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

International Nuclear Information System (INIS)

Epiney, A.S.

2010-01-01

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

Decay heat uncertainty quantification of MYRRHA

OpenAIRE

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

2017-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

Nicols, A.L.

2006-01-01

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

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

International Nuclear Information System (INIS)

Epiney, A. S.

2010-09-01

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

Uncertainties in fission-product decay-heat calculations

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

An application program for fission product decay heat calculations

International Nuclear Information System (INIS)

Pham, Ngoc Son; Katakura, Jun-ichi

2007-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Advances in technologies for decay heat removal

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Directory of Open Access Journals (Sweden)

Giacomino Bandini

2008-01-01

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

Status of the Japanese decay heat standard

International Nuclear Information System (INIS)

Katakura, Jun-ichi

1992-01-01

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

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

International Nuclear Information System (INIS)

Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

2000-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-10-01

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

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

International Nuclear Information System (INIS)

Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

1999-10-01

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

Contribution of short-lived nuclides to decay heat

International Nuclear Information System (INIS)

Katakura, Jun-ichi

1987-01-01

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

Consistency among integral measurements of aggregate decay heat power

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Application of least-squares method to decay heat evaluation

International Nuclear Information System (INIS)

Schmittroth, F.; Schenter, R.E.

1976-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

Summary report of RAMONA investigations into passive decay heat removal

International Nuclear Information System (INIS)

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

1995-07-01

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

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

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1986-12-01

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

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

International Nuclear Information System (INIS)

Oyamatsu, Kazuhiro; Mori, Hideki

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1986-06-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

A decay heat removal system requiring no external energy

International Nuclear Information System (INIS)

Costes, D.; Fermandjian, J.

1983-12-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Decay heat removal for the liquid metal fast breeder reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

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

1975-10-15

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

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

International Nuclear Information System (INIS)

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

1995-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Rebah, J

1998-08-01

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

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

International Nuclear Information System (INIS)

Nayak, A.K.

2015-01-01

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

Method for removal of decay heat of radioactive substances

International Nuclear Information System (INIS)

Hesky, H.; Wunderer, A.

1981-01-01

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

Sensitivity and uncertainty analysis for fission product decay heat calculations

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

Davies, B.S.J.

1978-07-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

International Nuclear Information System (INIS)

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

1996-03-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Radionuclide inventory and heat generation analysis in disposal of radioactive waste

International Nuclear Information System (INIS)

Suryanto

1997-01-01

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

A decay heat removal methodology for reuseable orbital transfer vehicles

Science.gov (United States)

McDaniel, Patrick J.; Perkins, David R.

1992-07-01

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

Passive decay heat removal from the core region

International Nuclear Information System (INIS)

Hichen, E.F.; Jaegers, H.

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-15

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

Tests for removal of decay heat by natural convection

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Generating cryptographic keys by radioactive decays

International Nuclear Information System (INIS)

Grupen, Claus; Maurer, Ingo; Schmidt, Dieter; Smolik, Ludek

2001-01-01

We are presenting a new method for the generation of statistically genuine random bitstream with very high frequency which can be employed for cryptographic purposes. The method uses the feature of statistically unpredictable radioactive decays as the source of randomness. The measured quantity is the time distance between the responses of a small ionisation chamber due to the recording of ionising decay products. This time measurement is converted into states representing 0o r 1. The data generated in our experiment successfully passed FIPS PUB 140-1 and die hard statistical tests. For the simulation of systematic effects Monte Carlo techniques were used

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

International Nuclear Information System (INIS)

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

1982-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

Directory of Open Access Journals (Sweden)

Porta A.

2016-01-01

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

A frame work for heat generation/absorption and modified homogeneous–heterogeneous reaction in flow based on non-Darcy–Forchheimer medium

Directory of Open Access Journals (Sweden)

Tasawar Hayat

2018-04-01

Full Text Available The present work aims to report the consequences of Darcy–Forchheimer medium in flow of Cross fluid model toward a stretched surface. Flow in porous space is categorized by Darcy–Forchheimer medium. Further heat transfer characteristics are examined via thermal radiation and heat generation/absorption. Transformation procedure is used. The arising system of nonlinear ordinary differential equations is solved numerically by means of shooting method. The effects of different flow variables on velocity, temperature, concentration, skin friction, and heat transfer rate are discussed. The obtained outcomes show that velocity was enhanced with the increase in the Weissenberg number but decays with increase in the porosity parameter and Hartman number. Temperature field is boosted by thermal radiation and heat generation; however, it decays with the increase in the Prandtl number. Keywords: Cross Fluid, Heat Generation/Absorption, Homogeneous–Heterogeneous Reactions, Non-Darcy–Forchheimer Medium, Thermal Radiation

Current status of decay heat measurements, evaluations, and needs

International Nuclear Information System (INIS)

Dickens, J.K.

