WorldWideScience

Sample records for activation decay heat

  1. Decay heat calculations for reactors

    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. The method of decay heat estimation relies on the measurements over practical time intervals as well as on calculation for predictions over very long time intervals. Neutron cross-sections, fission yields and decay data together with operational history are the basic inputs to such. A code used to calculate decay heat would require to generate isotopic inventory that would be present at the shutdown based on operational history of the reactor and follow up the decay over an extended period of time. Aspects of decay heat estimation based on standards like ANS 5.1 and by fuel cycle analysis codes shall be discussed. 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)

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

    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. THIDA-2: an advanced code system for calculation of transmutation, activation, decay heat and dose rate

    In a D-T burning fusion reactor, the radioactivity induced by the 14 MeV neutrons causes many problems. It limits personnel access to the reactor during shutdown, generates decay heat and produces radwastes. A code system THIDA had been developed in 1978 to calculate the radioactivity and dose rate around a fusion device. The THIDA system consisted of the followings: one- and two-dimensional discrete ordinates radiation transport codes; induced activity calculation code; three libraries for transmutation and decay chain data, transmutation cross sections and delayed gamma-ray emission data. The present report gives a complete description of THIDA-2, a new advanced version of the THIDA system which has the following major improvements: 1. Capability to treat three-dimensional calculation models by the use of a Monte Carlo transport code. 2. Accurate decay heat calculation following the transport of delayed gamma rays. 3. Simplification of the data input process by the use of free format scheme and closer coupling between the radiation transport codes and the induced activity calculation code. 4. Self-descriptive output format and additional plotter output. 5. Capability to calculate problems requiring larger core memory by the use of variable dimension. (author)

  4. Decay Heat Code Validation Activities at ORNL: Supporting Expansion of NRC Regulatory Guide 3.54

    Oak Ridge National Laboratory (ORNL) has a long history of involvement in the development and validation of the ORIGEN series of isotope summation codes and nuclear data libraries, widely recognized and used to predict the decay heat for spent nuclear fuel. In particular, the ORIGEN-S code, the depletion/decay module of the SCALE code system, has been extensively validated using experimental isotopic assay data and decay heat measurements for commercial spent fuel. This work was used in the development of the technical basis for NRC Regulatory Guide 3.54 on spent fuel decay heat. The bulk of the experimental data used to validate spent fuel decay heat predictions are from programs of the 1970s and 1980s and consequently involve older-design fuel assemblies with a relatively low enrichment and burnup. This has led to a situation where the spent fuel now being discharged from operating reactors extends well beyond the regime of the experimental data and area of code applicability based on the data. The absence of validation data for modern fuel designs has potentially serious consequences for decay heat predictions in terms of added safety factors to account for larger uncertainties and lower volumetric transport and storage capacities

  5. Consolidated fuel decay heat calculations

    Wittekind, W.D.

    1994-06-24

    The radiological decay heat generated from all irradiated fuel presently in K East (KE) and K West (KW) Basins was calculated in support of consolidated fuel storage. There are four sources of heat inflow into the fuel storage basins: (1) radiological decay heat from irradiated fuel; (2) mechanical heat from operating machinery (e.g., pumps); (3) heat flow from surroundings (mainly the ground through the concrete walls into the basin water if it is maintained below ambient); and (4) exothermic chemical reactions of uranium oxidation (although at basin temperatures this reaction rate is slow). This report details the radiological decay heat from irradiated fuel source in the K basins. Decay heat calculations using ORIGEN2 (Wittekind 1994 and Schmittroth 1993) for irradiated fuel presently (April 1994) in KE and KW Basins gave results for January 31 of each year.

  6. Data for decay Heat Predictions

    These proceedings of a specialists' meeting on data for decay heat predictions are based on fission products yields, on delayed neutrons and on comparative evaluations on evaluated and experimental data for thermal and fast fission. Fourteen conferences were analysed

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

    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 calculation consistency with the experimental results, the knowledge of the material impurities content is mandatory

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

    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...

  9. Decay Heat Removal System of Monju

    MONJU has three decay heat removal systems. The intermediate heat exchanger of the decay heat removal system is incorporated within the main IHX shell, and the heat from the secondary system is rejected to the air. Forced circulation is adopted for both primary and secondary coolant, though natural circulation capability is designed into the plant itself. Feasibility of rejecting the decay heat through steam plant is also being studied. In this paper, MONJU's decay heat removal system design, operational procedures, and the considerations behind the concept will be presented. (author)

  10. Passive decay heat removal during shutdown

    During shutdown the decay heat in commercial Boiling Water Reactors is removed from the core region by active and redundant pump/heat exchanger-systems which are, in addition, supported by emergency power. To study the capability of the newly developed emergency condensers to remove energy produced within the core region to a large water pool outside the Reactor Pressure Vessel by natural convection, a related test in the NOKO facility as performed. The pressure vessel in the NOKO facility has been flooded above the inlet line to the emergency condenser and heated up to about 100 deg. C. The natural circulation resulted in a cool down of the water within the pressure vessel. With two specially designed grids equipped with up to 12 thermocouples the temperature fields in two cross sections were measured: no plume-effect was identified. The vertical temperature profile was measured with thermocouples. The test showed that decay heat could be removed some time after scram to an outside pool by natural convection processes: the time after scram depends on the emergency condenser heat exchange area. (author)

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

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

    2014-10-15

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

  12. Estimation of decay heat in fusion DEMO reactor

    The decay heat of activated materials is important in safety assessment of fusion DEMO reactor against loss of coolant-flow accidents. Decay heat for reactor main components of the SlimCS DEMO reactor was studied with a one-dimensional code THIDA-2. The reactor main components consist of the inboard (IB) blanket module, outboard (OB) blanket module and divertor. For a reactor with a fusion output of 3.0 GW, the decay heat of IB blanket, OB blanket, divertor and radiation shield were estimated to be as high as 8.6 MW, 30.9 MW, 10.6 MW and 1.8 MW, respectively, immediately after the shutdown of operation. The total decay heat was as high as 52 MW immediately after the shutdown and 3.1 MW one month later. The blanket produces the largest portion of decay heat, about 76%. (author)

  13. Safety characteristics of decay heat removal systems

    Safety features of the decay heat removal systems including power sunply and final heat sink are described. A rather high reliability and an utmost degree of independence from energy supply are goals to be attained in the design of the European Fast Reactor (EFR) decay heat removal scheme. Natural circulation is an ambitious design goal for EFR. All the considerations are performed within the frame of risk minimization

  14. Development of limiting decay heat values

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

  15. Heat Kernels, Smoothness Estimates and Exponential Decay

    Boggess, Albert; Raich, Andrew

    2010-01-01

    In this article, we establish Gaussian decay for the Box_b-heat kernel on polynomial models in C^2. Our technique attains the exponential decay via a partial Fourier transform. On the transform side, the problem becomes finding quantitative smoothness estimates on a heat kernel associated to the weighted dbar-operator on L^2(C). The bounds are established with Duhamel's formula and careful estimation.

  16. Decay heat studies for nuclear energy

    Algora, A.; Jordan, D.; Taín, J. L.; Rubio, B.; Agramunt, J.; Caballero, L.; Nácher, E.; Perez-Cerdan, A. B.; Molina, F.; Estevez, E.; Valencia, E.; Krasznahorkay, A.; Hunyadi, M. D.; Gulyás, J.; Vitéz, A.; Csatlós, M.; Csige, L.; Eronen, T.; Rissanen, J.; Saastamoinen, A.; Moore, I. D.; Penttilä, H.; Kolhinen, V. S.; Burkard, K.; Hüller, W.; Batist, L.; Gelletly, W.; Nichols, A. L.; Yoshida, T.; Sonzogni, A. A.; Peräjärvi, K.

    2014-01-01

    The energy associated with the decay of fission products plays an important role in the estimation of the amount of heat released by nuclear fuel in reactors. In this article we present results of the study of the beta decay of some refractory isotopes that were considered important contributors to the decay heat in reactors. The measurements were performed at the IGISOL facility of the University of Jyväskylä, Finland. In these studies we have combined for the first time a Penning trap (JYFLTRAP), which was used as a high resolution isobaric separator, with a total absorption spectrometer. The results of the measurements as well as their consequences for decay heat summation calculations are discussed.

  17. Decay heat of fast reactor spent fuel

    Decay heat from JOYO Mk-II spent fuel subassemblies was measured to obtain experimental data and to improve the accuracy of related calculations. The measurement was taken in the JOYO spent fuel storage pond. The fuel burn-up was approximately 66 GWd/t and the cooling time was between 40 and 385 days. The decay heat was calculated with the ORIGEN2 code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. The fuel power used as an input to ORIGEN2 was calculated by the MAGI core management code system. The ratios between calculated and experimental values were between 0.94 and 0.89 and decreased with a longer cooling time. This systematic discrepancy is not fully understood, but the change with cooling time appears to be due to the actinide decay heat uncertainty. This indicated that cross sections of actinides are important to evaluate decay heat accurately. (author)

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

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

  19. Fission product decay heat for thermal reactors

    Dickens, J. K.

    1979-01-01

    In the past five years there have been new experimental programs to measure decay heat (i.e., time dependent beta- plus gamma-ray energy release rates from the decay of fission products) following thermal-neutron fission of /sup 235/U, /sup 239/Pu, and /sup 241/Pu for times after fission between 1 and approx. 10/sup 5/ sec. Experimental results from the ORNL program stress the very short times following fission, particularly in the first few hundred sec. Complementing the experimental effort, computer codes have been developed for the computation of decay heat by summation of calculated individual energies released by each one of the fission products. By suitably combining the results of the summation calculations with the recent experimental results, a new Decay Heat Standard has been developed for application to safety analysis of operations of light water reactors. The new standard indicates somewhat smaller energy release rates than those being used at present, and the overall uncertainties assigned to the new standard are much smaller than those being used at present.

  20. Applications of TAGS data in beta decay energies and decay heat calculations

    Pham, N. S.; 片倉 純一

    2007-01-01

    The recent data of beta-decay intensity measured by using the total absorption gamma-ray spectrometer (TAGS), for several fission products (FP), has been applied for calculations of the average energies and spectra, and decay heat summations. The calculations were performed based on the Gross theory of beta decay, in which the beta strength functions were experimentally derived from TAGS data. The deviations of decay heat power predictions from the original decay data of JENDL Decay Data File...

  1. Jeff-3 and decay heat calculations

    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 UO2 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

  2. Study on diverse passive decay heat removal approach

    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)

  3. Experimental validation of decay heat calculations for ITER

    This article reports an international benchmark exercise to determine nuclear data and activation code uncertainties in the context of decay heat projections for ITER, the subject of a workshop, held a the ITER San Diego, Joint Work Site, October 2-3, 1997, under a Task Agreement between JCT and the Japanese Home Team. In view of the importance of the new results, JCT opened this meeting to other Home Teams to allow interested specialists to participate if they wish

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

    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)

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

    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)

  6. Experimental Investigation on Decay Heat Removal System of CEFR

    2001-01-01

    The decay heat exists in a relative long time period after reactor shut down. The decay heat is removed by the decay heat removal system (DHRS).The manager of CEFR demanded that experiments should be carried out to make sure that natural circulation could be established under the conditions of < 1.0 percent of the normal power. Experiments are therefore performed under the heating power of 2, 3, 4, 6 and 8.5 kW respectively. The measured temperature in the hot plenum will be used to validate the computer codes. Moreover,

  7. Consistency among integral measurements of aggregate decay heat power

    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)

  8. Decay heat calculation an international nuclear code comparison

    The results of an international code comparison on decay heat are presented and discussed. Participants from more than ten laboratories calculated, using the same input data, decay heat for thirteen cooling times between 1 and 1013 sec. Two irradiation cases were proposed: fission pulse and 3x107 seconds of irradiation of 235U fuel. The results are analysed and compared. This inter-comparison shows that, if the same input data are given, most of the codes give very similar results for the decay heat and consequently also for the fission product contribution

  9. LWR decay heat calculations using a GRS improved ENDF/B-6 based ORIGEN data library

    The known ORNL ORIGEN code is widely spread over the world for inventory, activity and decay heat tasks and is used stand-alone or implemented in activation, shielding or burn-up systems. More than 1000 isotopes with more than six coupled neutron capture and radioactive decay channels are handled simultaneously by the code. The characteristics of the calculated inventories, e.g., masses, activities, neutron and photon source terms or the decay heat during short or long decay time steps are achieved by summing over all isotopes, characterized in the ORIGEN libraries. An extended nuclear GRS-ORIGENX data library is now developed for practical appliance. The library was checked for activation tasks of structure material isotopes and for actinide and fission product burn-up calculations compared with experiments and standard methods. The paper is directed to the LWR decay heat calculation features of the new library and shows the differences of dynamical and time integrated results of Endf/B-6 based and older Endf/B-5 based libraries for decay heat tasks compared to fission burst experiments, ANS curves and some other published data. A multi-group time exponential evaluation is given for the fission burst power of 235U, 238U, 239Pu and 241Pu, to be used in quick LWR reactor accident decay heat calculation tools. (authors)

  10. Decay heat measurement of fusion related materials in an ITER-like neutron field

    Morimoto, Y.; Ochiai, K.; Maekawa, F.; Wada, M.; Nishitani, T.; Takeuchi, H.

    2002-12-01

    Decay heat is one of the most important factors for the safety aspect of ITER. Especially, the prediction of decay heat with an uncertainty less than 15% for the three most important materials, i.e., copper, type-316 stainless steel (SS316) and tungsten, is strongly requested by designers of ITER. To provide experimental decay heat data needed for validation of decay heat calculations for SS316 and copper, an experiment was conducted as the ITER/EDA task T-426. An ITER-like neutron field was constructed, and decay heat source distributions in thick copper and SS316 plates were measured with the whole energy absorption spectrometer. The measured decay heat distributions in the thick sample plates were compared with the predicted values by MCNP calculations. It was found that the use of an effective activation cross-section calculated by MCNP was needed to consider the self-shielding effects and, for both cases, MCNP calculations could predict the decay heat adequately.

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

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

    2005-09-01

    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 850ºC 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).

  12. Proliferation resistance of plutonium based on decay heat

    Proliferation resistance of plutonium can be enhanced by increasing the decay heat of plutonium. For example, it can be enhanced by increasing the isotopic fraction of 238Pu, which has the largest decay heat among plutonium isotopes, produced by transmutation of Minor Actinides (Protected Plutonium Production: P3). In the present paper, proliferation resistance of plutonium was evaluated based on decay heat with physical assessment model. As a summary of the evaluation, new criteria to evaluate proliferation resistance of plutonium based on its isotopic composition from the view point of decay heat were suggested. The present methodology and the criteria were applied to evaluate the impact of P3 by the transmutation of Minor Actinides in fast breeder reactor blanket on proliferation resistance of plutonium. (author)

  13. Method for removal of decay heat of radioactive substances

    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.)

  14. Decay heat analysis of MNSR reactor core using ORIGEN-2 code

    Highlights: • Analysis of the decay heat parameters of the MNSR reactor was performed using the ORIGEN-2 code. • A new one-group cross-section data base of the ORlGEN-2 computer code for the MNSR was developed. • It is recommended to adopt the results of the ORIGIEN-2 code for future reactor safety analysis of MNSR. - Abstract: The knowledge of the decay heat of nuclear fuel is necessary for performing the reactor safety analysis, determining the heating load in fuel pools, shielding requirements on fuel discharge and transport routes when irradiated reactor fuel is transferred from the reactor, via some intermediate storage location, to the final disposal or the chemical reprocessing plant. In this study, analysis of the decay heat parameters of the Miniature Neutron Source Reactor (MNSR) including radioactivity, decay heat and the isotopic mass variation with time since reactor shutdown for the potential Low Enriched Uranium (LEU) (UO2-Zircaloy and U3Si-Al) and the standard Highly Enriched Uranium (HEU) (HEU-Al4) cores has been performed using the ORIGEN-2 code. For this purpose, a new one-group cross-section data base of the ORlGEN-2 computer code for the MNSR with LEU and HEU fuels has been developed using the MCNP-4C code. The variation of fission products, actinides and daughters and activation products with post shutdown time for the standard core and the potential LEU cores have been considered in the analysis of the decay heat power resources. It was found that, all the three types of MNSR fuels show close agreement in the total decay heat, which is mainly due to the fission products. This behavior continued for about 1.0E05 days. After this time, the fission products decay heat became comparable with the corresponding actinides decay heat in which the standard HEU UAl4-Al fuel showed the smallest decay heat values while the potential LEU-UO2 fuel had the highest decay heat followed by the LEU-U3Si fuel. The time variation of the total radioactivity

  15. Development of passive decay heat removal system PDRC

    One of the most important tasks in a successful design of a nuclear power plant is the demonstration of safe and reliable decay heat removal (DHR). In order to assure a plant safety, a safety grade DHR system, PDRC (passive decay heat removal circuit), has been developed to cope with the design basis events in KALIMER 600. The PDRX employs an innovative design concept distinct from conventional sodium cooled fast reactor designs currently being developed or operated in the world. An operational reliance of the PDRC system can be remarkably enhanced by excluding either any operator's action or active components operated by an external power supply or actuation signal. Based on the established design concept, system design methodologies have been developed and its performance analyses were reasonably carried out. Also, for the verification of the advanced design concept, a large scale sodium thermal hydraulic test facility is being designed. Starting with the basic design of the test facility in 2008, its installation is scheduled to be completed by the end of 2011

  16. Uncertainties in fission-product decay-heat calculations

    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)

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

    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)

  18. Performance of ALMR passive decay heat removal system

    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)

  19. Optimization design of emergency decay heat removal system for CEFR

    Background: China experimental fast reactor (CEFR) is the first fast reactor in China, the piping designs are very important for the fast reactor. Purpose: The purpose is to improve the economy of pipeline design for the fast reactor. Methods: According to the optimization principles, redesign the support system of emergency decay heat removal system. Results: The piping of emergency decay heat removal system meets the ASME standard stress limit value under various prospective loads, and reduces the number of dampers, spring hangers. Conclusions: We accumulate layout experience for the support design of high temperature piping by the optimization design for emergency decay heat removal system, lay the foundation for the fast reactor piping design in the future. (authors)

  20. Tests for removal of decay heat by natural convection

    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)

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

    Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

    1998-03-01

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

  2. Decay Heat Measurements Using Total Absorption Gamma-ray Spectroscopy

    Rice, S.; Valencia, E.; Algora, A.; Taín, J. L.; Regan, P. H.; Podolyák, Z.; Agramunt, J.; Gelletly, W.; Nichols, A. L.

