-01-01

    Calculated and experimental gamma heating rates of europia in the Engineering Mockup Critical Assembly (EMC) were correlated. A calculated to experimental (C/E) ratio of 1.086 was established in validating the theoretical approach and computational technique applied in the calculations. Gamma heat deposition rates in the FTR with Eu 2 O 3 control absorbers were determined from three-dimensional calculations. Maximum gamma heating was found to occur near the tip of a half-inserted row 5 control rod assembly--12.8 watts/gm of europia. Gamma heating profiles were established for a single half-inserted europia absorber assembly. Local heat peaking was found not to alter significantly heating rates computed in the FTR core model, where larger mesh interval sizes precluded examination of spatially-limited heating gradients. These computations provide the basis for thermal-hydraulic analyses to ascertain temperature profiles in the FTR under europia control

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-01-01

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

  2. Rare decays and CP asymmetries in charged B decays

    International Nuclear Information System (INIS)

    Deshpande, N.G.

    1991-01-01

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

  3. Study of dryout heat fluxes in beds of inductively heated particles

    International Nuclear Information System (INIS)

    Dhir, V.K.; Catton, I.

    1977-02-01

    Experimental observations of the dryout heat fluxes for inductively heated particulate beds have been made. The data were obtained when steel and lead particles in the size distribution 295-787 microns were placed in a 4.7 cm diameter pyrex glass jar and inductively heated by passing radio frequency current through a 13.3 cm diameter multi-turn work coil encircling the jar. Distilled water, methanol and acetone were used as coolants in the experiments, while the bed height was varied from 1.0 to 8.9 cm. Different mechanisms for the dryout in deep and shallow beds have been identified. Dryout in shallow beds is believed to occur when the vapor velocity in the gas jets exceeds a certain critical velocity at which choking of the vapor occurs, leading to obstruction in the flow of the liquid toward the bed. However, deep beds dry out when gravitational force can no longer maintain a downward coolant flow rate necessary to dissipate the heat generated in the bed. The heat flux data of the investigation and that from two previous investigations made at Argonne Laboratory and at UCLA have been correlated with semi-theoretical correlations based on the proposed hydrodynamic models. The deep and shallow bed correlations are used to predict the bed height at which transition from deep to shallow bed would occur. An application of the study has been made to determine the maximum coolable depths of the core debris as a function of the particle size, bed porosity and decay heat

  4. Pulsed radiation decay logging

    International Nuclear Information System (INIS)

    Mills, W.R. Jr.

    1983-01-01

    There are provided new and improved well logging processes and systems wherein the detection of secondary radiation is accomplished during a plurality of time windows in a manner to accurately characterize the decay rate of the secondary radiation. The system comprises a well logging tool having a primary pulsed radiation source which emits repetitive time-spaced bursts of primary radiation and detector means for detecting secondary radiation resulting from the primary radiation and producing output signals in response to the detected radiation. A plurality of measuring channels are provided, each of which produces a count rate function representative of signals received from the detector means during successive time windows occurring between the primary radiation bursts. The logging system further comprises means responsive to the measuring channels for producing a plurality of functions representative of the ratios of the radiation count rates measured during adjacent pairs of the time windows. Comparator means function to compare the ratio functions and select at least one of the ratio functions to generate a signal representative of the decay rate of the secondary radiation

  5. Design and optimization of automotive thermoelectric generators for maximum fuel efficiency improvement

    International Nuclear Information System (INIS)

    Kempf, Nicholas; Zhang, Yanliang

    2016-01-01

    Highlights: • A three-dimensional automotive thermoelectric generator (TEG) model is developed. • Heat exchanger design and TEG configuration are optimized for maximum fuel efficiency increase. • Heat exchanger conductivity has a strong influence on maximum fuel efficiency increase. • TEG aspect ratio and fin height increase with heat exchanger thermal conductivity. • A 2.5% fuel efficiency increase is attainable with nanostructured half-Heusler modules. - Abstract: Automotive fuel efficiency can be increased by thermoelectric power generation using exhaust waste heat. A high-temperature thermoelectric generator (TEG) that converts engine exhaust waste heat into electricity is simulated based on a light-duty passenger vehicle with a 4-cylinder gasoline engine. Strategies to optimize TEG configuration and heat exchanger design for maximum fuel efficiency improvement are provided. Through comparison of stainless steel and silicon carbide heat exchangers, it is found that both the optimal TEG design and the maximum fuel efficiency increase are highly dependent on the thermal conductivity of the heat exchanger material. Significantly higher fuel efficiency increase can be obtained using silicon carbide heat exchangers at taller fins and a longer TEG along the exhaust flow direction when compared to stainless steel heat exchangers. Accounting for major parasitic losses, a maximum fuel efficiency increase of 2.5% is achievable using newly developed nanostructured bulk half-Heusler thermoelectric modules.