1986-01-01

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

Analysis of decay heat removal by natural convection in PFBR

International Nuclear Information System (INIS)

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

1993-01-01

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

Passive Decay Heat Removal System for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Self-disposal option for heat-generating waste - 59182

International Nuclear Information System (INIS)

Ojovan, Michael I.; Poluektov, Pavel P.; Kascheev, Vladimir A.

2012-01-01

Self-descending heat generating capsules can be used for disposal of dangerous radioactive wastes in extremely deep layers of the Earth preventing any release of radionuclides into the biosphere. Self-disposal option for heat-generating radioactive waste such as spent fuel, high level reprocessing waste or spent sealed radioactive sources, known also as rock melting concept, was considered in the 70's as a viable alternative disposal option by both Department of Energy in the USA and Atomic Industry Ministry in the USSR. Self-disposal is currently reconsidered as a potential alternative route to existing options for solving the nuclear waste problem and is associated with the renaissance of nuclear industry. Self- disposal option utilises the heat generated by decaying radionuclides of radioactive waste inside a heavy and durable capsule to melt the rock on its way down. As the heat from radionuclides within the capsule partly melts the enclosing rock, the relatively low viscosity and density of the silicate melt allow the capsule to be displaced upwards past the heavier capsule as it sinks. Eventually the melt cools and solidifies (e.g. vitrifies or crystallizes), sealing the route along which the capsule passed. Descending or self-disposal continues until enough heat is generated by radionuclides to provide partial melting of surrounding rock. Estimates show that extreme depths of several tens and up to hundred km can be reached by capsules which could never be achieved by other techniques. Self- disposal does not require complex and expensive disposal facilities and provides a minimal footprint used only at operational stage. It has also an extremely high non- proliferation character and degree of safety. Utilisation of heat generated by relatively short-lived radionuclides diminishes the environmental uncertainties of self-disposal and increases the safety of this concept. Self-sinking heat-generating capsules could be launched from the bottom of the sea as

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

Nishi, Y.; Kinoshita, I.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Passive safety systems for decay heat removal of MRX

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Control of the ASTRA decay heat removal system

International Nuclear Information System (INIS)

Nedelik, A.

1982-11-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1982-06-01

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

PANDA passive decay heat removal transient test results

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Studies related to emergency decay heat removal in EBR-II

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Parametric decay instabilities in ECR heated plasmas

International Nuclear Information System (INIS)

Porkolab, M.

1982-01-01

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

Possible generation of heat from nuclear fusion in Earth's inner core.

Science.gov (United States)

Fukuhara, Mikio

2016-11-23

The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

Reduction of weighing errors caused by tritium decay heating

International Nuclear Information System (INIS)

Shaw, J.F.

1978-01-01

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

Formation and decay of laser-generated shock waves

Energy Technology Data Exchange (ETDEWEB)

Cottet, F.; Romain, J.P.

1982-01-01

The process of formation and decay of laser-generated shock waves is described by a hydrodynamic model. Measurements of shock velocities are performed on copper foils for incident intensities between 3 x 10/sup 11/ and 3 x 10/sup 12/ W/cm/sup 2/, with the use of piezoelectric detectors. Maximum induced pressures are found between 0.5 and 1.2 Mbar in the intensity range considered. Analysis of the results with the shock-evolution model outlines the importance of the decay process of laser-generated shocks.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-02-01

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

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

International Nuclear Information System (INIS)

1993-02-01

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

Microscopic beta and gamma data for decay-heat needs

International Nuclear Information System (INIS)

Dickens, J.K.

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1995-12-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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

International Nuclear Information System (INIS)

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

1986-08-01

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

Application of photoconductivity decay and photocurrent generation ...