    2012-09-01

    A knowledge of the decay heat emitted by thermal neutron-irradiated nuclear fuel is an important factor in ensuring safe reactor design and operation, spent fuel removal from the core, and subsequent storage prior to and after reprocessing, and waste disposal. Decay heat can be readily calculated from the nuclear decay properties of the fission products, actinides and their decay products as generated within the irradiated fuel. Much of the information comes from experiments performed with HPGe detectors, which often underestimate the beta feeding to states at high excitation energies. This inability to detect high-energy gamma emissions effectively results in the derivation of decay schemes that suffer from the pandemonium effect, although such a serious problem can be avoided through application of total absorption γ-ray spectroscopy (TAS). The beta decay of key radionuclei produced as a consequence of the neutron-induced fission of 235U and 239Pu are being re-assessed by means of this spectroscopic technique. A brief synopsis is given of the Valencia-Surrey (BaF2) TAS detector, and their method of operation, calibration and spectral analysis.

  3. Decay heat removal in HTGR by natural circulation

    The coolability of 1000MWt HTGR under complete loss of forced flow conditions was investigated. A dynamic simulator for nuclear power plants (DSNP) program was developed to study the evolution of natural circulation during loss of forced flow (LOF). It was concluded that a passive cooling device, having 0.3% heat removal capacity of nominal decay heat rate, is sufficient for emergency cooling of the proposed plant under total LOF. (author)

  4. Decay heat removal system in a medium size HTGR

    The present study investigates the operation of the decay heat removal system (DHR) during loss of flow (LOF) accident in which the primary flow is reduced to zero at full power operation of medium size high temperature gas reactor (HTGR). (author)

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

    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

  6. Decay heat calculations for a 500 kW W-Ta spallation target

    Yu, Quanzhi; Lu, Youlian; Hu, Zhiliang; Zhou, Bin; Yin, Wen; Liang, Tianjiao

    2015-05-01

    The China Spallation Neutron Source (CSNS) is a short-pulsed neutron scattering facility. The beam power is designed to be 100 kW in Phase I, with the capability of upgrading to 500 kW. Tantalum (Ta)-cladded tungsten (W) was chosen as the spallation target due to its high neutron yield. Ta claddings can solve the problem of the corrosiveness of W plates, although they produce high decay heat after intense irradiation. This paper presents the decay heat distributions and evolutions for the future upgraded 500 kW W-Ta spallation target. The calculations are performed using the MCNPX2.5 Monte Carlo code and the CINDER'90 activation code. The decay heat distributions show that for the W plates, decay heat is mainly produced via the spallation reaction process, whereas for the Ta claddings, it is mainly produced via the neutron capture process. An effective method of reducing the decay heat in the W-Ta target is also presented and discussed.

  7. Multidimensional Thermal-Hydraulic Analysis for Decay Heat Exchanger of PGSFR

    Hong, Jonggan; Yoon, Jung; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The decay heat exchanger (DHX) of PGSFR is a shell-and-tube type counter-current flow sodium heat exchanger, and each unit is designed for the rated thermal power of 1.0 MWt, which is corresponding to the nominal design capacity of a single passive decay heat removal system (PDHRS) and active decay heat removal system (ADHRS) loops. The DHX unit is fully immersed in the cold sodium pool region and removes the system heat load sufficiently and reliably during the temperature transient. In this work, a multidimensional thermal-hydraulic analysis for the DHX was carried out numerically and the numerical results were compared with the calculated results of the 1-D DHX design code to verify the reliability of the design code. In addition, an influence of the cold pool sodium which flows into the shell-side of the DHX through the shell outlet was evaluated. The SHXSA code was conservative in calculating the pressure drop of the shell-side which is our major concern in designing the natural circulation of the decay heat removal system. It was revealed that the buffer region is needed to reduce the thermal stress in the lower tubesheet by the inflow of the cold pool sodium.

  8. Multidimensional Thermal-Hydraulic Analysis for Decay Heat Exchanger of PGSFR

    The decay heat exchanger (DHX) of PGSFR is a shell-and-tube type counter-current flow sodium heat exchanger, and each unit is designed for the rated thermal power of 1.0 MWt, which is corresponding to the nominal design capacity of a single passive decay heat removal system (PDHRS) and active decay heat removal system (ADHRS) loops. The DHX unit is fully immersed in the cold sodium pool region and removes the system heat load sufficiently and reliably during the temperature transient. In this work, a multidimensional thermal-hydraulic analysis for the DHX was carried out numerically and the numerical results were compared with the calculated results of the 1-D DHX design code to verify the reliability of the design code. In addition, an influence of the cold pool sodium which flows into the shell-side of the DHX through the shell outlet was evaluated. The SHXSA code was conservative in calculating the pressure drop of the shell-side which is our major concern in designing the natural circulation of the decay heat removal system. It was revealed that the buffer region is needed to reduce the thermal stress in the lower tubesheet by the inflow of the cold pool sodium

  9. Decaying particles do not heat up the universe

    It is usually assumed that a massive relic species, which comes to dominate the mass density of the Universe and later decays, heats up the Universe when the age of the Universe approx. = its lifetime. We show that if its decay follows the usual exponential decay law, then the Universe is never reheated, rather it tust cools more slowly. We calculate the evolution of the temperature and entropy, and find that to within numerical factors of order unity, the usual estimates for the entropy increase are found to be correct. Our results have implications for primordial nucleosynthesis in scenarios where a massive relic with lifetime approx. = 10-2 to 103 sec is present, and for baryogenesis in the new inflationary Universe scenario

  10. An analysis of decay heat power in the experimental VHTR

    Decay heat power of the typical fuels loaded in the experimental multipurpose very high temperature gas-cooled reactor (VHTR) has been calculated, based on the American National Standard which has been newly developped as to be applicable to light water reactors. Physical constants of the reactor core, such as fission power during operation, burn-up of fissile atoms and change in composition, have been evaluated with the VHTR lattice burn-up code DELIGHT-5. The analysis has been done for the fuels of different enrichments under a given power density and for the cases in which the power density changes during operation, then some characteristics of the VHTR decay heat, especially the effect of operation period, are shown. (author)

  11. Decay heat removal in GEN IV gas cooled fast reactors

    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.

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

    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 104 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 109 seconds, but provides only an ''upper bound'' on the capture correction in the interval 104 9 seconds. (author)

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

    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., 4 figs., 1 tab

  14. HTGR spent fuel element decay heat and source term analysis

    Sund, R.E.; Strong, D.E.; Engholm, B.A.

    1977-02-01

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm/sup 3/ and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations.

  15. HTGR spent fuel element decay heat and source term analysis

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm3 and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations

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

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

  17. Effect of yeast antagonist in combination with heat treatment on postharvest blue mold decay and Rhizopus decay of peaches.

    Zhang, Hongyin; Wang, Lei; Zheng, Xiaodong; Dong, Ying

    2007-04-01

    The potential of using heat treatment alone or in combination with an antagonistic yeast for the control of blue mold decay and Rhizopus decay of peaches caused by Penicillium expansum and Rhizopus stolonifer respectively, and in reducing natural decay development of peach fruits, as well as its effects on postharvest quality of fruit was investigated. In vitro tests, spore germination of pathogens in PDB was greatly controlled by the heat treatment of 37 degrees C for 2 d. In vivo test to control blue mold decay of peaches, heat treatment and antagonist yeast, as stand-alone treatments, were capable of reducing the percentage of infected wounds from 92.5% to 52.5% and 62.5%, respectively, when peach fruits stored at 25 degrees C for 6 d. However, in fruit treated with combination of heat treatment and Cryptococcus laurentii, the percentage of infected wounds of blue mold decay was only 22.5%. The test of using heat treatment alone or in combination with C. laurentii to control Rhizopus decay of peaches gave a similar result. The application of heat treatment and C. laurentii resulted in low average natural decay incidences on peaches after storage at 4 degrees C for 30 days and 20 degrees C for 7 days ranging from 40% to 30%, compared with 20% in the control fruit. The combination of heat treatment and C. laurentii was the most effective treatment, and the percentage of decayed fruits was 20%. Heat treatment in combination with C. laurentii had no significant effect on firmness, TSS, ascorbic acid or titratable acidity compared to control fruit. Thus, the combination of heat treatment and C. laurentii could be an alternative to chemicals for the control of postharvest decay on peach fruits. PMID:17140691

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

    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 235U. 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

  19. Beta decay of fission products for the non-proliferation and decay heat of nuclear reactors

    Today, nuclear energy represents a non-negligible part of the global energy market, most likely a rolling wheel to grow in the coming decades. Reactors of the future must face the criteria including additional economic but also safety, non-proliferation, optimized fuel management and responsible management of nuclear waste. In the framework of this thesis, studies on non-proliferation of nuclear weapons are discussed in the context of research and development of a new potential tool for monitoring nuclear reactors, the detection of reactor antineutrinos, because the properties of these particles may be of interest for the International Agency of Atomic Energy (IAEA), in charge of the verification of the compliance by States with their safeguards obligations as well as on matters relating to international peace and security. The IAEA encouraged its member states to carry on a feasibility study. A first study of non-proliferation is performed with a simulation, using a proliferating scenario with a CANDU reactor and the associated antineutrinos emission. We derive a prediction of the sensitivity of an antineutrino detector of modest size for the purpose of the diversion of a significant amount of plutonium. A second study was realized as part of the Nucifer project, an antineutrino detector placed nearby the OSIRIS research reactor. The Nucifer antineutrino detector is dedicated to non-proliferation with an optimized efficiency, designed to be a demonstrator for the IAEA. The simulation of the OSIRIS reactor is developed here for calculating the emission of antineutrinos which will be compared with the data measured by the detector and also for characterizing the level of background noises emitted by the reactor detected in Nucifer. In general, the reactor antineutrinos are emitted during radioactive decay of fission products. These radioactive decays are also the cause of the decay heat emitted after the shutdown of a nuclear reactor of which the estimation is an

  20. Development of a combined heat exchanger design concept for an SFR decay heat removal system

    A fast neutron spectrum reactor is one of the most promising options for efficient uranium resources utilization and a substantial reduction of radioactive waste to be disposed. To this end, the Korea Atomic Energy Research Institute (KAERI) has developed the own sodium cooled fast reactor(SFR) design concept since 1992, and recent efforts putting into this area have been focused on enhancement of plant safety complying with the lessons learned from the Fukushima nuclear power plant accident. In particular, a reliable decay heat removal (DHR) becomes one of the most important tasks in successful SFR design. Therefore, to achieve more reliable DHR performance, KAERI has developed the innovative design concept called the PDRC, which is similar to conventional DRACS but its detailed flow path inside the reactor vessel is very creative. The schematic of the heat transport system in DSFR 600 is depicted in Figure 1 as an example. In regard to the DHR operation, the internal flow path passing through the reactor core should be maintained at all time but, in accident conditions, converted from the normal heat transport mode using the intermediate heat transport system (IHTS) to the alternate path with the DRACS loops. If the normal heat transport path via the IHTS is not available, the DRACS shall substitute the normal path and remove total heat load including core decay and sensible heat from the primary sodium pool. Since heat rejection from the intermediate heat exchanger (IHX) is not guaranteed in this situation, the decay heat exchanger (DHX) becomes the only heat sink and thus its arrangement inside the reactor vessel plays an important role in determining DHR capability. In the current SFR design, however, the internal flow path from the hot pool to the cold pool is somewhat ambiguous due to the split flow ratio formed in a parallel path between IHXs and DHXs. This ambiguity results in a large uncertainty in DHX shell side flowrate and corresponding heat transfer

  1. Applications of the total absorption technique to improve reactor decay heat calculations: study of the beta decay of 102,104,105Tc

    The decay heat of the fission products plays an important role in predicting the heat-up of nuclear fuel after reactor shutdown. This form of energy release is calculated as the sum of the energy-weighted activities of all fission products P(t) = ΣEiλiNi(t), where Ei is the decay energy of nuclide i(gamma and beta component), λi is the decay constant of nuclide i and Ni(t) is the number of nuclide i at cooling time t. Even though the reproduction of the measured decay heat has improved in recent years, there is still a long standing discrepancy at t∼1000 s cooling time for some fuels. A possible explanation for this disagreement can been found in the work of Yoshida et al., who demonstrated that an incomplete knowledge of the β-decay of some Tc isotopes could be the source of the systematic discrepancy. We have recently measured the β-decay process of some Tc isotopes using a total absorption spectrometer at the IGISOL facility in Jyvaeskylae. The results of the measurements are discussed, along with their impact on summation calculations.

  2. Measurement and analysis of JOYO MK-II spent MOX fuel decay heat. 2

    Decay heat of the spent MOX fuel is important not only from the viewpoint of reactor safety concerning a decay heat removal at the reactor shut down, but also for the thermal design of the spent fuel storage and handling facility. In order to obtain the experimental data and to improve the accuracy of calculation, the decay heat of spent fuel subassemblies of the JOYO Mk-II core was measured. The burn-up was 66 GWd/t and the cooling time was between 40 and 150 days. Measured decay heat of the spent fuel subassemblies was approximately 1446 ± 24 - 663 ± 20 W. The decay heat was calculated by 'ORIGEN2' code and then it was compared with the measured value. In the 'ORIGEN2' calculation, the JENDL-3.2 cross section library and the JNDC-V2 decay data library were used and fuel power calculated by the core management code system 'MAGI' was used as a input. The ratios between calculated and experimental values, C/Es, were approximately between 0.94 and 0.90. The discrepancy between calculation and measurement was considered to be larger than the experimental error (1σ = 1.7 - 3.0%) or the uncertainty of calculated FP decay heat (1 - 2%). It appears due to the uncertainty of actinides decay heat and that indicates cross sections of actinides and initial composition of actinides are important to evaluate decay heat accurately. (author)

  3. Evaluation of Heat Removal Performance of Passive Decay Heat Removal system for S-CO2 Cooled Micro Modular Reactor

    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

  4. Measurement and analysis of JOYO MK-II spent MOX fuel decay heat (3)

    It is important to precisely evaluate decay heat of the spent MOX fuel not only from the viewpoint of reactor safety concerning a decay heat removal at the reactor shut down, but also for thermal design of the spent fuel storage and handling facility. In order to obtain the experimental data and to improve the accuracy of calculation, the decay heat of spent fuel subassemblies (Burn-up: 66 GWd/t) of the JOYO MK-II core was measured from the cooling time of 319 to 729 days. Measured decay heat of the spent fuel subassemblies was 351±16∼158±9W. The low flow rate measurement to make large temperature gradient between inlet and outlet of measurement system has allowed us to measure decay heat after cooling time of 500 days by less than 6% error (1σ). The decay heat was calculated by 'ORIGEN2' code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. Then it was compared with the measured value. The ratios between calculated and experimental value, C/Es, were approximately between 0.96 and 0.90. There exists a systematic discrepancy (6∼8%) between calculation and measurement, which was larger than the experimental error (1σ = 1.7∼6.0%). Decay heat was generated from actinides and fission products (FP). Major heat sources in actinides are 242Cm, 238Pu and 241Am. However, nuclides except for 242Cm were not considered as the cause of systematic discrepancy because these decay heat was almost constant throughout this measurement. After cooling time of 100 days, 95Zr, 95Nb, 106Rh and 144Pr remained as major FP decay heat source. As as result, the discrepancy appears due to the decay heat calculation uncertainty of 242Cm and these FP, or measurement error and further investigation will be required. (author)

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

    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

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

    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)

  7. Mixed-oxide fuel decay heat analysis for BWR LOCA safety evaluation

    The mixed-oxide (MOX) fuel decay heat behavior is analyzed for Boiling Water Reactor (BWR) Loss of Coolant Accident (LOCA) safety evaluation. The physical reasoning on why the decay heat power fractions of MOX fuel fission product (FP) are significantly lower than the corresponding decay heat power fractions of uranium-oxide (UOX) fuel FP is illustrated. This is primarily due to the following physical phenomena. -The recoverable energies per fission of plutonium (Pu)-239 and Pu-241 are significantly higher than those of uranium (U)-235 and U-238. Consequently, the fission rate required to produce the same amount of power in MOX fuel is significantly lower than that in UOX fuel, which leads to lower subsequent FP generation rate and associated decay heat power in MOX fuel than those in UOX fuel. - The effective FP decay energy per fission of Pu-239 is significantly lower than the corresponding effective FP decay energy per fission of U-235, e.g., Pu-239's 10.63 Mega-electron-Volt (MeV) vs. U-235's 12.81 MeV at the cooling time 0.2 second. This also leads to lower decay heat power in MOX fuel than that in UOX fuel. The FP decay heat is shown to account for more than 90% of the total decay heat immediately after shutdown. The FP decay heat results based on the American National Standard Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979 standard method are shown very close to the corresponding FP decay heat results based on the ANSI/ANS-5.1-2005 standard method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method are shown very close to but mostly slightly lower than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1971 method. The FP decay heat results based on the ANSI/ANS-5.1-1979 simplified method or the ANSI/ANS-5.1-1971 method are shown significantly larger than the corresponding FP decay heat results based on the ANSI/ANS-5.1-1979 standard method or the ANSI/ANS-5.1-2005 standard method. (authors)