  6. Physical properties and heat transfer characteristics of materials for krypton-85 storage

    International Nuclear Information System (INIS)

    Christensen, A.B.

    1977-09-01

    Krypton-85 decay results in heat generation, and the subsequent temperature increase in the krypton-85 storage media must be evaluated. This report compiles the physical properties of krypton and of potential krypton-85 storage materials which are required to calculate the maximum temperature developed during storage. Temperature calculations were made for krypton-85 stored as a gas or immobilized solid in steel storage cylinders. The effects of krypton-85 loading, cylinder radius, storage media properties, and exterior cooling on storage temperature were shown

  7. Determination of maximum reactor power level consistent with the requirement that flow reversal occurs without fuel damage

    International Nuclear Information System (INIS)

    Rao, D.V.; Darby, J.L.; Ross, S.B.; Clark, R.A.

    1990-01-01

    The High Flux Beam Reactor (HFBR) operated by Brookhaven National Laboratory (BNL) employs forced downflow for heat removal during normal operation. In the event of total loss of forced flow, the reactor will shutdown and the flow reversal valves open. When the downward core flow becomes sufficiently small then the opposing thermal buoyancy induces flow reversal leading to decay heat removal by natural convection. There is some uncertainty as to whether the natural circulation is adequate for decay heat removal after 60 MW operation. BNL- staff carried out a series of calculations to establish the adequacy of flow reversal to remove decay heat. Their calculations are based on a natural convective CHF model. The primary purpose of the present calculations is to review the accuracy and applicability of Fauske's CHF model for the HFBR, and the assumptions and methodology employed by BNL-staff to determine the heat removal limit in the HFBR during a flow reversal and natural convection situation

  8. Generic maximum likely scale selection

    DEFF Research Database (Denmark)

    Pedersen, Kim Steenstrup; Loog, Marco; Markussen, Bo

    2007-01-01

    in this work is on applying this selection principle under a Brownian image model. This image model provides a simple scale invariant prior for natural images and we provide illustrative examples of the behavior of our scale estimation on such images. In these illustrative examples, estimation is based......The fundamental problem of local scale selection is addressed by means of a novel principle, which is based on maximum likelihood estimation. The principle is generally applicable to a broad variety of image models and descriptors, and provides a generic scale estimation methodology. The focus...

  9. Heat pumps: heat recovery

    Energy Technology Data Exchange (ETDEWEB)

    Pielke, R

    1976-01-01

    The author firstly explains in a general manner the functioning of the heat pump. Following a brief look at the future heat demand and the possibilities of covering it, the various methods of obtaining energy (making use of solar energy, ground heat, and others) and the practical applications (office heating, swimming pool heating etc.) are explained. The author still sees considerable difficulties in using the heat pump at present on a large scale. Firstly there is not enough maintenance personnel available, secondly the electricity supply undertakings cannot provide the necessary electricity on a wide basis without considerable investments. Other possibilities to save energy or to use waste energy are at present easier and more economical to realize. Recuperative and regenerative systems are described.

  10. Search for proton decay: introduction

    International Nuclear Information System (INIS)

    Goldhaber, M.

    1984-01-01

    In interpreting contained events observed in various proton decay detectors one can sometimes postulate, though usually not unambiguously, a potential decay mode of the proton, called a candidate. It is called a candidate, because for any individual event it is not possible to exclude the possibility that it is instead due to cosmic ray background, chiefly atmospheric neutrinos. Some consistency checks are proposed which could help establish proton decay, if it does occur in the presently accessible lifetime window

  11. Rare beauty and charm decays

    International Nuclear Information System (INIS)

    Blake, T.

    2016-01-01

    Rare beauty and charm decays can provide powerful probes of physics beyond the Standard Model. These proceedings summarise the latest measurements of rare beauty and charm decays from the LHCb experiment at the end of Run 1 of the LHC. Whilst the majority of the measurements are consistent with SM predictions, small differences are seen in the rate and angular distribution of b → sℓ"+ℓ"− decay processes.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  13. Weak decays of heavy quarks

    International Nuclear Information System (INIS)

    Gaillard, M.K.