Indian Academy of Sciences (India)

Unknown

... (PCD) and photocurrent generation (PCG) methods are simple and low cost methods of ... flection based detection of the change in photoconductivity .... 2.2 Photoconductivity decay behaviour .... con solar cell of known spectral response. 4.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Micro thrust and heat generator

Science.gov (United States)

Garcia, E.J.

1998-11-17

A micro thrust and heat generator have a means for providing a combustion fuel source to an ignition chamber of the micro thrust and heat generator. The fuel is ignited by a ignition means within the micro thrust and heat generator`s ignition chamber where it burns and creates a pressure. A nozzle formed from the combustion chamber extends outward from the combustion chamber and tappers down to a narrow diameter and then opens into a wider diameter where the nozzle then terminates outside of said combustion chamber. The pressure created within the combustion chamber accelerates as it leaves the chamber through the nozzle resulting in pressure and heat escaping from the nozzle to the atmosphere outside the micro thrust and heat generator. The micro thrust and heat generator can be microfabricated from a variety of materials, e.g., of polysilicon, on one wafer using surface micromachining batch fabrication techniques or high aspect ratio micromachining techniques (LIGA). 30 figs.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Vate, J.F. van de

1976-01-01

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

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

International Nuclear Information System (INIS)

Tobias, A.

1989-05-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

Rudstam, G.

1979-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Heat transfer enhancement for fin-tube heat exchanger using vortex generators

International Nuclear Information System (INIS)

Yoo, Seong Yeon; Park, Dong Seong; Chung, Min Ho; Lee, Sang Yun

2002-01-01

Vortex generators are fabricated on the fin surface of a fin-tube heat exchanger to augment the convective heat transfer. In addition to horseshoe vortices formed naturally around the tube of the fin-tube heat exchanger, longitudinal vortices are artificially created on the fin surface by vortex generators. The purpose of this study is to investigate the local heat transfer phenomena in the fin-tube heat exchangers with and without vortex generators, and to evaluate the effect of vortices on the heat transfer enhancement. Naphthalene sublimation technique is employed to measure local mass transfer coefficients, then analogy equation between heat and mass transfer is used to calculate heat transfer coefficients. Experiments are performed for the model of fin-circular tube heat exchangers with and without vortex generators, and of fin-flat tube heat exchangers with and without vortex generators. Average heat transfer coefficients of fin-flat tube heat exchanger without vortex generator are much lower than those of fin-circular tube heat exchanger. On the other hand, fin-flat tube heat exchanger with vortex generators has much higher heat transfer value than conventional fin-circular tube heat exchanger. At the same time, pressure losses for four types of heat exchanger is measured and compared

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

International Nuclear Information System (INIS)

Neely, G.J.; Schmittroth, F.

1982-08-01

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

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

International Nuclear Information System (INIS)

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

1995-12-01

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

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

Directory of Open Access Journals (Sweden)

Stankunas Gediminas

2017-01-01

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

Simultaneous power generation and heat recovery using a heat pipe assisted thermoelectric generator system

International Nuclear Information System (INIS)

Remeli, Muhammad Fairuz; Tan, Lippong; Date, Abhijit; Singh, Baljit; Akbarzadeh, Aliakbar

2015-01-01

Highlights: • A new passive power cogeneration system using industrial waste heat was introduced. • Heat pipes and thermoelectrics were used for recovering waste heat and electricity. • Theoretical model predicted the 2 kW test rig could recover 1.345 kW thermal power. • 10.39 W electrical power was produced equivalent to 0.77% conversion efficiency. - Abstract: This research explores a new method of recovering waste heat and electricity using a combination of heat pipes and thermoelectric generators (HP-TEG). The HP-TEG system consists of Bismuth Telluride (Bi 2 Te 3 ) based thermoelectric generators (TEGs), which are sandwiched between two finned heat pipes to achieve a temperature gradient across the TEG for thermoelectricity generation. A theoretical model was developed to predict the waste heat recovery and electricity conversion performances of the HP-TEG system under different parametric conditions. The modelling results show that the HP-TEG system has the capability of recovering 1.345 kW of waste heat and generating 10.39 W of electrical power using 8 installed TEGs. An experimental test bench for the HP-TEG system is under development and will be discussed in this paper

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

International Nuclear Information System (INIS)

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

2009-01-01

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

A revised ANS standard for decay heat from fission products

International Nuclear Information System (INIS)

Schrock, V.E.