  8. Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE

    Highlights: • Evaluates SCALE capabilities of predicting decay heat in used nuclear fuel. • Validation basis includes 121 full-assembly decay heat measurements. • Key modeling assumptions and data are discussed. • Results indicate good agreement between calculated and measured decay heat. - Abstract: The energy release rate from the decay of radionuclides can be a critical design parameter for used nuclear fuel storage, transportation, and repository engineered systems. Validation of the SCALE nuclear analysis code system capabilities in predicting decay heat for commercial used fuel applications has been performed using decay heat measurements for fuel assemblies irradiated in pressurized and boiling water reactors. The experimental data used for validation include a large number of full-length-assembly decay heat measurements that were performed between 2003 and 2010 at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, Clab, operated by the Swedish Nuclear Fuel and Waste Management Company, SKB. The measured fuel assemblies cover the burnup range 14–51 GWd/MTU and cooling times between 12 and 27 years, which are times of interest to used fuel transportation and storage applications. The validation results indicate good agreement between calculated and measured decay heat values, generally within the reported measurement uncertainty. The effects of key modeling assumptions and data used in the calculations are presented and discussed

  9. Experimental investigation on melt coolability under bottom flooding with and without decay heat simulation

    Highlights: • The effect of decay heat on melt coolability under bottom flooding was studied. • Decay heat of 0.5 MW/m3 was simulated. • A single simulant material with same mass and initial temperature was used. • Quenching of melt pool does not depend on decay heat. • Comparison of with and without decay heat experiments has been presented. - Abstract: Investigations on severe accident phenomena help us in understanding the realistic accidental phenomena for the assessment of associated risk. The societal impact of radiological leakage to the environment has demanded further robustness in the line of defence of nuclear safety. Thus, to ensure the cooling and stabilization of corium within reactor containment in case of severe accident scenarios, many new reactors have been envisaged with core catcher. In this regard, corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. After studying the various cooling strategies, it has been demonstrated that melt coolability using bottom flooding of water is one of the most efficient techniques so far. To study the effect of decay heat on melt pool coolability under bottom flooding condition, two experiments have been performed in this paper; one without the decay heat and the other with decay heat. The test section used for carrying out these experiments consisted of two parts viz. lower part for retaining the melt from furnace, water inlet and melt quenching, and upper part for steam expansion and its outlet. The total height of the test section was 1400 mm and was made of 33 mm thick carbon steel. Total six stainless steel nozzles of diameter 12 mm were used for injecting water at the bottom of the melt pool. The lower part was surrounded by 10 radiative heaters to simulate decay heat of 10 kW which corresponds to 0.5 MW/m3. The experiments showed that quenching of about 25 l of melt at initial temperature of nearly 1200 °C took only a few minutes

  10. Probabilistic approach for decay heat uncertainty estimation using URANIE platform and MENDEL depletion code

    Tsilanizara, A.; Gilardi, N.; Huynh, T. D.; Jouanne, C.; Lahaye, S.; Martinez, J. M.; Diop, C. M.

    2014-06-01

    The knowledge of the decay heat quantity and the associated uncertainties are important issues for the safety of nuclear facilities. Many codes are available to estimate the decay heat. ORIGEN, FISPACT, DARWIN/PEPIN2 are part of them. MENDEL is a new depletion code developed at CEA, with new software architecture, devoted to the calculation of physical quantities related to fuel cycle studies, in particular decay heat. The purpose of this paper is to present a probabilistic approach to assess decay heat uncertainty due to the decay data uncertainties from nuclear data evaluation like JEFF-3.1.1 or ENDF/B-VII.1. This probabilistic approach is based both on MENDEL code and URANIE software which is a CEA uncertainty analysis platform. As preliminary applications, single thermal fission of uranium 235, plutonium 239 and PWR UOx spent fuel cell are investigated.

  11. Sodium Experiments on Natural Circulation Decay Heat Removal and 3D Simulation of Plenum Thermal Hydraulics

    Natural circulation decay heat removal is one of the significant issues for fast reactor safety, especially in long term station blackout events. Several sodium experiments were carried out using a 7-subassembly core model for core thermal hydraulics under natural circulation conditions and for onset transients of natural circulation in a decay heat removal system (DHRS) including natural draft. Significant heat removal via inter-wrapper flow was confirmed in the experiments. Solidification of sodium in an air cooler is one of key issues in loss of heat sink events. Natural circulation characteristics under long-term decay heat removal were also obtained. Multi-dimensional phenomena, e.g., thermal stratification and bypass flow in plenums and/or heat exchangers, may influence the natural circulation. Thus, 3D simulation method was developed for entire region in the primary loop. Comparison of temperature distributions in a DHRS heat exchanger between experiment and analysis was done. (author)

  12. Study of decay heat removal and structural assurance by LBB concept of tokamak components

    Since decay heat density in ITER is quite low, thermal analyses have shown that only natural dissipation due to thermal radiation can be sufficient for removal of decay heat even in loss of all coolant. Owing to this attractiveness, no cooling system would be required for decay heat removal. In addition, because a magnetically confined plasma terminates by a small amount of impurity ingress, there is no possibility of uncontrolled production of energy, which will damage the integrity of the vacuum vessel containing tritium and other radioactive materials. This statement can be assured with a high level of confidence resulted from the LBB (Leak Before Break) concept. (author)

  13. A study on the decay heat removal capability of a reactor vessel auxiliary cooling system

    A reactor vessel auxiliary cooling system (RVACS) is a potential candidate as a fully passive decay heat removal system for small FBRs. In this study the heat transfer performance of a collector with fins is discussed through experiment and the evaluation method is proposed for the heat removal capability of the system. (author)

  14. Time decay rates for the equations of the compressible heat-conductive flow through porous media

    Chen, Qing; Tan, Zhong; Wu, Guochun

    2015-11-01

    We consider the time decay rates of smooth solutions to the Cauchy problem for the equations of the compressible heat-conductive flow through porous media. We prove the global existence and uniqueness of the solutions by the standard energy method. Moreover, we establish the optimal decay rates of the solution as well as its higher-order spatial derivatives. And the damping effect on the time decay rates of the solution is studied in detail.

  15. Activation experiments on vanadium alloy NIFS-HEAT-2

    In the present study, activation analysis of impurities and evaluation of activation properties were performed on NIFS-HEAT-2 by DT neutron irradiation at FNS facility. Similar analysis and evaluation were performed on US and Chinese V-4Cr-Ti samples for comparison. For impurity analysis and direct evaluation of activation properties of vanadium alloys, activation experiments with DT neutron irradiations were performed on NIFS-HEAT-2 and Round-Robin samples from the US and China. Eight nuclides of 24Na, 28Al, 54Mn, 56Mn, 57Co, 58Co, 89Zr and 92mNb were identified form analysis of the gamma peaks and concentrations of Al, Si, Mn, Fe, Ni, Co, Zr, Nb and Mo were evaluated. It was confirmed that the concentration of Al in NIFS-HEAT-2, which is harmful for low activation property, was lower than the criteria required for recycling of used material after reactor shutdown. The results were almost consistent with those by chemical analysis. Until ∼8 months after irradiation, significant influence of impurities was not observed in the decay heat measurement. Results of decay heat measurement with the Whole Energy Absorption Spectrometer and those of calculation with MNCP-4C, ACT-4 and JENDL Activation File were consistent within ∼ 15%. Activation calculation considering impurity concentrations from the present analysis indicated that decay heat of 60Co transmuted from Ni impurity will be dominant ∼6 years after irradiation. (author)

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

    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)

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

    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.

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

    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 (Rcd) of 420 and neutron flux (Φth) of 1.6x106 n/cm2.s. This thermal neutron beam has been applied for measurements of capture cross-sections for nuclide of 51V, 55Mn, 180Hf and 186W by the activation method relative to the standard reaction 197Au(n,g)198Au. 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 235U, 238U, 239Pu and 232Th are introduced in this report. (author)

  19. [The influence of oil heat treatment on wood decay resistance by Fourier infrared spectrum analysis].

    Wang, Ya-Mei; Ma, Shu-Ling; Feng, Li-Qun

    2014-03-01

    Wood preservative treatment can improve defects of plantation wood such as easy to corrupt and moth eaten. Among them heat-treatment is not only environmental and no pollution, also can improve the corrosion resistance and dimension stability of wood. In this test Poplar and Mongolian Seoteh Pine was treated by soybean oil as heat-conducting medium, and the heat treatment wood was studied for indoor decay resistance; wood chemical components before and after treatment, the effect of heat treatment on wood decay resistance performance and main mechanism of action were analysed by Fourier infrared spectrometric. Results showed that the mass loss rate of poplar fell from 19.37% to 5% and Mongolian Seoteh Pine's fell from 8.23% to 3.15%, so oil heat treatment can effectively improve the decay resistance. Infrared spectrum analysis shows that the heat treatment made wood's hydrophilic groups such as hydroxyl groups in largely reduced, absorbing capacity decreased and the moisture of wood rotting fungi necessary was reduced; during the heat treatment wood chemical components such as cellulose, hemicellu lose were degraded, and the nutrient source of wood rotting fungi growth necessary was reduced. Wood decay fungi can grow in the wood to discredit wood is because of that wood can provide better living conditions for wood decay fungi, such as nutrients, water, oxygen, and so on. The cellulose and hemicellulose in wood is the main nutrition source of wood decay fungi. So the oil heat-treatment can reduce the cellulose, hemicellulose nutrition source of wood decay fungi so as to improve the decay resistance of wood. PMID:25208386

  20. Fission Product Decay Heat Calculations for Neutron Fission of 232Th

    Son, P. N.; Hai, N. X.

    2016-06-01

    Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.

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

    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)

  2. Heat transfer in turbulent decaying swirl flow in a circular pipe

    Algifri, A. H.; Bhardwaj, R. K.; Rao, Y. V. N.

    1988-08-01

    Heat transfer coefficients for air are measured along a heated pipe for decaying swirl flow, generated by radial blade cascade. The results are compared with an expression proposed for predicting the heat transfer coefficients in swirling flow. The theoretical predictions are in good agreement with the experimental data, with average and maximum deviations of 7 and 11 percent, respectively. The application of the theoretical approach to the experimental results obtained by other investigators for heat transfer in a decaying swirl flow generated by short-twisted tapes and tangential slots at inlet also give rise to encouraging agreement.

  3. Beta and gamma decay-heat measurements for fast and thermal reactors, particularly for Pu-239

    Dickens, J.K.

    1981-01-01

    The mounting evidence suggests (a) that the ORNL thermal-neutron fission decay heat data are likely correct to within stated uncertainties; (b) present data files (ENDF/B-V and others related to it) require more effort, and as improvements are made agreement with ORNL results is likely to become better; and (c) fast fission decay heat, induced by reactor-spectrum neutrons, appears to be little, if any, different from thermal-neutron fission-decay heat. The ANS 5.1 working group seems to be moving toward the above position but the standard for /sup 235/U has not yet been revised. Using /sup 235/U decay-heat data for /sup 239/Pu is quite conservative. Use of /sup 235/U decay-heat data for /sup 241/Pu decay-heat data as required in the standard is conservative only because of the over-statement (by approx. = 8%) of the /sup 235/U data in the standard.

  4. Decay heat removal in pool type fast reactor using passive systems

    Highlights: ► Three dimensional thermal hydraulic analysis of decay heat system in a fast reactor model predictions compared with experimental results from PHENIX. ► Calculations confirm adequacy of natural convection in decay heat removal. ► Inter-wrapper flow found to reduce peak temperatures by 50 K in the blanket zone. - Abstract: Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement

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

    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. Decay heat and anti-neutrino energy spectra in fission fragments from total absorption spectroscopy

    Rykaczewski, Krzysztof

    2015-10-01

    Decay studies of over forty 238U fission products have been studied using ORNL's Modular Total Absorption Spectrometer. The results are showing increased decay heat values, by 10% to 50%, and the energy spectra of anti-neutrinos shifted towards lower energies. The latter effect is resulting in a reduced number of anti-neutrinos interacting with matter, often by tens of percent per fission product. The results for several studied nuclei will be presented and their impact on decay heat pattern in power reactors and reactor anti-neutrino physics will be discussed.

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

    Porta A.

    2016-01-01

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

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

    Porta, A.; Zakari-Issoufou, A.-A.; Fallot, M.; Algora, A.; Tain, J. L.; Valencia, E.; Rice, S.; Bui, V. M.; Cormon, S.; Estienne, M.; Agramunt, J.; Äystö, J.; Bowry, M.; Briz, J. A.; Caballero-Folch, R.; Cano-Ott, D.; Cucouanes, A.; Elomaa, V.-V.; Eronen, T.; Estévez, E.; Farrelly, G. F.; Garcia, A. R.; Gelletly, W.; Gomez-Hornillos, M. B.; Gorlychev, V.; Hakala, J.; Jokinen, A.; Jordan, M. D.; Kankainen, A.; Karvonen, P.; Kolhinen, V. S.; Kondev, F. G.; Martinez, T.; Mendoza, E.; Molina, F.; Moore, I.; Perez-Cerdán, A. B.; Podolyák, Zs.; Penttilä, H.; Regan, P. H.; Reponen, M.; Rissanen, J.; Rubio, B.; Shiba, T.; Sonzogni, A. A.; Weber, C.

    2016-03-01

    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.

  9. Measurement and analysis of decay heat of fast reactor spent fuel

    Decay heat of the JOYO Mk-II spent fuel subassemblies was measured using the calorimetric method as a non-destructive examination. The measurement of the subassembly was taken in the spent fuel storage pond at JOYO. Its burn-up was approximately 60 GWd/t and the cooling time was between 24 and 258 days. The measured decay heat was compared with the calculated values by 'ORIGEN2' code using the cross section library of JENDL-3.2, and the decay data and the fission yield data library of JNDC-V2. Calculated to Experimental (C/E) values for the measured subassembly were 0.96 - 0.90. The decay heat generated by 238Pu and 241Am, which amount to 1% of initial composition of the fresh fuel, reached 6% - 17% of decay heat during 24 -258 days of cooling time. These results indicate that the initial composition and the burn-up calculation of actinides are important to accurately evaluate the decay heat of spent fuel. (author)

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

    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.

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

    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. Los Alamos PWR decay-heat-removal studies. Summary results and conclusions

    Boyack, B E; Henninger, R J; Horley, E; Lime, J F; Nassersharif, B; Smith, R

    1986-03-01

    The adequacy of shutdown-decay-heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is the review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators were unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performance of the Oconee-1 and Calvert Cliffs-1 reactors of Bobcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss of secondary heat sink. Feed and bleed was successfully applied in two of the plants, Oconee-1 and Zion-1, provided it was initiated no later than the time of primary system saturation. Feed and bleed in Calvert Cliffs-1 when initiated at the time of primary system saturation did result in core dryout; however, the core heatup was eventually terminated by coolant injection. Feed-and-bleed initiation at primary system saturation was not studied for H.B. Robinson-2. Insights developed during the analyses of specific plant transients have been identified and documented. 33 refs., 107 figs., 26 tabs.

  13. Los Alamos PWR decay-heat-removal studies. Summary results and conclusions

    The adequacy of shutdown-decay-heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is the review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators were unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performance of the Oconee-1 and Calvert Cliffs-1 reactors of Bobcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss of secondary heat sink. Feed and bleed was successfully applied in two of the plants, Oconee-1 and Zion-1, provided it was initiated no later than the time of primary system saturation. Feed and bleed in Calvert Cliffs-1 when initiated at the time of primary system saturation did result in core dryout; however, the core heatup was eventually terminated by coolant injection. Feed-and-bleed initiation at primary system saturation was not studied for H.B. Robinson-2. Insights developed during the analyses of specific plant transients have been identified and documented. 33 refs., 107 figs., 26 tabs

  14. Decay heat analysis of a VHTR core using the HELIOS and origen-2 codes

    This paper describes the procedure and results of a decay heat analysis in a relatively short time after a shutdown for the safety analysis of a VHTR core. In this analysis, HELIOS provides the one-group actinide cross sections to ORIGEN-2 through the 190 group lattice calculation for a single fuel block. Then, ORIGEN-2 performs the depletion and decay heat calculations using these actinide cross sections. After benchmarking this procedure against a PWR core, it was applied to a 200 MWth prismatic VHTR core. The results showed that the decay heat per unit operating power is very comparable to that for a typical large power PWR core, although the decay heat per unit heavy metal mass is three times higher than that in the PWR core. It was also found from the results that the decay heat fraction to operating power decreases very slowly with the core burnup after it reaches a maximum value of 6.1 percents at 5 GWd/tHM. (authors)

  15. Revisions to the decay heat standard: ANSI/ANS 5.1

    Brady Raap, M. C. [Pacific Northwest National Laboratory, 902 Battelle Blvd., Richland, WA 99352 (United States); Gauld, I. C. [Oak Ridge National Laboratory, Oak Ridge, TN 37832 (United States); Dickens, J. K. [Dept. of Physics, Univ. of Tennessee, Knoxville, TN 37996 (United States); Wilson, W. B. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2006-07-01

    This paper presents a summary of the 2005 revision to ANSI/ANS-5.1, Decay Heat Power for Light Water Reactors. The 2005 revision to ANSI/ANS-5.1 [1] represented an important milestone for this Standard. The reconstitution of the working group, incorporation of ENDF/B-VI data, and revisions to the 'simplified method' are the most significant accomplishments realized in the approval and publication of this Standard. A discussion comparing the content of the 2005 revision to the previous version of the Standard [2] is provided. In addition, an outline of the current activities of the working group and goals for the next revision is presented. (authors)

  16. Validation of ORIGEN-S decay heat predictions for LOCA analysis

    Recent developments in the nuclear data libraries used by the ORIGEN-S isotope generation and depletion code have enabled the extension of the code to accurately predict the delayed energy release rates from nuclear decay (decay heat) at very short cooling times of interest to reactor accident analysis. Historically this time domain has required integral methods, such as the ANS-5.1 decay heat standard, because isotopic summation codes could not be reliably applied due to incomplete nuclear data. This paper describes work to validate ORIGEN-S against experimental measurements for decay times that extend down to about 1 second after fission. Benchmarks using measured gamma ray spectra following fission are also included because these results are important to predicting spatial energy deposition from delayed gamma energy release. (authors)

  17. The decay heat of fission products and actinides of the SNR-300

    The report describes the computer code RASPA, which calculates the build-up and decay of fission products and actinides. The verification of the code and its library has been performed by comparison with theoretical and experimental results of other authors, whereby a good agreement has been achieved. Furthermore, an error analysis has shown, that the error of the calculated decay heat, which is induced by uncertainties of nuclear data, is less than 10 % up to decay times of one month. The results of calculations of the time dependent decay heat and the gamma source strength in various zones of the cores Mark-Ia and Mark-II of the SNR-300 are documented and discussed in detail

  18. Investigation of interaction between heat transport systems during the natural circulation decay heat removal in FBRs. Influence of decay heat removal system type and the secondary heat transport system

    Steady state sodium experiments were performed to investigate interactions between the heat transport systems, i.e., the primary system, the secondary system, and the decay heat removal system, during the natural circulation decay heat removal in FBRs. The test rig was used for the experiments. The core model has seven subassemblies; the center assembly simulates pin bundle geometry of a core fuel subassembly in a large scale FBR and consists of 37 pins, six outer subassemblies consists of 7 pins. As the decay heat removal system, Direct Reactor Auxiliary Cooling System (DRACS) and Primary Reactor Auxiliary Cooling System (PRACS) can be selected. Experiments were carried out under natural circulation conditions in the primary loop and force convection conditions in the decay heat removal system. In cases using DRACS, natural circulation flow rate in the primary loop was smaller by 20% than that in cases using PRACS due to the low temperature in the upper plenum and also in the upper non-heated section of the core. When natural circulation was allowed in the secondary heat transport system, the natural circulation flow rate in the primary system increased in spite of the operation of DRACS. In cases using DRACS, inter-subassembly flow redistribution occurred; the center subassembly had larger flow rate than those in outer subassemblies due to the low natural circulation head in the outer subassemblies which were cooled by the inter-wrapper flow (IWF). The highest temperature in the core was reduced by IWF via not only the direct cooling effect but also the inter-subassembly flow redistribution. (J.P.N.)