    1978-08-01

    The properties that may help to identify the two additional quark flavors that are expected to be discovered. These properties are lifetime, branching ratios, selection rules, and lepton decay spectra. It is also noted that CP violation may manifest itself more strongly in heavy particle decays than elsewhere providing a new probe of its origin. The theoretical progress in the understanding of nonleptonic transitions among lighter quarks, nonleptonic K and hyperon decay amplitudes, omega minus and charmed particle decay predictions, and lastly the Kobayashi--Maskawa model for the weak coupling of heavy quarks together with the details of its implications for topology and bottomology are treated. 48 references

  14. Statistical decay of giant resonances

    International Nuclear Information System (INIS)

    Dias, H.; Teruya, N.; Wolynec, E.

    1986-01-01

    Statistical calculations to predict the neutron spectrum resulting from the decay of Giant Resonances are discussed. The dependence of the resutls on the optical potential parametrization and on the level density of the residual nucleus is assessed. A Hauser-Feshbach calculation is performed for the decay of the monople giant resonance in 208 Pb using the experimental levels of 207 Pb from a recent compilation. The calculated statistical decay is in excelent agreement with recent experimental data, showing that the decay of this resonance is dominantly statistical, as predicted by continuum RPA calculations. (Author) [pt

  15. Statistical decay of giant resonances

    International Nuclear Information System (INIS)

    Dias, H.; Teruya, N.; Wolynec, E.

    1986-02-01

    Statistical calculations to predict the neutron spectrum resulting from the decay of Giant Resonances are discussed. The dependence of the results on the optical potential parametrization and on the level density of the residual nucleus is assessed. A Hauser-Feshbach calculation is performed for the decay of the monopole giant resonance in 208 Pb using the experimental levels of 207 Pb from a recent compilation. The calculated statistical decay is in excellent agreement with recent experimental data, showing that decay of this resonance is dominantly statistical, as predicted by continuum RPA calculations. (Author) [pt

  16. Is Radioactive Decay Really Exponential?

    OpenAIRE

    Aston, Philip J.

    2012-01-01

    Radioactive decay of an unstable isotope is widely believed to be exponential. This view is supported by experiments on rapidly decaying isotopes but is more difficult to verify for slowly decaying isotopes. The decay of 14C can be calibrated over a period of 12,550 years by comparing radiocarbon dates with dates obtained from dendrochronology. It is well known that this approach shows that radiocarbon dates of over 3,000 years are in error, which is generally attributed to past variation in ...

  17. Extreme Maximum Land Surface Temperatures.

    Science.gov (United States)

    Garratt, J. R.

    1992-09-01

    There are numerous reports in the literature of observations of land surface temperatures. Some of these, almost all made in situ, reveal maximum values in the 50°-70°C range, with a few, made in desert regions, near 80°C. Consideration of a simplified form of the surface energy balance equation, utilizing likely upper values of absorbed shortwave flux (1000 W m2) and screen air temperature (55°C), that surface temperatures in the vicinity of 90°-100°C may occur for dry, darkish soils of low thermal conductivity (0.1-0.2 W m1 K1). Numerical simulations confirm this and suggest that temperature gradients in the first few centimeters of soil may reach 0.5°-1°C mm1 under these extreme conditions. The study bears upon the intrinsic interest of identifying extreme maximum temperatures and yields interesting information regarding the comfort zone of animals (including man).

  18. CP violation in K decays and rare decays

    International Nuclear Information System (INIS)

    Buchalla, G.

    1996-12-01

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

  19. HAGRID/ VANDLE spectroscopy of Rb decays

    Science.gov (United States)

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

    2017-09-01

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

  20. Magnetically induced vacuum decay

    International Nuclear Information System (INIS)

    Xue Shesheng

    2003-01-01

    We study the fermionic vacuum energy of vacua with and without application of an external magnetic field. The energetic difference of two vacua leads to the vacuum decaying and the vacuum energy being released. In the context of quantum field theories, we discuss why and how the vacuum energy can be released by spontaneous photon emission and/or paramagnetically screening the external magnetic field. In addition, we quantitatively compute the vacuum energy released, the paramagnetic screening effect, and the rate and spectrum of spontaneous photon emission. The possibilities of experimentally detecting such an effect of vacuum-energy release and that this effect accounts for the anomalous x-ray pulsar are discussed