1978-01-01

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

Conjugate transient natural convection in a cylindrical enclosure with internal volumetric heat generation

International Nuclear Information System (INIS)

Sharma, Anil Kumar; Velusamy, K.; Balaji, C.

2008-01-01

This paper reports the results of a numerical investigation of transient turbulent natural convection heat transfer from a volumetric energy generating source placed inside a cylindrical enclosure filled with low Prandtl number fluid (liquid sodium, Pr = 0.005). Two-dimensional conservation equations of mass, momentum and energy, coupled with the Boussinesq approximation, are solved using a finite volume based discretisation method employing the SIMPLE algorithm for the pressure velocity coupling. Turbulence is modeled using the k-ε model with physical boundary conditions. The study presents the transient features of confined turbulent natural convection, due to time varying generation of heat in the volumetric source. The intensity of heat source exponentially decays with time and the source is placed over circular plates with a central opening. Results obtained from the numerical model compare favorably with those reported in the literature for steady state natural convection. Numerical simulations are carried out to display the sequential evolution of flow and thermal fields and the maximum temperature reached in the source. The advantages of distributing the heat source on multi trays have been quantified

Next Generation Microchannel Heat Exchangers

CERN Document Server

Ohadi, Michael; Dessiatoun, Serguei; Cetegen, Edvin

2013-01-01

In Next Generation Microchannel Heat Exchangers, the authors’ focus on the new generation highly efficient heat exchangers and presentation of novel data and technical expertise not available in the open literature.  Next generation micro channels offer record high heat transfer coefficients with pressure drops much less than conventional micro channel heat exchangers. These inherent features promise fast penetration into many mew markets, including high heat flux cooling of electronics, waste heat recovery and energy efficiency enhancement applications, alternative energy systems, as well as applications in mass exchangers and chemical reactor systems. The combination of up to the minute research findings and technical know-how make this book very timely as the search for high performance heat and mass exchangers that can cut costs in materials consumption intensifies.

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

Energy Technology Data Exchange (ETDEWEB)

Bourganel, St

2002-11-01

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

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

Directory of Open Access Journals (Sweden)

مریم قربانی

2015-05-01

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

Assessment of NJOY generated neutron heating factors based on JEF/EFF-1

International Nuclear Information System (INIS)

Vontobel, P.

1990-01-01

Using the NJOY nuclear data processing system, a coupled neutron-photon multigroup MATXS-formatted nuclear data library was generated based on the files JEF/EFF-1. The neutron heating factors contained in this VITAMIN-J structured library are compared with those of MACLIB-IV. The main differences are due to the included decay heat of shortlived reaction products in MACKLIB-IV and/or due to too high/low photon production data of some JEF/EFF-1 isotopes. It is recommended to check carefully the energy balance of new evaluations containing photon production data. How this can be done with the help of the NJOY HEATR module is shown in an example. (author) 35 figs., 9 refs

Distributed heat generation in a district heating system

OpenAIRE

Lennermo, Gunnar; Lauenberg, Patrick

2016-01-01

District heating (OH) systems need to be improved  regarding integration  of decentralised  heat generation. Micro production, prosumers and smart grids are terms becoming more and more common  in  connection  to  the  power  grid.  Concerning district  heating,  the  development  is slower, although improving. Today there are a number of such decentralised units for heat generation,  mainly  solar,  that have been partly evaluated.  Previous  studies  have shown  that there is a need to deve...

Ocean disposal of heat generating waste

International Nuclear Information System (INIS)

1985-06-01

A number of options for the disposal of vitrified heat generating waste are being studied to ensure that safe methods are available when the time comes for disposal operations to commence. This study has considered the engineering and operational aspects of the Penetrator Option for ocean disposal to enable technical comparisons with other options to be made. In the Penetrator Option concept, waste would be loaded into carefully designed containers which would be launched at a suitable deep ocean site where they would fall freely through the water and would embed themselves completely within the seabed sediments. Radiological protection would be provided by a multi-barrier system including the vitrified waste form, the penetrator containment, the covering sediment and the ocean. Calculations and demonstration have shown that penetrators could easily achieve embedment depths in excess of 30m and preliminary radiological assessments indicate that 30m of intact sediment would be an effective barrier for radionuclide isolation. The study concludes that a 75mm thickness of low carbon steel appears to be sufficient to provide a containment life of 500 to 1000 years during which time the waste heat output would have decayed to an insignificant level. Disposal costs have been assessed. (author)

Heat and electricity generating methods

International Nuclear Information System (INIS)

Buter, J.