  19. Preliminary Test of a small heat pipe for hybrid control rod in-core passive decay heat removal system

    This paper introduces 'Hybrid control rod' combining its original function and heat removal ability. The high temperature operation and high resistance of radiation should be considered to adopt the hybrid heat pipe at the in-core condition. Other design consideration is to make extra inlet parts because it has a high risk of inlet boundary failure. It means that the introduction of heat pipe system is difficult to present nuclear power plants. The other concepts are presented to out-core cooling design but it has low performance compared with in-core heat removal system. Hybrid heat pipe for in-core heat removal system suggests the solution of these problems. Ultimate objective of this research is to develop the passive emergency decay heat removal system using hybrid heat pipes targeting design bases accidents such as station black-out (SBO) and small break loss of coolant accident (SBLOCA). The purpose of this work is to confirm the performance and heat transfer behavior of hybrid heat pipe. The hybrid heat pipe has special condition for operation. Therefore, it is hard to analyze their behavior in core. Table I shows the characteristics of hybrid heat pipe and consideration for manufacturing the heat pipe

  20. Preliminary Test of a small heat pipe for hybrid control rod in-core passive decay heat removal system

    Kim, In Guk; Ban, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    This paper introduces 'Hybrid control rod' combining its original function and heat removal ability. The high temperature operation and high resistance of radiation should be considered to adopt the hybrid heat pipe at the in-core condition. Other design consideration is to make extra inlet parts because it has a high risk of inlet boundary failure. It means that the introduction of heat pipe system is difficult to present nuclear power plants. The other concepts are presented to out-core cooling design but it has low performance compared with in-core heat removal system. Hybrid heat pipe for in-core heat removal system suggests the solution of these problems. Ultimate objective of this research is to develop the passive emergency decay heat removal system using hybrid heat pipes targeting design bases accidents such as station black-out (SBO) and small break loss of coolant accident (SBLOCA). The purpose of this work is to confirm the performance and heat transfer behavior of hybrid heat pipe. The hybrid heat pipe has special condition for operation. Therefore, it is hard to analyze their behavior in core. Table I shows the characteristics of hybrid heat pipe and consideration for manufacturing the heat pipe.

  1. A simple method for evaluation of uncertainties in fission product decay heat summation calculations

    The present precision of nuclear data for the aggregate decay heat evaluation is analyzed quantitatively for 50 fissioning systems. In the practical calculation, a simple approximate method is proposed in order to avoid complication of the calculation and to point out easily the main causal nuclei of the uncertainties in decay heat calculations. As for the independent yield, the correlation among the values is taken into account. For this evaluation, nuclear data and their uncertainty data are taken from ENDF/B-VI nuclear data library. (author)

  2. Water test of natural circulation for decay heat removal in JSFR

    Japan Atomic Energy Agency (JAEA) is conducting a design study of Japan Sodium-cooled Fast Reactor (JSFR), in which a decay heat removal system (DHRS) utilizing natural circulation (NC) is applied as an innovative technology. The Central Research Institute of Electric Power Industry (CRIEPI) has carried out a water test to verify the applicability of NC to decay heat removal. The test used a 1/10-scale model of JSFR. Key issues on thermal-hydraulics were identified through simulation tests on representative events. Measures were also proposed to resolve these issues. The present study has demonstrated that a sufficient and stable NC is established in each event. (author)

  3. A probabalistic reliability evaluation of decay heat removal system using phased mission method and Markov model

    This paper presents a reliability evaluation model based on the 'phased mission method' combined with the Markov model and the results of its application to the decay heat removal system (DHRS) of a large tank-type FBR. The new approach divides the mission time into several time phases and the unreliability value in each time phase is calculated by solving simultaneous differential equations based on the Markov model. An application to an FBR decay heat removal system evaluation has shown that this realistic approach offers a methodology for improving reliability values by a factor of 10 as compared with the conventional fault tree method. (author)

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

    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)

  5. Distribution of the decay heat in various modul HTRs and influence on peak fuel temperatures

    A unique feature of modular high-temperature reactors (MODUL-HTRs) is their benign response to a Loss-Of-Coolant Accident (LOCA). The reactor inherently becomes subcritical; the decay power partly heats up the reactor and partly is removed to the environment via thermal conduction and radiation, while avoiding overheating of the fuel. Production, storage, and removal of the decay heat is studied for different MODUL-HTR concepts having annular-core designs and thermal-powers of 350 MWt. Based on use of Low-Enriched-Uranium/Thorium (LEU/Th) fuel cycles in Prismatic-Fueled Reactors (PFRs), and LEU fuel cycles in Pebble-Bed-Reactors (PBRs), the following has been determined: (1) Comparison of a PFR and a PBR having essentially the same design shows higher decay heat production in the PFR due to a higher fission-product inventory and to the decay of 233Pa bred from 232Th; comparison also shows lower heat-transport rates from the pebble-bed core during a LOCA due to the lower thermal conductivity of the core. (2) Changing the PBR design to utilize carbon bricks and an additional coolant gap in the outer regions of the reactor adds significant barriers to the transport of decay heat to the Reactor Cavity Cooling System (RCCS) which is external to the reactor vessel. (3) In HTRs for Gas-Turbine (GT) applications, the operating temperature of the reactor is higher than in Steam turbine Cycle (SC) HTRs; consequently, under LOCAs the relative heat transport to the RCCS is higher in GT-HTRs. As a result, during a LOCA the amount of heat energy stored in the GT-HTR cores was only about 50% of the amount stored in SC-HTR cores. (author). 8 refs, 9 figs, 5 tabs

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

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

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

    Gierse, N.; Coenen, J W; C. Thomser; Panin, A.; Ch. Linsmeier; Unterberg, B.; Philipps, V

    2015-01-01

    Ferromagnetic pebbles are investigated as high heat flux (q∥) plasma facing components in fusion devices with short power decay length (λq) on a conceptual level. The ability of a pebble concept to cope with high heat fluxes is retained and extended by the acceleration of ferromagnetic pebbles in magnetic fields. An alloying concept suited for fusion application is outlined and the compatibility of ferromagnetic pebbles with plasma operation is discussed. Steel grade 1.4510 is chosen as a ...

  8. Thyroid activity in heat adaptation

    The effect of acute and chronic (22 day round-the-clock) exposure to microenvironmental heat stress (37 deg C DBT) on thyroid activity was studied in Hariana x Holstein Frisian, Hariana x Brown Swiss and Hariana x Jersey non-cycling F1 crossbred heifers. Vis-a-vis their no-heat norms, the percentage uptake of tri-iodothyronine-125I by resin registered a steep fall (about 45 to 60 percent) on acute heat exposures reaching a minimum value in about 2 hrs. The levels started recouping by the 2nd day, plateuing out on the 5th day onwards at slightly subnormal level with up and down fluctuations throughout the three week duration of exposure to heat. There were no significant differences in the pattern or magnitude of response amongst breeds, though in case of Holstein Frisian and Brown Swiss cross-breds the levels of T3 tended, at times, to overshoot the no-stress norm. (author)

  9. DECAY RESISTANCE AND PHYSICAL PROPERTIES OF OIL HEAT TREATED ASPEN WOOD

    Behzad Bazyar

    2011-12-01

    Full Text Available The decay resistance of oil-heat treated aspen wood (Populus tremula l. against white rot fungi (Coriolus versicolor and brown rot fungi (Coniophora puteana was investigated. Three different temperature stages and two time levels for oil heat treatment for the selection of optimum conditions were determined. Linseed oil as a heating medium was used. The mass loss of treated samples that were exposed to both fungi was significantly lower than that of the control samples. Results also showed improvement in dimensional stability after oil heat treatment. Decay resistance and dimensional stability of aspen wood were increased significantly with temperature increasing, but time seemed to have no effect on those properties. Oil heat treatment is a suitable method to improve decay resistance of aspen wood as it reduced the mass loss by 71% and 77% against Coriolus versicolor and Coniophora puteana compared with control samples, respectively. On the other hand, oil heat treatment improved the dimensional stability by about 20.5%.

  10. Feasibility study on enhancement of decay heat removal capacity in LMR using radiation structures

    In this study, a quantitative analysis of heat transfer enhancement by using the radiation structure installed in hot air riser of PSDRS in KALIMER is performed and a feasibility of designing a large thermal rating power plant is investigated in view of decay heat removal capacity. The heat transfer enhancement by radiation structure is directly proportional to the increasing the number of radiation structure, gap width of air flow path, and the temperature of heat transfer area contacting to the natural circulating air. Based on these analyses results of the enhancement of decay heat removal capacity, the feasible maximum core power of a large thermal rating power plant is investigated, and it is founded that the higher operation temperature of reactor coolant system, the larger thermal rating power plant design is possible and the enhancement is about 30% ∼ 40%, quantitatively. From the analyses results, it is expected that a large thermal rating power plant of about 1,000MWt is feasible to design in view of the enhancement of decay heat removal capability

  11. Short term fission product and actinide decay heat for a PWR

    This note gives the results of best estimate calculations of the decay heat following reactor trip for the UK PWR using UK recommended methods. It is intended that these values, together with the uncertainties identified, should be used for the analysis of reactor transients following shutdown. This requires the use of the computer code FISPIN (or a similar code FISP) together with the First UK Library of Fission Product Decay Data (UKPFDD-1), the Crouch 2 fission yields and group averaged fission product capture cross sections recommended individually for each reactor type. The calculations reported here conform to this standard. Decay heat from heavy elements (identified as actinides in this report) is also calculated in FISPIN. (U.K.)

  12. VVER-1000/V320 decay heat analysis involving TVS-M and TVSA fuel assemblies

    Petkov, Plamen V. [' Kozloduy' NPP, 3321 Kozloduy, Vratsa (Bulgaria)], E-mail: pvpetkov@yahoo.com; Hristov, Danail V. [' Kozloduy' NPP, 3321 Kozloduy, Vratsa (Bulgaria)], E-mail: dvhristov@npp.bg

    2008-12-15

    MELCOR 1.8.4 is an integral computer code, developed for severe accident calculations. It is used primarily for the simulation of PWR and BWR types of reactors as there exists an internal database, suitable for modeling of their core inventory. Despite similarity between VVER-1000/V320 and PWR, accounting of specificities of Russian reactor designs is still required. Part of it is the simulation of core decay heat rate after the shutdown. MELCOR 1.8.4 distinguishes fifteen classes. Each of them contains chemical elements with similar properties. Twelve are involved in radioactive products decay. In current paper the authors present two boundary reactor core loadings, designed with corresponding fuel assemblies: TVS-M and TVSA. They have calculated decay heat after reactor shutdown from 100% and 104% of nominal power by SCALE 4.4a package. The amount of generated nuclides had also been estimated. Irradiation history had been accounted as proposed in Kolobashkin et al. (p. 141) [Kolobashkin, V.M., Rubtsov, P.M., Rujanskiy, P.A., Sidorenko, V.D., 1983. Radionuclide Inventory Estimation Handbook (on Russian). Energoatomizdat, Moscow, pp. 138-188]. Newly developed Core Inventory Estimation Tool (CIET), described in this paper, written and tested previously, has been used for the evaluation of core decay heat fractions, distributed over chemical classes. Twelve curves were generated by following the same numerical procedure implemented in MELCOR for representation of decay in W/kg. Comparison of chemical element decay rates to the defaults for PWR shows deviations from the expectations to maximal values of 37% in Uranium for TVSA fuel assemblies. The total number of radionuclides, separated in chemical classes, given in Gauntt et al. [Gauntt, R.O., Cole, R.K., Rodrigez, S.B., Sanders, R.L., Smith, R.C., Stuard, D.S., Summers, R.M., Young, M.F., 1997. MELCOR Computer Code Manuals. NUREG/CR-6119 Report, Vol. 1 and Vol. 2, SAND97-2398] was compared to the ones involved in

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

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

    1978-05-22

    Results from recent integral decay-power experiments are presented and compared with summation calculations. The experiments include the decay power following thermal fission of /sup 233/U, /sup 235/U, and /sup 239/Pu. The summation calculations use ENDF/B-IV decay data and yields from Versions IV and V. Limited comparisons of experimental ..beta.. and ..gamma.. spectra with summation calculations using ENDF/B-IV are included. Generalized least-squares methods are applied to the recent /sup 235/U and /sup 239/Pu decay-power experiments and summation calculations to arrive at evaluated values and uncertainties. Results for /sup 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 /sup 239/Pu are larger. Accurate analytical representations of the decay power are presented for /sup 235/,/sup 238/U, and /sup 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.

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

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

    1978-07-01

    Results from recent integral decay-power experiments are presented and compared with summation calculations. The experiments include the decay power following thermal fission of /sup 233/U, /sup 235/U, and /sup 239/Pu. The summation calculations use ENDF/B-IV decay data and yields from Versions IV and V. Limited comparisons of experimental ..beta.. and ..gamma.. spectra with summation calculations using ENDF/B-IV are included. Generalized least-squares methods are applied to the recent /sup 235/U and /sup 239/Pu decay-power experiments and summation calculations to arrive at evaluated values and uncertainties. Results for /sup 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 /sup 239/Pu are larger. Accurate analytical representations of the decay power are presented for /sup 235/ /sup 238/U, and /sup 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.

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

    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

  16. Markov analysis for time dependent success criteria of passive decay heat removal system

    Highlights: • Unavailability estimation of decay heat removal system with time dependent success criteria. • Dynamic modeling of the system under both continuous and periodic monitoring schemes. • Sensitivity of unavailability to various factors. • Comprehensive Markov modeling with and without considering common cause failures. • Estimation of upper bound and lower bound for the mean unavailability of decay heat removal system. - Abstract: Safety systems deployed in nuclear industry are generally required to operate for a particular mission time. Most of such systems employ redundancy to ensure their high availability over the stipulated mission time. However, availability of a system with redundant configuration depends upon success criteria and is application-specific. The Safety Grade Decay Heat Removal System of Indian Prototype Fast Breeder Reactor is required to operate with different success criteria during the specified mission time on account of steady decline in decay heat produced by the reactor core. In this paper, Markov analysis is carried out to evaluate the availability of the system under both continuous and periodic monitoring schemes. The study estimates the upper bound and lower bound for mean unavailability of SGDHR system for the specified mission time. Sensitivity analysis of the system attributable to important parameters is also carried out. The study provides a comprehensive approach to model scenarios with time dependent success criteria and provides an insight on the factors affecting availability of such systems

  17. FIP Bias Evolution in a Decaying Active Region

    Baker, D; Démoulin, P; Yardley, S L; van Driel-Gesztelyi, L; Long, D M; Green, L M

    2015-01-01

    Solar coronal plasma composition is typically characterized by first ionization potential (FIP) bias. Using spectra obtained by Hinode's EUV Imaging Spectrometer (EIS) instrument, we present a series of large-scale, spatially resolved composition maps of active region (AR) 11389. The composition maps show how FIP bias evolves within the decaying AR from 2012 January 4-6. Globally, FIP bias decreases throughout the AR. We analyzed areas of significant plasma composition changes within the decaying AR and found that small-scale evolution in the photospheric magnetic field is closely linked to the FIP bias evolution observed in the corona. During the AR's decay phase, small bipoles emerging within supergranular cells reconnect with the pre-existing AR field, creating a pathway along which photospheric and coronal plasmas can mix. The mixing time scales are shorter than those of plasma enrichment processes. Eruptive activity also results in shifting the FIP bias closer to photospheric in the affected areas. Final...

  18. Thermal Capacitance (Slug) Calorimeter Theory Including Heat Losses and Other Decaying Processes

    Hightower, T. Mark; Olivares, Ricardo A.; Philippidis, Daniel

    2008-01-01

    A mathematical model, termed the Slug Loss Model, has been developed for describing thermal capacitance (slug) calorimeter behavior when heat losses and other decaying processes are not negligible. This model results in the temperature time slope taking the mathematical form of exponential decay. When data is found to fit well to this model, it allows a heat flux value to be calculated that corrects for the losses and may be a better estimate of the cold wall fully catalytic heat flux, as is desired in arc jet testing. The model was applied to the data from a copper slug calorimeter inserted during a particularly severe high heating rate arc jet run to illustrate its use. The Slug Loss Model gave a cold wall heat flux 15% higher than the value of 2,250 W/sq cm obtained from the conventional approach to processing the data (where no correction is made for losses). For comparison, a Finite Element Analysis (FEA) model was created and applied to the same data, where conduction heat losses from the slug were simulated. The heat flux determined by the FEA model was found to be in close agreement with the heat flux determined by the Slug Loss Model.

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

    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 CO2 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 CO2 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)

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

    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

  1. Study on a decay heat removal system of light water reactors using air coolers

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  2. Near-term decay heat and age of spent fuel in commercial power reactor inventories

    Projections based upon utility supplied data indicate that several commercial nuclear power reactors will require additional spent fuel storage capacity. These storage needs are beyond those which currently exist in the reactor spent fuel storage pools. A key parameter needed to design out-of-pool or consolidated storage capacity to handle these needs is the decay heat of the spent fuel to be stored. Calculations of the heat of the additional spent fuel requiring storage have been made. These calculations indicate small differences in the decay heat of the selected fuel assemblies depending upon whether the oldest or the coldest fuel in the existing inventory is selected first for out-of-pool or consolidated storage

  3. Detailed processes accompanying the decay of an active region

    High resolution (better than 1'') magnetograms obtained at the Sacramento Peak Vacuum Tower Telescope were used to study the decay of a small active region. The reduction process allows one to match intensity and magnetic pictures exactly. Some of the main results are: (i) The granulation massages a magnetic pore, probably inducing its fragmentation. The supergranules get rid of the decayed pieces transporting them away from the pore. (ii) Magnetic flux is removed from the photosphere through its submergence. (author). 5 figs., 10 refs

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

    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.

  5. Relationship between computed ANSI/ANS-5.1 and ORIGEN-S decay heat powers for BWR LOCA safety analysis

    The decay heat power fraction computed using ANSI/ANS-5.1-1979 with CASMO-4 decay heat parameters is compared with the decay heat power fraction computed using ANSI/ANS-5.1-1979 with ORIGEN-S decay heat parameters. The comparison indicates that the ORIGEN-based ANS-5.1 total decay power fraction appears very close to the CASMO-based ANS-5.1 total decay power fraction due to compensating effect between fission-product decay heat power fraction and U-239 and Np-239 decay heat power fraction, although the CASMO-4 fission fractions and U-238 neutron capture ratio are considered more accurate than the ORIGEN-S fission fractions and U-238 neutron capture ratio. Therefore, it seems acceptable to calculate the total decay heat fraction using ANSI/ANS-5.1-1979 with ORIGEN-S decay heat parameters. This result is useful, since ORIGEN-S /SCALE 5.1 are easier to run than CASMO-4. The decay heat power fraction computed using ANSI/ANS-5.1-1979 is also compared with the decay heat power fraction computed using ORIGEN-S directly. The comparison indicates that the ORIGEN-S total decay heat power fraction is much smaller than the corresponding ANS 5.1 total decay heat power fraction, which is due to the fact that the ORIGEN-S fission-product decay heat power fraction is much smaller than the corresponding ANS 5.1 fission-product decay heat power fraction. This demonstrates that the total decay heat power fraction calculated using ORIGEN-S directly is not conservative and that ANSI/ANS-5.1 must be used to calculate the total decay heat power fraction for LOCA safety analysis. (author)

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

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

  7. Activation product decay data: UKPADD-2 data files

    The decay data of various radionuclides have been evaluated on the basis of a series of well-defined specifications and the requirements of the UK nuclear power, fuel reprocessing and waste management programmes. These radionuclides are primarily activation products and standards that are commonly used in gamma-ray spectroscopy. Recommended data include half-life, branching fractions, alpha, beta and gamma-ray energies and emission probabilities, total decay energy, mean alpha, beta and gamma energies, internal conversion coefficients, and all associated uncertainties. Computer-based files have been generated in ENDF-6 format, including lists of the references used to produce the proposed decay scheme and comments that identify any inadequacies. (author)

  8. Decay heat calculations with the CEA radioactivity data bauk and the code PEPIN

    The CEA radioactivity data bank, has been updated mainly from ENSDF and from some recent experimental results. This library contains the decay data for about 700 fission products (F.P.), 220 actinides and more than 1400 other nuclides. A comparison between our data and ENDF/B5 is shown for the fission products. The fission products part of this library is currently used for shielding and decay heat calculations with the PEPIN code. Calculations and spectral comparisons of the available experiments (Dickens, Lott, Yarmell ...) and other recent calculations is made for thermal fission of 235U and 239Pu using our data bank as input

  9. Experimental studies on natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS were performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increases of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan. (author)

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

    Giacomino Bandini

    2008-01-01

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

  11. Study on decay heat removal characteristics by natural circulation for top entry loop-type FBR

    The thermal-hydraulic circulation characteristics of a top entry loop-type FBR are studied experimentally with a 1/8 scaled three-loop model using water as the working fluid. The model is designed to simulate reactor decay heat removal operations including natural circulation mode as well as three types of decay heat removal systems, a direct reactor auxiliary cooling system (DRACS) with a direct heat exchangers immersed in a hot plenum of a reactor vessel (Immersed type DRACS), a DRACS with DHXs penetrating from a hot plenum to a cold plenum of a reactor vessel (Penetrating type DRACS) and a primary reactor auxiliary cooling system (PRACS). The feasibility of reactor decay heat removal operations including natural circulation mode of a top entry loop-type FBR is confirmed based on the experimental results. The natural circulation performance of the Immersed type DRACS is compared with that of the PRACS for the water test results. The results are used to evaluate the accuracy of the one-dimensional flow-network analysis codes and the multi-dimensional thermal-hydraulic analysis code. The analytical results are generally in good agreement with the measurements. (author)

  12. Compilation and review of loss of decay heat removal during reactor shutdown in PWRs

    This report provides review of the U.S.NRC guidances (Information Notice, Generic Letter, Bulletin) related to loss of decay heat removal during reactor shutdown in PWRs. Referring to the U.S.NRC's Licensee Event Reports (LERs), we also identify events involving actual or potential loss of decay heat removal which occurred in the period from the second half of 1990 to the end of 1992, summarize event description, causes and actions taken for each event, and analyze the direct and root causes. Compilation and review of the U.S.NRC guidances indicate that 15 documents have been issued for loss of decay heat removal event after 1980, most of which pointed out the problems concerning operating procedures of residual heat removal during reduced inventory conditions, procedure and/or administration for maintenance and reactor water level instrumentation. Although many actions taken according to the U.S.NRC guidances have decreased the number of such events, the fact that one or more documents per year have been issued after 1986 implies more importance of such events than before. Review of recent events shows that the primary direct cause has been 'vortexing' or 'air entrainment' due to lowering reactor water level too far or loss of coolant. As for root causes, it is found that most of events resulted from human factors such as procedural deficiencies and personnel errors. This report is a sequel to our previous report, JAERI-M 91-143 (Analysis of Operating Experience Data in Nuclear Power Plants - Loss of Decay Heat Removal during Reactor Shutdown -). (author)

  13. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment. PMID:16381764

  14. Experimental study on a heat pipe towards in-core decay heat removal control rod

    A novel in-core heat removal concept can be adopted in control rods for passive safety of nuclear power plants. The new concept is featured by a passive heat transfer device called heat pipe and combined with control rod. As the first step for this concept, stainless steel 316L heat pipes were tested in terms of heat removal capability under the same diameter condition with an actual control rod in a typical PWR. It has outer diameter of 3/4 inch (17.4 mm inner diameter), and the length of 1000 mm. Also, the capillary-driven heat pipe was compared with a bare tube with same diameter without wick structures called thermosyphon. As the results, heat transfer coefficients of the heat pipe were ∼34% higher than those of thermosyphon. The results were compared with existing correlations and a CFD analysis. The overall heat transfer characteristics of heat pipes such as thermal resistances were checked for potential uses in terms of in-core heat removal. (author)

  15. Evaluation on decay heat removal capability of isolation condenser for depressurization and cooling system, July 2014

    In this study, we aimed to quantitatively evaluate cooling capacity of isolation condenser (IC) for core cooling in Fukushima Daiichi Nuclear Plant Unit 1. In order to evaluate cooling capacity, decay heat emitted from fuels in reactor core was calculated. We used the data of nuclear fuel and rated thermal power which is released by TEPCO. We analyzed IC with TRAC code. We made an experiment in order to check IC's input data. Because analysis data is close to experimental data, we made sure validity of input data. We analyzed IC under conditions which are recorded when IC started up and tsunami rushed toward plant. As a result of analysis, we confirmed that cooling capacity exceed decay heat under both conditions. In conclusion, IC had a capability to prevent core meltdown accident in Fukushima Daiichi Nuclear Plant Unit 1. (author)

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

    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.

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

    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

  18. Study of Nuclear Decay Data Contribution to Uncertainties in Heat Load Estimations for Spent Fuel Pools

    Ferroukhi, H.; Leray, O.; Hursin, M.; Vasiliev, A.; Perret, G.; Pautz, A.

    2014-04-01

    At the Paul Scherrer Institut (PSI), a methodology for nuclear data uncertainty propagation in CASMO-5M (C5M) assembly calculations is under development. This paper presents a preliminary application of this methodology to C5M decay heat calculations. Applying a stochastic sampling method, nuclear decay data uncertainties are first propagated for the cooling phase only. Thereafter, the uncertainty propagation is enlarged to gradually account for cross-section as well as fission yield uncertainties during the depletion phase. On that basis, assembly heat load uncertainties as well as total uncertainty for the entire pool are quantified for cooling times up to one year. The relative contributions from the various types of nuclear data uncertainties are in this context also estimated.

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

    Lap-Yan Cheng

    2009-01-01

    Full Text Available 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.

  20. Online calculation of the decay heat of assemblies at the Fast Flux Test Facility

    The Fast Flux Test Facility (FFTF) is utilized by the US Department of Energy and the international community as a fast reactor research tool. Its use includes, among other things, the irradiation testing of nuclear reactor fuels and materials required for the development of commercial liquid metal reactors. The decay heat rate of assemblies irradiated in the FFTF is an important parameter in establishing the transportation, examination, and storage of irradiated assemblies. The decay heat program which is maintained on a Cray super computer along with a Symphony speadsheet program running on a personal computer (PC) were created to accommodate this need. This unique synthesis provides a method of combing the capabilities of a mainframe computer with those of a PC

  1. Reliability of the Decay Heat Removal System of Phenix, Concepts and Qualitative Reliability Analysis

    The decay heat removal in the Phenix Reactor has been designed with high level of safety and redundancy so that it should exist at least one way of removal whatever incident would be generated. In the following paragraph, we indicate the principal criteria used in the design, the different ways of operating and the global and qualitative reliability analysis in order to show the high degree of safety

  2. Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

    The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam-Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1-0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using 'CRAFT' software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ≤1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement

  3. An automated software for analysis of experimental data on decay heat from spent nuclear fuel

    Llerena Herrera, Isbel

    2012-01-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has developed a method for final disposal of spent nuclear fuel. This technique requires accurate measurement of the residual decay heat of every assembly. For this purpose, depletion codes as well as calorimetric and gamma-ray spectroscopy experimental methods have been developed and evaluated. In this work a prototype analysis tool has been developed to automate the analysis of both calorimetric and gamma-ray spectroscopy measureme...

  4. Fast neutron reactor safety. Reliability analysis of Phenix decay heat removal function

    This paper presents a reliability analysis of the Phenix decay heat removal function. After summarizing analysis results of sequences leading to failure of this function, the main steps in a probabilistic risk assessment are described: computation of reliability and maintenance parameters for the various systems involved, followed by probabilistic analysis of failure sequences. The extent to which the various systems contribute to the total risk is analyzed. Certain maintenance recommendations were made accordingly

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

    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

  6. Decay heat removal under boiling condition in a pin-bundle geometry

    Decay heat removal capability under boiling condition was investigated using an electrically heated 37-pin bundle test section. The flow was driven by natural circulation force of the out-of-pile sodium loop SIENA in O-arai Engineering Center, PNC. As the heater power was increased, the two-phase flow regime changed from bubbly flow to slug flow and then to annular or annular mist flow. In 15 runs, dry-out was not observed in the average exit quality region of less than 0.5. The results indicated the existance of a large ''boiling window'' for low flow rate and low power conditions. (author)

  7. Testing JEFF-3.1.1 and ENDF/B-VII.1 Decay and Fission Yield Nuclear Data Libraries with Fission Pulse Neutron Emission and Decay Heat Experiments

    Cabellos, O.; de Fusco, V.; Diez de la Obra, C. J.; Martinez, J. S.; Gonzalez, E.; Cano-Ott, D.; Alvarez-Velarde, F.

    2014-04-01

    The aim of this work is to test the present status of Evaluated Nuclear Decay and Fission Yield Data Libraries to predict decay heat and delayed neutron emission rate, average neutron energy and neutron delayed spectra after a neutron fission pulse. Calculations are performed with JEFF-3.1.1 and ENDF/B-VII.1, and these are compared with experimental values. An uncertainty propagation assessment of the current nuclear data uncertainties is performed.

  8. Physical distributions of radon decay chain activities in air

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

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

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

  10. Heating entrepreneur activity in 2003

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

  11. Decay Heat Analyses after Thermal-Neutron Fission of 235U and 239Pu by SCALE-6.1.3 with Recently Available Fission Product Yield Data

    The heat reaches about 1.5% after one hour and falls to 0.4% after a day. After a week it will be about 0.2%. The reactor, however, still requires further cooling for several years to keep the fuel rods safe. In general, the decay heat in the reactors can be calculated using a summation calculation method, which is simply the sum of the activities of the fission products produced during the fission process and after the reactor shutdown weighted by the mean decay energies. Consequently, the method is strongly dependent on the available nuclear structure data. Nowadays, the method has been implemented in various burnup and depletion programs such as ORIGEN and CINDER. In this study, the decay heat measurements after thermal-neutron fission of 235U and 239Pu have been evaluated by the ORIGEN-S with the decay data and fission product yield libraries included in the SCALE-6.1.3 software package. The new libraries were applied to the decay heat calculations, and the results were compared with those by the ORIGEN reference calculation. The decay heat measurements for very short cooling times after thermal-neutron fission of 235U and 239Pu have been evaluated by the ORIGEN-S summation calculation. The reference calculation results by the latest ORIGEN data libraries of the SCALE-6.1.3 have been validated with the measurements by ORNL and Studsvik. In addition, the generation of the new ORIGEN yield libraries has been completed based on the ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0. The new libraries have been successfully applied to the decay heat calculations and comparative analyses have been devoted to verifying the importance of the fission product yield data when estimating the decay heat values for each isotope in a very short time. The decay data library occupies an important position in the ORIGEN summation calculation along with the fission product yield library

  12. Recruitment and activation of mRNA decay enzymes by two ARE-mediated decay activation domains in the proteins TTP and BRF-1

    Lykke-Andersen, Jens; Wagner, Eileen

    2005-01-01

    In human cells, a critical pathway in gene regulation subjects mRNAs with AU-rich elements (AREs) to rapid decay by a poorly understood process. AREs have been shown to directly activate deadenylation, decapping, or 3′-to-5′ exonucleolytic decay. We demonstrate that enzymes involved in all three of these mRNA decay processes, as well as 5′-to-3′ exonucleolytic decay, associate with the protein tristetraprolin (TTP) and its homolog BRF-1, which bind AREs and activate mRNA decay. TTP and BRF-1 ...

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

    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

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

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

  15. Interim storage of solidified fission products from fuel element reprocessing with utilization of obtaining post-decay heat

    It is noted that the out-lined interim store for HRW with industrial utilization of decay heat (production of saturated steam and hydrogen) does include a certain risk potential like any technical plant but that it does not represent a danger to the population living nearby. All internal and external impacts on the store result in safely triggering natural convection cooling. A further emergency cooling system is provided by the water irrigation facility so that obtaining after-heat can be safely removed under all circumstances. Therefore, there are no safety-technology arguments against any realization of the concept presented for interim storage of solidified high-level radio-active wastes. An interim store of this type may be built and operated even in densely populated regions and urban agglomerations. (orig./HP)

  16. FIP Bias Evolution in a Decaying Active Region

    Baker, D.; Brooks, D. H.; Démoulin, P.; Yardley, S. L.; van Driel-Gesztelyi, L.; Long, D. M.; Green, L. M.

    2015-04-01

    Solar coronal plasma composition is typically characterized by first ionization potential (FIP) bias. Using spectra obtained by Hinode’s EUV Imaging Spectrometer instrument, we present a series of large-scale, spatially resolved composition maps of active region (AR)11389. The composition maps show how FIP bias evolves within the decaying AR during the period 2012 January 4-6. Globally, FIP bias decreases throughout the AR. We analyzed areas of significant plasma composition changes within the decaying AR and found that small-scale evolution in the photospheric magnetic field is closely linked to the FIP bias evolution observed in the corona. During the AR’s decay phase, small bipoles emerging within supergranular cells reconnect with the pre-existing AR field, creating a pathway along which photospheric and coronal plasmas can mix. The mixing timescales are shorter than those of plasma enrichment processes. Eruptive activity also results in shifting the FIP bias closer to photospheric in the affected areas. Finally, the FIP bias still remains dominantly coronal only in a part of the AR’s high-flux density core. We conclude that in the decay phase of an AR’s lifetime, the FIP bias is becoming increasingly modulated by episodes of small-scale flux emergence, i.e., decreasing the AR’s overall FIP bias. Our results show that magnetic field evolution plays an important role in compositional changes during AR development, revealing a more complex relationship than expected from previous well-known Skylab results showing that FIP bias increases almost linearly with age in young ARs.