1977-01-01

A short synopsis on the actual methods of heating of lodgings and of industrial heat generation is given. Electricity can be generated in steam cycles heated by burning of fossil fuels or by nuclear energy. A valuable contribution to the electricity economy is produced in the hydroelectric power plants. Besides these classical methods, also the different procedures of direct electricity generation are treated: thermoelectric, thermionic, magnetohydrodynamic power sources, solar and fuel cells. (orig.) [de

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

International Nuclear Information System (INIS)

Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-01-01

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

Heat transfer enhancement in cross-flow heat exchanger using vortex generator

International Nuclear Information System (INIS)

Yoo, S. Y.; Kwon, H. K.; Kim, B. C.; Park, D. S.; Lee, S. S.

2003-01-01

Fouling is very serious problem in heat exchanger because it rapidly deteriorates the performance of heat exchanger. Cross-flow heat exchanger with vortex generators is developed, which enhance heat transfer and reduce fouling. In the present heat exchanger, shell and baffle are removed from the conventional shell-and-tube heat exchanger. The naphthalene sublimation technique is employed to measure the local heat transfer coefficients. The experiments are performed for single circular tube, staggered array tube bank and in-line array tube bank with and without vortex generators. Local and average Nusselt numbers of single tube and tube bank with vortex generator are investigated and compared to those of without vortex generator

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

International Nuclear Information System (INIS)

Livadiotis, George

2007-01-01

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

Heat generated by knee prostheses.

Science.gov (United States)

Pritchett, James W

2006-01-01

Temperature sensors were placed in 50 knees in 25 patients who had one or both joints replaced. Temperature recordings were made before walking, after walking, and after cycling. The heat generated in healthy, arthritic, and replaced knees was measured. The knee replacements were done using eight different prostheses. A rotating hinge knee prosthesis generated a temperature increase of 7 degrees C in 20 minutes and 9 degrees C in 40 minutes. An unconstrained ceramic femoral prosthesis articulating with a polyethylene tibial prosthesis generated a temperature increase of 4 degrees C compared with a healthy resting knee. The other designs using a cobalt-chrome alloy and high-density polyethylene had temperature increases of 5 degrees-7 degrees C with exercise. Frictional heat generated in a prosthetic knee is not immediately dissipated and may result in wear, creep, and other degenerative processes in the high-density polyethylene. Extended periods of elevated temperature in joints may inhibit cell growth and perhaps contribute to adverse performance via bone resorption or component loosening. Prosthetic knees generate more heat with activity than healthy or arthritic knees. More-constrained knee prostheses generate more heat than less-constrained prostheses. A knee with a ceramic femoral component generates less heat than a knee with the same design using a cobalt-chromium alloy.

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

International Nuclear Information System (INIS)

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

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

Analysis of decay heat removal following loss of RHR

International Nuclear Information System (INIS)

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

1991-01-01

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

Filtered thermal neutron captured cross sections measurements and decay heat calculations

International Nuclear Information System (INIS)

Pham Ngoc Son; Vuong Huu Tan

2015-01-01

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

Performance of ALMR passive decay heat removal system

International Nuclear Information System (INIS)

Boardman, C.E.; Hunsbedt, A.

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Bourganel, St

2002-11-01

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

Heat generated by dental implant drills during osteotomy-a review: heat generated by dental implant drills.