  17. Resonance shielding-factor cross-section processing technique validation based on tungsten decay heat experimental data

    This study presents a method to obtain corrected self-shielded radiative capture cross-sections for tungsten isotopes to be used for activation calculations. The approach used is based on the application of the Bondarenko shielding factor method to the 175-group AMPX master library by means of the Bonami-Nitawl scale-4.3 sequence calculation. The ANITA-4M activation code calculates the tungsten radioisotopes production and the decay heat using the self-shielded cross-sections from ENDF/B-VI, JEF-2.2 and JENDL-3.2 data files. Two irradiation scenarios (5 min and 7 h) in the international thermonuclear experimental reactor (ITER)-like neutron flux spectrum defined by the fusion neutron source experiments are analyzed. The unshielded calculations result in discrepancy with experiment up to 70%, while the self-shielding treatment reduces drastically that discrepancy to less than few percents. In comparison to the experimental integral decay heat values provides a validation of the method used to deal with the self-shielding treatment

  18. Resonance shielding-factor cross-section processing technique validation based on tungsten decay heat experimental data

    Cepraga, D G; Frisoni, M

    2000-01-01

    This study presents a method to obtain corrected self-shielded radiative capture cross-sections for tungsten isotopes to be used for activation calculations. The approach used is based on the application of the Bondarenko shielding factor method to the 175-group AMPX master library by means of the Bonami-Nitawl scale-4.3 sequence calculation. The ANITA-4M activation code calculates the tungsten radioisotopes production and the decay heat using the self-shielded cross-sections from ENDF/B-VI, JEF-2.2 and JENDL-3.2 data files. Two irradiation scenarios (5 min and 7 h) in the international thermonuclear experimental reactor (ITER)-like neutron flux spectrum defined by the fusion neutron source experiments are analyzed. The unshielded calculations result in discrepancy with experiment up to 70%, while the self-shielding treatment reduces drastically that discrepancy to less than few percents. In comparison to the experimental integral decay heat values provides a validation of the method used to deal with the sel...

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

    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)

  20. Development of numerical analysis methods for natural circulation decay heat removal system applied to a large scale JSFR

    A decay heat removal system utilizing passive natural circulation is applied to a large scale Japan Sodium-cooled Fast Reactor. As preparing for the future licensing, a one-dimensional flow network method and a three-dimensional numerical analysis method were developed to evaluate core cooling capability and thermal transient under decay heat removal modes after reactor trip. The one-dimensional method was applied to a water test simulating the primary system of the reactor, while the three-dimensional method was applied to the water test and a sodium test focusing on the decay heat removal system. The numerical results of both methods have turned out to agree well with the test results. And then the thermal-hydraulic behavior under a typical decay heat removal mode of the reactor has been predicted by the three-dimensional method. (author)

  1. Radiotoxicity and decay heat power of spent uranium-plutonium and thorium fuel at long-term storage

    Changes of radiotoxicity and decay heat power of actinides from spent uranium- plutonium and thorium nuclear fuel of WWER-1000 type reactors at storage during 300 years are investigated in report. (author)

  2. Validation of the CFD code NEPTUNE for a full scale simulator for decay heat removal systems with in-pool heat exchangers

    Bassenghi, Federica

    2013-01-01

    In the present work, a multi physics simulation of an innovative safety system for light water nuclear reactor is performed, with the aim to increase the reliability of its main decay heat removal system. The system studied, denoted by the acronym PERSEO (in Pool Energy Removal System for Emergency Operation) is able to remove the decay power from the primary side of the light water nuclear reactor through a heat suppression pool. The experimental facility, located at SIET laboratories (PIACE...

  3. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time

  4. Passive Decay Heat Removal System Options for S-CO{sub 2} Cooled Micro Modular Reactor

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO{sub 2} Brayton power cycle. The S-CO{sub 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{sub 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.

  5. Contribution to decay heat calculation: fission product mean beta and gamma assessment

    Following a reactor shutdown, after the fission chain process has completely faded out, a significant quantity of energy (around seven per cent of the total power of the reactor) continues to be generated in the core. This is known as residual power or decay heat. The principal source of this energy is due to the radioactive decay of fission products and is at any time equal to the sum of the powers released by these different nuclei (P = Σ = Pi). Each of the powers Pi is the product of three terms: the concentration of the relevant nuclide, its decay constant and its mean decay energy. The evaluation of the first two term is straightforward. On the other hand the evaluation of the mean energies presents some difficulties due to a lack of data in beta and gamma spectra of some fission products. This study intends, after a critical analysis of the current method of evaluation of the mean energies, to propose a new model for this calculation. The new model tested on several well known nuclides, has been proved correct and precise. It has then been applied to approximatively sixty nuclides among the lesser known ones. The results obtained have lead to a better prediction of both beta and gamma ray components of the residual power. Consequently, this new model, which allows to take into account the lack of beta branching ratio corresponding to the highest levels of the product nucleus in the beta decay reaction, can be adopted to replace the current method, for calculation of the mean energies of fission products, especially in the case of the lesser known nuclides

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

    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

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

    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.)

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

    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.

  9. Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

    This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto the drywell floor

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

    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)

  11. Radioactive decay products in neutron star merger ejecta: heating efficiency and $\\gamma$-ray emission

    Hotokezaka, Kenta; Tanaka, Masaomi; Bamba, Aya; Terada, Yukikatsu; Piran, Tsvi

    2015-01-01

    The radioactive decay of the freshly synthesized $r$-process nuclei ejected in compact binary mergers power optical/infrared macronovae (kilonovae) that follow these events. The light curves depend critically on the energy partition among the different products of the radioactive decay and this plays an important role in estimates of the amount of ejected $r$-process elements from a given observed signal. We study the energy partition and $\\gamma$-ray emission of the radioactive decay. We show that $20$-$50\\%$ of the total radioactive energy is released in $\\gamma$-rays on timescales from hours to a month. The number of emitted $\\gamma$-rays per unit energy interval has roughly a flat spectrum between a few dozen keV and $1$ MeV so that most of this energy is carried by $\\sim 1$ MeV $\\gamma$-rays. However at the peak of macronova emission the optical depth of the $\\gamma$-rays is $\\sim 0.02$ and most of the $\\gamma$-rays escape. The loss of these $\\gamma$-rays reduces the heat deposition into the ejecta and h...

  12. Quantification of modeling uncertainties based on scaling laws in natural circulation decay heat removal

    A Best Estimate Plus Uncertainty (BEPU) analysis is one of the good methods to estimate the uncertainty of phenomenon in a nuclear power plant dynamics. In BEPU analysis, a number of numerical analyses, in which input parameters are varied based on their probabilistic distributions, are carried out to obtain statistical characteristics of the output result. In general, the uncertainty of input parameters, such as a probabilistic distribution form and variance, are estimated based on experimental knowledge and/or engineering judgment. In the present research, we focus on a scaling law (dimensionless number) in constitutive equations from a view point of phenomenological theory. An influence of uncertainty in the dimensionless number and its dependency on BEPU analysis has been investigated. Plant dynamics analyses of Super Safe, Small and Simple (4S) reactor, being developed by Toshiba, are carried out under a natural circulation decay heat removal condition. In the analysis, uncertainties of the dimensionless numbers such as Nusselt, Reynolds, and Prandtl numbers are taken into consideration, as well as an uncertainty of decay heat power. The Latin Hypercube Sampling is applied to determine the input deck set. As a result, it is demonstrated that the parameter dependency on the output result can be revealed by using the dimensionless numbers. (author)

  13. VVER-1000/V320 decay heat analysis, involving TVS-M and TVS-A fuel assemblies

    Petkov, P.V.; Hristov, D.V. [Kozloduy NPP, Vratsa (Bulgaria)

    2007-07-01

    MELCOR-1.8.4 is an integrated computer code, developed for severe accident calculations. It is used primarily for simulation of PWR and BWR types of reactors since the code includes an internal database, suitable for modeling of their cores inventory. Despite similarity between VVER-1000/V320 and PWR, there is still required accounting of specificities of Russian reactor design. MELCOR distinguishes 15 classes, each of them containing chemical elements with similar properties, 12 are involved in radioactive product decay. Part of it appears in the simulation of core decay heat rate after shutdown. In this paper the authors present two reactor core loadings corresponding to fuel assemblies: TVS-M and TVS-A. They have calculated decay heat after reactor shutdown from 100% and 104% of nominal power by SCALE 4.4a package. They also have estimated the amount of generated nuclides. The Newly developed Core Inventory Estimation Tool (CIET), described in this paper, written and tested previously, has been used for the evaluation of core decay heat fractions, distributed over chemical classes. Twelve curves were generated, following the same numerical procedure, implemented in MELCOR for representation of decay heat in W/kg. Comparison of curves shows deviations from the expectations. The total amount of radionuclides, separated in chemical classes, given in MELCOR Computer Code User's was compared to the ones involved in default MELCOR decay heat calculation. The results have confirmed that neglected chemical elements give 0.9% and 1.1% of total core mass for TVS-M and TVS-A and correspondingly 0.5% and 0.6% of total core decay heat.

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

    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 235U 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

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

    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)

  16. Radioactive decay products in neutron star merger ejecta: heating efficiency and γ-ray emission

    Hotokezaka, K.; Wanajo, S.; Tanaka, M.; Bamba, A.; Terada, Y.; Piran, T.

    2016-06-01

    The radioactive decay of the freshly synthesized r-process nuclei ejected in compact binary mergers powers optical/infrared macronovae (kilonovae) that follow these events. The light curves depend critically on the energy partition among the different decay products and it plays an important role in estimates of the amount of ejected r-process elements from a given observed signal. We show that 20-50 per cent of the total radioactive energy is released in γ-rays on time-scales from hours to a month. The number of emitted γ-rays per unit energy interval has roughly a flat spectrum between a few dozen keV and 1 MeV so that most of the energy is carried by ˜1 MeV γ-rays. However, at the peak of macronova emission the optical depth of the γ-rays is ˜0.02 and most of the γ-rays escape. The loss of these γ-rays reduces the heat deposition into the ejecta and hence reduces the expected macronova signals if those are lanthanides dominated. This implies that the ejected mass is larger by a factor of 2-3 than what was previously estimated. Spontaneous fission heats up the ejecta and the heating rate can increase if a sufficient amount of transuranic nuclei are synthesized. Direct measurements of these escaping γ-rays may provide the ultimate proof for the macronova mechanisms and an identification of the r-process nucleosynthesis sites. However, the chances to detect these signals are slim with current X-ray and γ-ray missions. New detectors, more sensitive by at least a factor of 10, are needed for a realistic detection rate.

  17. Natural-circulation decay heat removal from an SP-100, 550-kWe power system for a lunar outpost. Final Report

    This research investigated the decay heat removal from the SP-100 reactor core of a 550-kWe power system for a lunar outpost by natural circulation of lithium coolant. A transient model that simulates the decay heat removal loop (DHRL) of the power system was developed and used to assess the system's decay heat removal capability. The effects of the surface area of the decay heat rejection radiator, the dimensions of the decay heat exchanger (DHE) flow duct, the elevation of the DHE, and the diameter of the rise and down pipes in the DHRL on the decay heat removal capability were examined. Also, to determine the applicability of test results at earth gravity to actual system performance on the lunar surface, the effect of the gravity constant (1 g and 1/6 g) on the thermal behavior of the system after shutdown was investigated

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

    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

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

    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

  20. Dual active surface heat flux gage probe

    Liebert, Curt H.; Kolodziej, Paul

    1995-02-01

    A unique plug-type heat flux gage probe was tested in the NASA Ames Research Center 2x9 turbulent flow duct facility. The probe was fabricated by welding a miniature dual active surface heat flux gage body to the end of a hollow metal cylindrical bolt containing a metal inner tube. Cooling air flows through the inner tube, impinges onto the back of the gage body and then flows out through the annulus formed between the inner tube and the hollow bolt wall. Heat flux was generated in the duct facility with a Huels arc heater. The duct had a rectangular cross section and one wall was fabricated from 2.54 centimeter thick thermal insulation rigid surface material mounted onto an aluminum plate. To measure heat flux, the probe was inserted through the plate and insulating materials with the from of the gage located flush with the hot gas-side insulation surface. Absorbed heat fluxes measured with the probe were compared with absorbed heat fluxes measured with six water-cooled reference calorimeters. These calorimeters were located in a water-cooled metal duct wall which was located across from the probe position. Correspondence of transient and steady heat fluxes measured with the reference calorimeters and heat flux gage probe was generally within a satisfactory plus or minus 10 percent. This good correspondence was achieved even though the much cooler probe caused a large surface temperature disruption of 1000K between the metal gage and the insulation. However, this temperature disruption did not seriously effect the accuracy of the heat flux measurement. A current application for dual active surface heat flux gages is for transient and steady absorbed heat flux, surface temperature and heat transfer coefficient measurements on the surface of an oxidizer turbine inlet deflector operating in a space shuttle test bed engine.

  1. Reactor Decay Heat in 239Pu: Solving the γ Discrepancy in the 4-3000-s Cooling Period

    The β feeding probability of 102,104,105,106,107Tc, 105Mo, and 101Nb nuclei, which are important contributors to the decay heat in nuclear reactors, has been measured using the total absorption technique. We have coupled for the first time a total absorption spectrometer to a Penning trap in order to obtain sources of very high isobaric purity. Our results solve a significant part of a long-standing discrepancy in the γ component of the decay heat for 239Pu in the 4-3000 s range.

  2. Reactor decay heat in 239Pu: solving the γ discrepancy in the 4-3000-s cooling period.

    Algora, A; Jordan, D; Taín, J L; Rubio, B; Agramunt, J; Perez-Cerdan, A B; Molina, F; Caballero, L; Nácher, E; Krasznahorkay, A; Hunyadi, M D; Gulyás, J; Vitéz, A; Csatlós, M; Csige, L; Aysto, J; Penttilä, H; Moore, I D; Eronen, T; Jokinen, A; Nieminen, A; Hakala, J; Karvonen, P; Kankainen, A; Saastamoinen, A; Rissanen, J; Kessler, T; Weber, C; Ronkainen, J; Rahaman, S; Elomaa, V; Rinta-Antila, S; Hager, U; Sonoda, T; Burkard, K; Hüller, W; Batist, L; Gelletly, W; Nichols, A L; Yoshida, T; Sonzogni, A A; Peräjärvi, K

    2010-11-12

    The β feeding probability of (102,104,105,106,107)Tc, 105Mo, and 101Nb nuclei, which are important contributors to the decay heat in nuclear reactors, has been measured using the total absorption technique. We have coupled for the first time a total absorption spectrometer to a Penning trap in order to obtain sources of very high isobaric purity. Our results solve a significant part of a long-standing discrepancy in the γ component of the decay heat for 239Pu in the 4-3000 s range. PMID:21231223

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

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

  4. EPRI/WOG analysis of decay heat removal risk at Point Beach

    This paper provides a best-estimate probabilistic analysis of Decay Heat Removal (DHR) risk at the Point Beach nuclear power plant. It includes an evaluation of potential plant modifications proposed by Sandia National Laboratories as pare of the NRC's Unresolved Safety Issue program on DHR requirements (USI A-45). The EPRI/WOG analysis yielded a factor of thirty lower core-damage frequency for the sequences included in the scope of the Sandia study. This analysis also yielded a factor of seven reduction in off-site consequences (over and above the core-damage frequency reduction), and estimates of costs that are 50-400% higher for the various proposed backfit modifications. This evaluation, like the Sandia study, concludes that a dedicated SDHR system is not justifiable for Point Beach on a cost-benefit basis

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

    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

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

    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.

  7. Analysis of integral circulation and decay heat removal experiments in the lead-bismuth CIRCE facility with RELAP5 code

    In this paper, the results of the post-test analysis of some integral circulation experiments conducted on the lead-bismuth CIRCE facility are presented in comparison with the experimental data. These experiments include the simulation of unprotected loss of flow and unprotected loss of heat sink transients in a pool-type heavy liquid metal reactor. Furthermore, the results of the pre-test analysis of a protected loss of heat sink and flow transient with decay heat removal by a heat exchanger immersed in the pool and operating in natural circulation is presented. All transient analyses have been performed with the RELAP5 thermal-hydraulic code. (author)

  8. Impact of decay heat in concrete on water release/H2 production and its implications to LMRs

    The effects of fission product decay gamma energy deposition in concrete on the rate of water release from concrete is investigated. The results indicate a significant increase in the water release rate when gamma heating is included as compared to surface thermal heating only. Because such water release can lead to hydrogen production, it is concluded that this phenomena should be included in the analysis of the low probability severe accident when appropriate. (author)

  9. CdWO4 scintillating bolometer for Double Beta Decay: Light and Heat anticorrelation, light yield and quenching factors

    Arnaboldi, C.; Beeman, J.W.; Cremonesi, O.; Gironi, L.; M. Pavan; Pessina, G.(Sezione INFN di Milano Bicocca, Milan, Italy); Pirro, S.(INFN-Laboratori Nazionali del Gran Sasso, Assergi, 67010 , L’Aquila, Italy); Previtali, E.

    2010-01-01

    Abstract We report the performances of a 0.51 kg CdWO4 scintillating bolometer to be used for future Double Beta Decay Experiments. The simultaneous read-out of the heat and the scintillation light allows to discriminate between different interacting particles aiming at the disentanglement and the reduction of background contribution, key issue for next generation experiments. We will describe the observed anticorrelation between the heat and the light signal and we will show how t...

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

    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

  11. Presentation of decay heat removal computer codes used for gas cooled reactors

    For the existing French Magnox type reactors, two computer codes have been developed to analyze the transient after reactor shutdown: - The first one ('GITA', is representative of the short term evolution (less than 2 days) and it includes a refined representation of all the reactor components. - The second one, 'LOTE', has been developed to represent the long term evolution (from 2 days to several months) with a simplified representation of the main components of the reactor. One example of accident simulation is presented for existing Magnox reactor. Moreover, as a part of the French program on the future reactors, an analysis of the modular high temperature has been initiated. 2D and 3D general flow and conduction codes are used for this analysis: - DELFINE is a 2D conduction code including a 1D thermosyphon model, it has been used for decay heat removal analysis. TRIO is a 3D flow code including 3D radiation, conduction and convection heat transfer. It is used for detailed thermal analysis during accidental conditions

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

    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)

  13. CFD analysis of decay heat removal scenarios of the lead cooled ELSY reactor

    The lead cooled European reactor concept ELSY is characterized by its innovative, compact design, where all components of the primary loops are located inside the reactor vessel. The vessel includes 8 steam generators and pumps which generate a coolant flux of 126 tons/s. At nominal operation conditions the core releases about 1500 MW converted to an electric power of 600 MW. For the coolant temperatures 400 C at the core inlet and 480 C are envisaged. If the reactor is shut down and the pumps are switched off, the decay heat is removed by isolation condenser (IC) systems acting on the SG secondary circuits. If the IC's are not available, heat removal by 4 dip coolers is foreseen, which can be operated by gravitation driven water flow or by air. Additionally the outer vessel wall is permanently cooled by a RVACS (reactor vessel air cooling) system located between the outer vessel wall and the reactor cavity. The main intention of this work is the investigation of the passive cooling systems which are used for the decay heat removal. The CFD vessel model of about 20 million cells takes advantage of the components symmetry and simulates a 90 deg. section of the reactor. The spatial resolution of the computational grid varies between 5 mm close to walls and 100 mm in undisturbed regions where only small gradients are expected. The core, the pumps and the SG's are simulated with porous media models including volumetric source terms for momentum and energy. For the SG's detailed CFD studies with at least one order of magnitude finer grid resolution are performed in order to obtain data for pressure losses. Solids like pipe walls or the core barrel are taking into account by heat conduction. As the model considers a closed system the coolant flux is controlled by momentum sources of the pumps and frictional losses mainly by the core and the SG's. The simulations are performed as single phase flows, therefore the free lead surface at the unclosed upper part of the vessel

  14. Integrated assessment approach to ensure removal of decay heat with RCW out of service

    In the unusual event that Recirculated Cooling Water System of a CANDU NPP is drained for maintenance or generally incapacitated and the reactor is in the Guaranteed Shutdown State, an alternative source of cooling water is required to remove decay heat. For the Cernavoda Unit 1 NPP, CNE-PROD assessed and proposed a novel arrangement to address this issue, prepared a design, and requested AECL to assess the proposed approach. As part of the assessment, AECL developed and performed an integrated approach that brought together a variety of design, operational, maintenance and safety considerations. First, a simplified hydraulic analysis was performed, the heat removal requirements of the system were reviewed and alternative heat sinks were evaluated. Then, the maintenance and operational implications were considered in detail, using AECL's Systematic Assessment of Maintenance (SAM) process. This included a detailed assessment of failure modes of interest and associated effects (Failure Modes and Effects Analysis). This assessment included consideration of the impact on operational chemistry. In addition, the safety implications (potential for human error and safety consequences of failure) were reviewed. Based upon the system requirements, hydraulic assessment, and maintenance review, several alternatives or improvements to the CNE-PROD proposal were developed. In addition, a number of operational suggestions, as well as additional concerns and considerations were identified. Finally, potential alternative solutions were looked at based upon previous AECL experience. This paper will describe the novel CNE-PROD arrangement and provide an overview of the integrated assessment approach taken by AECL (using Plant Life Management technologies) to evaluate the proposed arrangement and to identify additional considerations. (author)

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

    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)

  16. Neutronic and thermo-hydraulic analyses of a small, long-life HTGR for passive decay-heat removal

    Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011, design concepts for nuclear reactors have been reconsidered with much greater emphasis placed upon passive systems for decay-heat removal. By considering this issue, the design parameter conditions for high temperature gas-cooled reactors (HTGRs) with passive safety features of decay-heat removal were obtained by residual-heat transfer calculation using equations for fundamental heat transfer mechanisms in our previous works. In the present study, the appropriate size of reactor core for a 100 MWt reactor operating at 1123 K of the initial core temperature was found using the conditions. Consequently, neutronics and thermo-hydraulic analyses for the proposed reactor core were performed and the proper optimizations to control the excess reactivity and flatten the change in power peaking factor during operation were done successfully. By the systematic method to decide the core design which satisfies the condition for passive decay-heat removal, a long-life small HTGR concept whose excess reactivity was small during the operation was shown. The small excess reactivity is a significant advantage from the view point of safety in reactivity accident. (author)

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

    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)

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

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

    2016-08-01

    We present the analysis of a decaying active region observed by the EUV Imaging Spectrometer on Hinode during 2009 December 7–11. We investigated the temporal evolution of its structure exhibited by plasma at temperatures from 300,000 to 2.8 million degrees, and derived the electron density, differential emission measure, effective electron temperature, and elemental abundance ratios of Si/S and Fe/S (as a measure of the First Ionization Potential (FIP) Effect). We compared these coronal properties to the temporal evolution of the photospheric magnetic field strength obtained from the Solar and Heliospheric Observatory Michelson Doppler Imager magnetograms. We find that, while these coronal properties all decreased with time during this decay phase, the largest change was at plasma above 1.5 million degrees. The photospheric magnetic field strength also decreased with time but mainly for field strengths lower than about 70 Gauss. The effective electron temperature and the FIP bias seem to reach a “basal” state (at 1.5 × 106 K and 1.5, respectively) into the quiet Sun when the mean photospheric magnetic field (excluding all areas physical properties are all positively correlated with each other and the correlation is the strongest in the high-temperature plasma. Such correlation properties should be considered in the quest for our understanding of how the corona is heated. The variations in the elemental abundance should especially be considered together with the electron temperature and density.

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

    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

  20. Shutdown Decay Heat Removal analysis of a Westinghouse 3-loop pressurized water reactor: Case study

    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 Westinghouse 3-loop 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. Decay heat removal and operator's intervention during a very small LOCA

    Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA (< or approx.0.05 ftsup(2)). For a water-side break with the break size larger than 0.006ftsup(2), fluid-loss through break exceeds the makeup. If the break sizeis larger than 0.008ftsup(2), decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 9000MWe or 1200MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervension was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system. (Author)

  2. Design Of Rectifier Controller For Decay Heat Simulation Using Thermalhydraulic Test Loop

    The reactor thermalhydraulic tast loop permits to study experimentally the thermalhydraulic phenomena in a PWR or PHWR nuclear reactor. The system is design to operate in normal mode, hence the power supplied to the test section will loose totally as soon as the system is shutdown. In this research, the power control system has been modified in order to be able to simulate the decay heat phenomena. The power setting circuit has been modified from the potentiometer circuit to another one using DAC and computer program. This modification is to overcome the slow setting changes caused by the limit of the motor rotation. The test results an absolute error 6.25 kW, that corresponds to the DAC conversion error 0.5 LSB. This error is less than any other general method, that provides error conversion 1 LSB. In term of electrical power, the error value depends on the setting configuration adapted to the maximum power. In this work, the maximum power value is 3.2 MW, which corresponds to the nominal power of the installation. Besides its better precision, this modification provides also a higher speed then the previous system using potentiometer. The smallest sampling time chosen 10 ms is much greater then the DAC time conversion, so the analog signal still well correspond to its digital value

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

    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)

  4. Numerical investigation of multidimensional natural circulation phenomenon in passive safety systems for decay heat removal in large pools

    Many advanced designs of nuclear reactors adopt a methodology of passive safety systems in which the decay heat generated from a reactor is transferred by natural circulation into large pool of water called the Gravity Driven Water Pool (GDWP). Three-dimensional convection flows develop, which in turn affect the heat transfer process and hence the temperature pattern. The heat transfer process can get compromised by the possible stratification of the temperature. The objective of this study is to investigate the transient multidimensional natural circulation phenomenon in GDWP having total volume of 9247 m3 using open source CFD code (OpenFOAM-2.2). In order to reduce the thermal stratification, various geometrical modifications have been incorporated on the heat exchanger design, such as distributing and submergence of heat source; provision of passive elements such as draft tube. A detailed CFD analysis has confirmed the mitigation of thermal stratification phenomenon. (author)

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

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

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

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

    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.

  8. Radiotoxicity and decay heat power of actinides and fission products of spent nuclear fuel of WWER type reactors at long-term storage

    Time dependence of a power and radiotoxicity of spent nuclear fuel of WWER-1000 reactors at its long-term storage or uniform accumulation in long-term storage on basis of 1 GW of electrical power is investigated. At calculations of decay heat power, the contributions of alpha-, beta-, and gamma- irradiation were taken into account, at calculations of a radiotoxicity - maximum permissible activity of nuclides in air and in water were taken into account. The submitted data can be used by choice of strategy of a long-term storage of spent nuclear fuel of power reactors. (author)

  9. The effect of air leakage and heat exchange on the decay of entrapped air pocket slamming oscillations

    Abrahamsen, Bjørn C.; Faltinsen, Odd M.

    2011-10-01

    The phenomenon studied in this work is that of an air pocket entrapped by a free surface water wave inside a rectangular tank at a high filling level. The wave, which is a gravity wave, is caused by forced horizontal motion which is constructed in a particular way, in order to entrap an air pocket as it approaches the upper left corner of the tank. As the wave touches the roof, the air is compressed and starts to oscillate. The oscillations resemble, to some extent, the free oscillations of an underdamped mass-spring system, where the mass is related to the generalized added mass effect of the water pressure associated with the air pocket oscillations. The stiffness is due to the compressibility of the air. The reason for the damping or, more generally, the decay of the air pocket oscillations is less understood. Air leakage has been proposed as one possible reason for this decay. In this work, the role of air leakage is found not to be the reason for the decay of the air pocket oscillations, because it is not present during major parts of the impact. However, by drilling holes in the roof of the tank, the effect of leakage during the oscillations is proven to cause decay. To explain the physical source of the decay of the oscillations, damping due to heat transfer to and from the air pocket is investigated through an analytical one-dimensional steady-state model. The damping due to heat transfer is observed to play an important role. The obtained understanding of the mechanisms causing the decay of the air-pocket impact at the upper corner is believed to be relevant to other types of impacts, particularly the entrapment of air pockets on walls by breaking waves.

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

    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)

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

    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.

  12. Report on the activities of the decay data evaluation project

    This report summarizes the work of the DDEP collaboration since its establishment in 1994. The aim of this group is to evaluate and propose decay data for approximately 250 radionuclides of interest in various applications. It also presents a projected schedule of nuclear data evaluations for 2002-2003, and the minutes of the DDEP Meeting held at Braunschweig, Germany, May 15, 2001. A sample of a DDEP evaluation is presented in this report. (authors)

  13. Low-latitude coronal holes, decaying active regions and global coronal magnetic structure

    Petrie, Gordon

    2013-01-01

    We study the relationship between decaying active region magnetic fields, coronal holes and the global coronal magnetic structure using Global Oscillations Network Group (GONG) synoptic magnetograms, Solar Terrestrial RElations Observatory (STEREO) extreme ultra-violet (EUV) synoptic maps and coronal potential-field source-surface (PFSS) models. We analyze 14 decaying regions and associated coronal holes occurring between early 2007 and late 2010, four from cycle 23 and 10 from cycle 24. We investigate the relationship between asymmetries in active regions' positive and negative magnetic intensities, asymmetric magnetic decay rates, flux imbalances, global field structure and coronal hole formation. Whereas new emerging active regions caused changes in the large-scale coronal field, the coronal fields of the 14 decaying active regions only opened under the condition that the global coronal structure remained almost unchanged. This was because the dominant slowly-varying, low-order multipoles prevented opposin...

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

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

  15. Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The DHRS of JSFR consists of one unit of DRACS (direct reactor auxiliary cooling system), which has a dipped heat exchanger in the reactor vessel and two units of PRACS, which has a heat exchanger in a primary-side inlet plenum of IHX in each loop. In this study, the sodium experiments were conducted using a sodium test loop PLANDTL in order to investigate the effect of operation mode on transient behavior of thermal hydraulic in PRACS loop. The experimental results revealed the effect of increasing heat removal capacity of PRACS and the forced flow operation in PRACS loop on the thermal transient in the PRACS loop and natural circulation behavior of PRACS. (author)

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

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

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

    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 oC which is substantially lower than ∼627 oC 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. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    Hung, T.C., E-mail: tchung@ntut.edu.t [Department of Mechanical Engineering, National Taipei University of Technology, 1, Sec. 3, Chung-hsiao E. Rd., Taipei 10608, Taiwan (China); Dhir, V.K. [Department of Mechanical and Aerospace Engineering, UCLA, CA (United States); Chang, J.C. [Graduate Institute of Mechanical and Electrical Engineering, National Taipei University of Technology, Taiwan (China); Wang, S.K. [Department of Mechanical and Automation Engineering, I-Shou University, Taiwan (China)

    2011-01-15

    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 {approx}551 {sup o}C which is substantially lower than {approx}627 {sup 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

  19. Evaluation of the heat-storage capability of shallow aquifers using active heat tracer tests and Fiber-Optics Distributed-Temperature-Sensing

    Suibert Oskar Seibertz, Klodwig; Chirila, Marian Andrei; Bumberger, Jan; Dietrich, Peter; Vienken, Thomas

    2015-04-01

    In the course of the energy transition, geothermal energy storage and heat generation and cooling have proven to be environmental friendly alternatives to conventional energy. However, to ensure sustain usage, the heat transport behavior of aquifers and its distribution has to be studied. A tool to achieve this is the active heat tracer test, eg. Leaf et al. (2012). If active heat tracer tests are combined with in aquifer heat testing via electric heating-cables, eg. Liu et al. (2013), it is possible to observe heat transport and temperature signal decay without disturbing the original pressure field within the aquifer. In this field study a two channel High-Resolution-Fiber-Optic-Distributed-Temperature-Sensing and Pt100 were used to measure temperature signals within in two wells of 1.4 m distance, where the temperature difference was generated using a self regulating heating cable in the upstream well. High resolution Distributed-Temperature-Sensing measurements were achieved by coiling the fiber around screened plastic tubes. The upstream well was also used to observe heating (Δ Tmax approx. 24K) and temperature signal decay, while the downstream well was used to observe heat transport between both wells. The data was analyzed and compared to thermal conductivity of soil samples and Direct-Push (DP) Electrical-Conductivity-Logging and DP Hydraulic-Profiling results. The results show good agreement between DP data and temperature measurements proving the active heat tracer test is a suitable tool for providing reliable information on aquifer heat-storage capability. References Leaf, A.T., Hart, D.J., Bahr, J.M.: Active Thermal Tracer Tests for Improved Hydrostratigraphic Characterization. Ground Water, vol. 50, 2012 Liu, G., Knobbe, S., Butler, J.J.Jr.: Resolving centimeter-scale flows in aquifers and their hydrostratigraphic controls. Geophysical Research Letters, vol. 40, 2013

  20. Efficacy and Pertinence 01 Heat Treatments against Monilia Decay in Commerciai Condition

    Warlop, Franç,ois

    2002-01-01

    Post-harvest decay can lead to very important damages on stone fruits. especially in the Rhone valley. Hot water treatments are known to be efficient against many fruit pathogens including monilia, but industriai application remains difficult, since it can present some risks according to fruit maturity, thermostat accuracy of the engine We went on trying this alternative on peach decay, by testing a prototype lend by Xeda, in a professional station. Several parameters were to be eval...

  1. Preliminary Test Requirements for the Performance Test of Passive Decay Heat Removal System of Sodium-Cooled Fast Reactor

    In order to verify the concept of safety grade passive decay removal system PDRC (Passive Decay heat Removal Circuit) of KALIMER-600 and the design features to resolve the design issues for securing the cooling performance, the performance test is implemented. In this report, the preliminary test requirements for using as a guideline to the design of the experimental facility were established. Since the experimental facility should be designed so as to simulate the various thermal- hydraulic phenomena, as closely as possible, to be occurred in reference reactor during the decay heat removal operation, the design characteristics of the reference reactor (KALIMER-600) were analyzed for drawing major constitutive elements to be simulated in the facility. The preliminary test matrix was set up by the analysis of various design basis events and then the key test matrix was determined. Also, the priority for various thermal hydraulic phenomena which should be considered in the design of the experimental facility was determined by analyzing the phenomena for each key test matrix. Based on the analysis, the general design requirements for experimental facility were prepared and the design requirements for fluid systems and instrumentation were established. The test requirements in this report will be reflected in the scaling analysis and the basic design of the experimental facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  2. Decay heat removal by natural circulation in a 550-kW (electric) SP-100 power system for a lunar outpost

    In this paper, the natural-circulation decay heat removal capability of a 550-kW(electric) SP-100 reactor power system for a lunar outpost is investigated. A transient thermal-hydraulic model of the decay heat removal loop (DHRL) is developed to investigate the effects of the radiator surface area, the dimensions and elevation of the decay heat exchanger (DHE), and the diameter of the rise and down pipes on the passive decay heat removal of the system. The effect of gravity is also investigated in order to examine the applicability of earth-based test results to the actual system on the lunar surface. Results show that natural circulation of lithium coolant in the DHRL would keep the SP-100 reactor safely coolable after shutdown. However, the lithium coolant in the adiabatic rise pipe, directly downstream from the reactor core, could overheat by as much as 175 K above its nominal operation value of 1355 K at ∼ 200 s after shutdown. This coolant temperature increase can be reduced by as much as 50 K by increasing the height of the DHE duct to 15 cm; a further increase in the duct height would have little effect n the decay heat removal. Increasing the elevation of the DHE slightly improves the decay heat removal

  3. The AU-rich element mRNA decay-promoting activity of BRF1 is regulated by mitogen-activated protein kinase-activated protein kinase 2

    Maitra, Sushmit; Chou, Chu-Fang; Luber, Christian A.; Lee, Kyung-Yeol; Mann, Matthias; Chen, Ching-Yi

    2008-01-01

    Regulated mRNA decay is a highly important process for the tight control of gene expression. Inherently unstable mRNAs contain AU-rich elements (AREs) in the 3′ untranslated regions that direct rapid mRNA decay by interaction with decay-promoting ARE-binding proteins (ARE-BPs). The decay of ARE-containing mRNAs is regulated by signaling pathways that are believed to directly target ARE-BPs. Here, we show that BRF1 involved in ARE-mediated mRNA decay (AMD) is phosphorylated by MAPK-activated p...