Science.gov (United States)

Mishra, Sunil Kumar; Chowdhary, Ramesh

2014-06-01

Osseointegration is the more stable situation and results in a high success rate of dental implants. Heat generation during rotary cutting is one of the important factors influencing the development of osseointegration. To assess the various factors related to implant drills responsible for heat generation during osteotomy. To identify suitable literature, an electronic search was performed using Medline and Pubmed database. Articles published in between 1960 to February 2013 were searched. The search is focused on heat generated by dental implant drills during osteotomy. Various factors related to implant drill such effect of number of blades; drill design, drill fatigue, drill speed and force applied during osteotomies which were responsible for heat generation were reviewed. Titles and abstracts were screened, and literature that fulfilled the inclusion criteria was selected for a full-text reading. The initial literature search resulted in 299 articles out of which only 70 articles fulfils the inclusion criteria and were included in this systematic review. Many factors related to implant drill responsible for heat generation were found. Successful preparation of an implant cavity with minimal damage to the surrounding bone depends on the avoidance of excessive temperature generation during surgical drilling. The relationship between heat generated and implant drilling osteotomy is multifactorial in nature and its complexity has not been fully studied. Lack of scientific knowledge regarding this issue still exists. Further studies should be conducted to determine the various factors which generate less heat while osteotomy such as ideal ratio of force and speed in vivo, exact time to replace a drill, ideal drill design, irrigation system, drill-bone contact area.

Removal of radon decay products with ion generators - comparison of experimental results with theory

International Nuclear Information System (INIS)

Maher, E.F.; Rudnick, S.N.; Moeller, D.W.

1985-01-01

The potential of ion generators to remove radon decay products from the airspace of residences or mines was investigated both experimentally and theoretically. A positive ion generator, producing an air ion current of less than 2 μA and operated in a 78 m 3 chamber with air exchange rates ranging from 0.2 to 0.8-hr -1 and relative humidities ranging from 20 to nearly 100%, reduced the concentrations of airborne radon decay products by as much as 85%. A negative ion generator, operated under the same range of conditions, was less effective, producing airborne radon decay product removals up to 67%. Experimental results compared favorably with a simple theoretical model that hypothesizes a three-part process: 1) radon decay products, as well as aerosol particles to which some of these decay products attach, are charged by diffusion of the air ions produced by the generator; 2) the air ions also produce a nonuniform space charge in the chamber that results in an electric field gradient radially outwards from the generator to the chamber surfaces; and 3) because of the influence of this electric field, the charged decay products and particles migrate toward nearby surfaces where they plate out. The net benefit of unipolar space charging is a substantial decrease in the steady-state radon decay product concentrations in the airspace with a corresponding reduction in the lung dose equivalent to the occupants

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

International Nuclear Information System (INIS)

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

1994-09-01

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

Natural convection in heat-generating fluids

International Nuclear Information System (INIS)

Bol'shov, Leonid A; Kondratenko, Petr S; Strizhov, Valerii F

2001-01-01

Experimental and theoretical studies of convective heat transfer from a heat-generating fluid confined to a closed volume are reviewed. Theoretical results are inferred from analytical estimates based on the relevant conservation laws and the current understanding of the convective heat-transfer processes. Four basic and one asymptotic regime of heat transfer are identified depending on the heat generation rate. Limiting heat-transfer distribution patterns are found for the lower boundary. Heat transfer in a quasi-two-dimensional geometry is analyzed. Quasi-steady-state heat transfer from a cooling-down fluid without internal heat sources is studied separately. Experimental results and theoretical predictions are compared. (reviews of topical problems)

Fourth-generation effects in Z0 to bb-bar decay

International Nuclear Information System (INIS)

Ma Wengan; Xu Zizong

1991-01-01

Possible fourth generation effects in partial widths in Z 0 into bb-bar decay are discussed with the consideration of all quark mixing within the minimal standard model. We show that it is possible to disentangle the effects from the fourth generation. (author)

A Modified Entropy Generation Number for Heat Exchangers

Institute of Scientific and Technical Information of China (English)

无

1996-01-01

This paper demonstrates the difference between the entropy generation number method proposed by Bejian and the method of entropy generation per unit amount of heat transferred in analyzing the ther-modynamic performance of heat exchangers,points out the reason for leading to the above difference.A modified entropy generation number for evaluating the irreversibility of heat exchangers is proposed which is in consistent with the entropy generation per unit amount of heat transferred in entropy generation analysis.The entropy generated by friction is also investigated.Results show that when the entropy generated by friction in heat exchangers in taken into account,there is a minimum total entropy generation number while the NTU and the ratio of heat capacity rates vary.The existence of this minimum is the prerequisite of heat exchanger optimization.

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

International Nuclear Information System (INIS)

Son, Pham Ngoc; Tan, Vuong Huu

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1976-03-01

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

A review of U-235 decay heat measurements and calculations

International Nuclear Information System (INIS)

Walker, W.H.

1979-08-01

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

Application of optimal estimation techniques to FFTF decay heat removal analysis

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

1977-09-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Sterbentz, J.W.

1997-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-03-01

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

Heat flow and heat generation in greenstone belts

Science.gov (United States)

Drury, M. J.

1986-01-01

Heat flow has been measured in Precambrian shields in both greenstone belts and crystalline terrains. Values are generally low, reflecting the great age and tectonic stability of the shields; they range typically between 30 and 50 mW/sq m, although extreme values of 18 and 79 mW/sq m have been reported. For large areas of the Earth's surface that are assumed to have been subjected to a common thermotectonic event, plots of heat flow against heat generation appear to be linear, although there may be considerable scatter in the data. The relationship is expressed as: Q = Q sub o + D A sub o in which Q is the observed heat flow, A sub o is the measured heat generation at the surface, Q sub o is the reduced heat flow from the lower crust and mantle, and D, which has the dimension of length, represents a scale depth for the distribution of radiogenic elements. Most authors have not used data from greenstone belts in attempting to define the relationship within shields, considering them unrepresentative and preferring to use data from relatively homogeneous crystalline rocks. A discussion follows.

General-purpose heat source: Research and development program, radioisotope thermoelectric generator/thin fragment impact test

International Nuclear Information System (INIS)

Reimus, M.A.H.; Hinckley, J.E.

1996-11-01

The general-purpose heat source provides power for space missions by transmitting the heat of 238 Pu decay to an array of thermoelectric elements in a radioisotope thermoelectric generator (RTG). Because the potential for a launch abort or return from orbit exists for any space mission, the heat source response to credible accident scenarios is being evaluated. This test was designed to provide information on the response of a loaded RTG to impact by a fragment similar to the type of fragment produced by breakup of the spacecraft propulsion module system. The results of this test indicated that impact by a thin aluminum fragment traveling at 306 m/s may result in significant damage to the converter housing, failure of one fueled clad, and release of a small quantity of fuel

Glas generator for the steam gasification of coal with nuclear generated heat

International Nuclear Information System (INIS)

Buchner, G.

1980-01-01

The use of heat from a High Temperature Reactor (HTR) for the steam gasification of coal saves coal, which otherwise is burnt to generate the necessary reaction heat. The gas generator for this process, a horizontal pressure vessel, contains a fluidized bed of coal and steam. An ''immersion-heater'' type of heat exchanger introduces the nuclear generated heat to the process. Some special design problems of this gasifier are presented. Reference is made to the present state of development of the steam gasification process with heat from high temperature reactors. (author)

Reliability assessment on decay heat removal system of a fast reactor

International Nuclear Information System (INIS)

Hioki, Kazumasa

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Event generators for {{\\eta }}/{{{\\eta }}}^{^{\\prime} } decays at BESIII

Science.gov (United States)

Qin, Nian; Zhang, Zhen-Yu; Fang, Shuang-Shi; Zhou, Xiang; Du, Lin-Lin; Qiao, Hao-Xue

2018-01-01

The light unflavoured meson {{η }}/{{{η }}}^{\\prime } decays are valuable for testing non-perturbative quantum chromodynamics and exploring new physics beyond the Standard Model. This paper describes a series of event generators, including {{η }}/{{{η }}}^{\\prime }\\to {{γ }}{{{l}}}+{{{l}}}-, {{η }}/{{{η }}}^{\\prime }\\to {{γ }}{{{π }}}+{{{π }}}-, {{{η }}}^{\\prime }\\to {{ω }}{{{e}}}+{{{e}}}-, {{η }}\\to {{{π }}}+{{{π }}}-{{{π }}}0, {{η }}/{{{η }}}^{\\prime }\\to {{{π }}}0{{{π }}}0{{{π }}}0, {{{η }}}^{\\prime }\\to {{η }}{{π }}{{π }} and {{{η }}}^{\\prime }\\to {{{π }}}+{{{π }}}-{{{π }}}+{{{π }}}-/{{{π }}}+{{{π }}}-{{{π }}}0{{{π }}}0, which have been developed for investigating {{η }}/{{{η }}}^{\\prime } decay dynamics. For most of these generators, their usability has been validated in BESIII analyses for determining the detection efficiency, and background studies. The consistency between data and Monte Carlo shows that these generators work well in the BESIII simulation, and will also be useful for ongoing BESIII analyses and other experiments for studying {{η }}/{{{η }}}^{\\prime } physics. Supported by National Natural Science Foundation of China (NSFC) (11205117, 11575133, 11675184), the China Scholarship Council (201506275156), the Wuhan University PhD Short-time Mobility Program 2016, the Joint Funds of the NSFC and Henan Province (U1504112)