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

    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 N2 gas for the comparison. The results of the experiments shows very clear features of self control characteristics. The self control mechanism was made clear, that is, in such system in which the condensing area in the condenser expands automatically in accordance with the increase of the heat input to keep the system temperature nearly constant. The working temperature of the system are clearly dependent on the pressure of the non-condensable gas initially enclosed, with higher system working temperature with higher initial gas pressure enclosed. The analyses were done on water and methanol as the working fluids, which show very good agreement with the experimental results. A lot of attractive applications are expected including the self switching feature with minimum heat loss during normal operation with maintaining the sufficient heat removal at accidents. (author)

  5. Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The DHRS of JSFR consists of one unit of DRACS (direct reactor auxiliary cooling system), which has a dipped heat exchanger in the reactor vessel and two units of PRACS, which has a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. The transient experiments for the start-up of DHRS loop showed that quick increase of natural draft in the air duct followed by smooth increase of sodium flow rate in the DHRS loop. Influences of the pressure loss coefficients in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS, respectively due to recovery of natural circulation head via the increase of temperature difference in each loop. (author)

  6. Experimental evaluation of decrease in the activities of polyphosphate/glycogen-accumulating organisms due to cell death and activity decay in activated sludge.

    Hao, Xiaodi; Wang, Qilin; Cao, Yali; van Loosdrecht, Mark C M

    2010-06-15

    Decrease in bacterial activity (biomass decay) in activated sludge can result from cell death (reduction in the amount of active bacteria) and activity decay (reduction in the specific activity of active bacteria). The goal of this study was to experimentally differentiate between cell death and activity decay as the cause of decrease in bacterial activity. By means of measuring maximal anaerobic phosphate release rates, verifying membrane integrity by live/dead staining and verifying presence of 16S rRNA with fluorescence in situ hybridization (FISH), the decay rates and death rates of polyphosphate-accumulating organisms (PAOs) in a biological nutrient removal (BNR) system and a laboratory phosphate removing sequencing batch reactor (SBR) system were determined, respectively, under famine conditions. In addition, the decay rate and death rate of glycogen-accumulating organisms (GAOs) in a SBR system with an enrichment culture of GAOs were also measured under famine conditions. Hereto the maximal anaerobic volatile fatty acid uptake rates, live/dead staining, and FISH were used. The experiments revealed that in the BNR and enriched PAO-SBR systems, activity decay contributed 58% and 80% to the decreased activities of PAOs, and that cell death was responsible for 42% and 20% of decreases in their respective activities. In the enriched GAOs system, activity decay constituted a proportion of 74% of the decreased activity of GAOs, and cell death only accounted for 26% of the decrease of their activity. PMID:20178124

  7. A model for the analysis of loss of decay heat removal accident in MTR pool type research reactors

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. In such conditions, a core overheat take place, and the heat is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a three dimensional geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the time and space dependent non-linear partial differential fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding. (author)

  8. Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF

    Highlights: ► Monte Carlo design of reactor facilities. ► Neutron coupling assessment between critical core and fresh fuel in the storage vessels. ► Power contribution by induced fission from neutrons leaving the core, spontaneous fission and (α, n) sources. ► Power decay heat estimation for different reactor fuel cycles scenarios. ► Material damage assessment in the storage vessels. - Abstract: The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental-scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facilities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall system that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Several parameters were analyzed, like the criticality behavior (namely the Keff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed.

  9. Development and testing of heat transport fluids for use in active solar heating and cooling systems

    Parker, J. C.

    1981-01-01

    Work on heat transport fluids for use with active solar heating and cooling systems is described. Program objectives and how they were accomplished including problems encountered during testing are discussed.

  10. Preliminary design activities for solar heating and cooling systems

    1978-01-01

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

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

    Frisoni Manuela

    2016-01-01

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

  12. Development of margin assessment methodology of decay heat removal function against external hazards. (1) Project overview and snow PRA methodology

    This paper describes mainly snow probabilistic risk assessment (PRA) methodology development in addition to the project overview. This project addresses extreme weathers (snow, tornado, wind and rainfall), volcanic phenomena and forest fire as representative external hazards. In this project, the methodologies of both PRA and margin assessment are developed for each external hazard through external hazard and accident sequence evaluations mainly in terms of decay heat removal function of a sodium-cooled fast reactor (SFR). Using recent 50 year weather data at a typical Japanese SFR site, snow hazard categories were set the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the accident sequence was evaluated by producing event trees which consist of several headings representing the loss of decay heat removal. Snow removal action and manual operation of the air cooler dampers were introduced into the event tree as accident managements. In this paper, the snow PRA showed less than 10-6/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1−2 m/day of snowfall speed and 0.5−0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution to secure the access routes. (author)

  13. Uncertainty evaluation of reliability of safety grade decay heat removal system of Indian prototype fast breeder reactor

    Highlights: • Uncertainty analysis of failure frequency of SGDHRS of a medium sized fast reactor is studied. • Lognormal distribution of failure rate of components is taken with error factor of 3. • The error factor in the distribution of failure frequency in most cases is 3. • The relative importance of the safety components is brought out. - Abstract: Deterministic and probabilistic safety assessment of nuclear power reactor technology is very important in assuring that the design is robust and safety systems perform as per requirement. The parameters required as input data for such analysis have uncertainties associated with them. Their impact is to be assessed on the results obtained for such analyses and it affects the overall decision making process. Safety Grade Decay Heat Removal System (SGDHRS) is one of the safety systems in fast breeder reactors and itremoves decay heat after reactor shutdown. It is a critical safety system; hence failure frequency for SGDHR is targeted to be less than 1.0 × 10−7 per reactor year. By bringing diversity in some of the components of SGDHRS, such as sodium-to-sodium decay heat exchanger (DHX), sodium to air heat exchanger (AHX) and valves, one can achieve the targeted low failure frequency of SGDHRS. We perform uncertainty analysis of the reliability of such SGDHRS here. Uncertainty in failure rate (of components of SGDHRS) is assumed to follow the log-normal distribution with error factor of three. Monte Carlo method of sampling is used in MATLAB environment. Results are obtained in terms of mean, median and standard deviation values of failure frequency. Percentile and confidence interval analysis of mean values are also obtained. These provide 95 and 98 percentile and confidence interval values of 98%, 99% and 99.8%. It is found that error factor of failure frequency of SGDHRS is found to be less than 3 in all the cases except the one in which DHX, AHX and Valves are designed with diversity in design. It is to

  14. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  15. Heat dissipation guides activation in signaling proteins.

    Weber, Jeffrey K; Shukla, Diwakar; Pande, Vijay S

    2015-08-18

    Life is fundamentally a nonequilibrium phenomenon. At the expense of dissipated energy, living things perform irreversible processes that allow them to propagate and reproduce. Within cells, evolution has designed nanoscale machines to do meaningful work with energy harnessed from a continuous flux of heat and particles. As dictated by the Second Law of Thermodynamics and its fluctuation theorem corollaries, irreversibility in nonequilibrium processes can be quantified in terms of how much entropy such dynamics produce. In this work, we seek to address a fundamental question linking biology and nonequilibrium physics: can the evolved dissipative pathways that facilitate biomolecular function be identified by their extent of entropy production in general relaxation processes? We here synthesize massive molecular dynamics simulations, Markov state models (MSMs), and nonequilibrium statistical mechanical theory to probe dissipation in two key classes of signaling proteins: kinases and G-protein-coupled receptors (GPCRs). Applying machinery from large deviation theory, we use MSMs constructed from protein simulations to generate dynamics conforming to positive levels of entropy production. We note the emergence of an array of peaks in the dynamical response (transient analogs of phase transitions) that draw the proteins between distinct levels of dissipation, and we see that the binding of ATP and agonist molecules modifies the observed dissipative landscapes. Overall, we find that dissipation is tightly coupled to activation in these signaling systems: dominant entropy-producing trajectories become localized near important barriers along known biological activation pathways. We go on to classify an array of equilibrium and nonequilibrium molecular switches that harmonize to promote functional dynamics. PMID:26240354

  16. Consideration on rationalization of reactor safety systems. Reliability assessment of decay heat removal systems in commercialized sodium cooled FBR concepts

    For commercialization of FBR (Fast Breeder Reactor), the reactor safety systems are needed not only to have necessary and enough reliability but also to decrease the amount of materials in order that the FBR has economical competitiveness against LWR (Light Water Reactor) and another electrical power supply systems. In this study, reliability assessment, which calculates the occurrence frequencies of PLOHS (Protected Loss Of Heat Sink) sequences, was performed for three kinds of large size sodium cooled fast reactors with decreased number of loops and of support systems examined in the Feasibility Studies on Commercialized FBR System. The realistic evaluation was performed using the failure rate data of components based on the domestic LWR operating experience. The result is: All of the three kinds of commercialized FBR concepts are expected to achieve the frequencies of PLOHS sequences caused by internal events of the plant under 10-6/ry, assuming that common-mode-failure is excluded. In addition, the dominating cause of the coincidence of the incidents and the information that improves reliability of decay heat removal systems are summarized for each concept. In order to evaluate design margin, the reliability assessment was performed in the case that the capacity of natural circulation cooling was rein forced from 100%/3 x 3 loops to 50% x 3 loops or from 25% x 4 loops to 100%/3 x 4 loops easing to succeed decay heat removal. In that case, it is confirmed that the frequencies of PLOHS sequences decrease by about one order of magnitude. (author)

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

    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.

  18. Actinide, Activation Product and Fission Product Decay Data for Reactor-based Applications

    The UK Activation Product Decay Data Library was first released in September 1977 as UK-PADD1, to be followed by regular improvements on an almost yearly basis up to the assembly of UKPADD6.12 in March 2013. Similarly, the UK Heavy Element and Actinide Decay Data Library followed in December 1981 as UKHEDD1, with the implementation of various modifications leading to UKHEDD2.6, February 2008. Both the data content and evaluation procedures are defined, and the most recent evaluations are described in terms of specific radionuclides and the resulting consistency of their recommended decay-data files. New versions of the UKPADD and UKHEDD libraries are regularly submitted to the NEA Data Bank for possible inclusion in the JEFF library

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

    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. Implementación del modelo de decay heat ANSI/ ANS-5.1-2005 EN TRAC-BF1

    SOLER MARTÍNEZ, MARÍA DESAMPARADOS; Barrachina Celda, Teresa María; Miró Herrero, Rafael; Concejal, A.; Melara, J.; Verdú Martín, Gumersindo Jesús

    2012-01-01

    En este artículo se muestran los resultados de la implementación del modelo de calor residual (Decay Heat) ANSI/ANS-5.1 2005 en el código termohidráulico TRAC-BF1. Los modelos para el cálculo del calor residual en el código TRAC-BF1 se corresponden con el estándar ANS 1973, por defecto, y con el estándar ANS 1979, si es seleccionado por parte del usuario. Con la entrada en vigor del estándar ANS 1994 y, su posterior revisión, el estándar ANS 2005 los modelos que presenta TRAC-...

  1. Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

    Lahaye S.

    2016-01-01

    Full Text Available Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation’s choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011 is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.

  2. Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

    Lahaye, S.; Huynh, T. D.; Tsilanizara, A.

    2016-03-01

    Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation's choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011) is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.

  3. Analytical solution for laser evaporative heating process: time exponentially decaying pulse case

    The modelling of the laser heating process gives insight into the laser workpiece interaction and minimizes the experimental cost. In the present study, analytical solution for the laser pulse heating process is considered and the closed form solution for the temperature rise due to time exponentially varying pulse is obtained. In the analysis, evaporation of the surface is taken into account. A Laplace transformation method was used when formulating the closed form solution for the temperature profiles. The effect of pulse parameters on the temperature profiles is examined in detail. It is found that the closed form solution derived from the present study reduces to the previously obtained analytical solution when the surface recession velocity is set to zero in the closed form solution. Moreover, the predictions of numerical simulation and closed form solution are found to be in good agreement. (author)

  4. Basic CFD investigation of decay heat removal in a pool type research reactor

    Safety is one of the most important and desirable characteristic in a nuclear plant. Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid-dynamics (CFD) code is used to simulate the process of natural circulation in an open pool research reactor after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous media, where the parameters were obtained from a separately detailed CFD analysis. (author)

  5. Plasma heating in the very early and decay phases of solar flares

    Falewicz, R.; Siarkowski, M.; Rudawy, P.

    2011-01-01

    In this paper we analyze the energy budgets of two single-loop solar flares under the assumption that non-thermal electrons are the only source of plasma heating during all phases of both events. The flares were observed by the Ramaty High Energy Solar Spectroscopic Imager (RHESSI) and Geostationary Operational Environmental Satellite (GOES) on September 20, 2002 and March 17, 2002, respectively. For both investigated flares we derived the energy fluxes contained in non-thermal electron beams...

  6. 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

    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

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

    Lefrois, R.T.; Knowles, G.R.; Mathur, A.K.; Budimir, J.

    1979-02-01

    The report describes active heat exchange concepts for use with thermal energy storage systems in the temperature range of 250/sup 0/C to 350/sup 0/C, using the heat of fusion of molten salts for storing thermal energy. It identifies over 25 novel techniques for active heat exchange thermal energy storage systems. Salt mixtures that freeze and melt in appropriate ranges are identified and are evaluated for physico-chemical, economic, corrosive and safety characteristics. Eight active heat exchange concepts for heat transfer during solidification are conceived and conceptually designed for use with selected storage media. The concepts are analyzed for their scalability, maintenance, safety, technological development and costs. A model for estimating and scaling storage system costs is developed and is used for economic evaluation of salt mixtures and heat exchange concepts for a large scale application. The importance of comparing salts and heat exchange concepts on a total system cost basis, rather than the component cost basis alone, is pointed out. Comparison of these costs with current state-of-the-art systems should be avoided due to significant differences in developmental status. The heat exchange concepts were sized and compared for 6.5 MPa/281/sup 0/C steam conditions and a 1000 MW(t) heat rate for six hours. A cost sensitivity analysis for other design conditions is also carried out. The study resulted in the selection of a shell and coated-tube heat exchanger concept and a direct contact-reflux boiler heat exchange concept. For the storage medium, a dilute eutectic mixture of 99 wt % NaNO/sub 3/ and 1 wt % NaOH is selected for use in experimenting with the selected heat exchanger concepts in subsequent tasks.

  8. Heat dissipation guides activation in signaling proteins

    Weber, Jeffrey K.; Shukla, Diwakar; Pande, Vijay S.

    2015-01-01

    As with their macroscopic counterparts, the moving parts of nanoscale protein machines grow hot while in operation. A portion of the energy biomolecules harness to perform meaningful work is always dissipated as heat into the surroundings. Here, we feature a methodology by which dominant dissipative trajectories can be extracted from detailed models of protein dynamics. In two important classes of signaling proteins [kinases and G-protein–coupled receptors (GPCRs)], we find that the regions o...

  9. Periphytic photosynthetic stimulation of extracellular enzyme activity in aquatic microbial communities associated with decaying typha litter.

    Francoeur, Steven N; Schaecher, Mark; Neely, Robert K; Kuehn, Kevin A

    2006-11-01

    We examined the effect of light on extracellular enzyme activities of periphytic/endogenous microbial assemblages associated with decomposing litter of an emergent macrophyte Typha angustifolia within a small inland wetland in southeastern Michigan. Standing-dead Typha leaf litter was collected, placed into floating wire mesh litter baskets, and submerged in a wetland pool. Enzyme saturation assays were conducted on three occasions following litter submergence (days 9, 28, and 44) to generate saturation curves for the individual enzymes tested and to examine potential differences in enzyme saturation kinetics during microbial colonization and development. Experimental light manipulations were conducted on two occasions during microbial development (days 10 and 29). Short-term (30 min) light exposure significantly increased extracellular beta-glucosidase activity of litter-associated microbial communities. Activities of beta-xylosidase and leucine-aminopeptidase were not stimulated, and stimulation of phosphatase activity was variable. The exact mechanism for increased enzyme activity remains unknown, but it may have been increased pH arising from periphytic algal photosynthesis. These results suggest that extracellular enzyme activity in microbial communities colonizing natural organic substrata may be influenced by light/photosynthesis, as has previously been demonstrated for periphyton communities grown on artificial, inert substrata. Thus, light/photosynthetic mediated stimulation of extracellular enzyme activities may be a common occurrence in microbial communities associated with natural decaying plant litter in wetlands and might engender diurnal patterns in other microbial decay processes (e.g., production, organic matter decomposition, and mineralization). PMID:17082997

  10. 多样化非能动衰变热排出方法研究%Study on diverse passive decay heat removal approach

    林千; 司胜义

    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.%事故情况下的衰变热排出是涉及核安全的重要方面.采用非能动方法来排出衰变热对于提高核反应堆的安全性非常有益.在目前一些先进反应堆中通过设置非能动余热排出系统、非能动安注系统、非能动安全壳冷却系统等安全子系统,形成多样化的从堆芯到最终热阱的非能动衰变热排出渠道.论文对多种非能动衰变热排出方法和系统设计方案进行了归纳总结,比较分析了这些非能动衰变热排出方法的共性特征和区别,探讨了非能动衰变热排出系统的设计原理.通过对传热过程分解,将这些衰变热排出方法表达为一些基本传热形式的不同组合方式,根据不同的组合可获得多样化的非能动衰变热排出方法和新的系统设计方案.