Kinetic study on non-thermal volumetric plasma decay in the early afterglow of air discharge generated by a short pulse microwave or laser

Energy Technology Data Exchange (ETDEWEB)

Yang, Wei, E-mail: yangwei861212@126.com; Zhou, Qianhong; Dong, Zhiwei [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China)

2016-08-28

This paper reports a kinetic study on non-thermal plasma decay in the early afterglow of air discharge generated by short pulse microwave or laser. A global self-consistent model is based on the particle balance of complex plasma chemistry, electron energy equation, and gas thermal balance equation. Electron-ion Coulomb collision is included in the steady state Boltzmann equation solver to accurately describe the electron mobility and other transport coefficients. The model is used to simulate the afterglow of microsecond to nanosecond pulse microwave discharge in N{sub 2}, O{sub 2}, and air, as well as femtosecond laser filament discharge in dry and humid air. The simulated results for electron density decay are in quantitative agreement with the available measured ones. The evolution of plasma decay under an external electric field is also investigated, and the effect of gas heating is considered. The underlying mechanism of plasma density decay is unveiled through the above kinetic modeling.

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

International Nuclear Information System (INIS)

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

1993-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

An Operators View of Reliability Testing and Decay Heat Rejection Systems

International Nuclear Information System (INIS)

Henderson, J.D.C.

1975-01-01

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

Estimation of heat transfer and heat source in a molten pool

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

Estimation of heat transfer and heat source in a molten pool

International Nuclear Information System (INIS)

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

2001-01-01

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

Spectrum-generating SU(4) in particle physics. II. Electromagnetic decays of vector mesons

International Nuclear Information System (INIS)

Bohm, A.; Teese, R.B.

1977-09-01

The decay rates for the electromagnetic decays of vector mesons are derived within the spectrum-generating SU(4) approach. Radiative as well as leptonic decays of vector mesons can be derived from one theoretical assumption and given in terms of three reduced matrix elements. The implication of the experimental value GAMMA(rho → πγ) = (35 +- 10) keV for the form of the electromagnetic current operator is discussed

Heat Transfer and Entropy Generation Analysis of an Intermediate Heat Exchanger in ADS

Science.gov (United States)

Wang, Yongwei; Huai, Xiulan

2018-04-01

The intermediate heat exchanger for enhancement heat transfer is the important equipment in the usage of nuclear energy. In the present work, heat transfer and entropy generation of an intermediate heat exchanger (IHX) in the accelerator driven subcritical system (ADS) are investigated experimentally. The variation of entropy generation number with performance parameters of the IHX is analyzed, and effects of inlet conditions of the IHX on entropy generation number and heat transfer are discussed. Compared with the results at two working conditions of the constant mass flow rates of liquid lead-bismuth eutectic (LBE) and helium gas, the total pumping power all tends to reduce with the decreasing entropy generation number, but the variations of the effectiveness, number of transfer units and thermal capacity rate ratio are inconsistent, and need to analyze respectively. With the increasing inlet mass flow rate or LBE inlet temperature, the entropy generation number increases and the heat transfer is enhanced, while the opposite trend occurs with the increasing helium gas inlet temperature. The further study is necessary for obtaining the optimized operation parameters of the IHX to minimize entropy generation and enhance heat transfer.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Heat generation during plunge stage in friction stir welding

Directory of Open Access Journals (Sweden)

Veljić Darko M.

2013-01-01

Full Text Available This paper deals with the heat generation in the Al alloy Al2024-T3 plate under different rotating speeds and plunge speeds during the plunge stage of friction stir welding (FSW. A three-dimensional finite element model (FEM is developed in the commercial code ABAQUS/Explicit using the arbitrary Lagrangian-Eulerian formulation, the Johnson-Cook material law and Coulomb’s Law of friction. The heat generation in FSW can be divided into two parts: frictional heat generated by the tool and heat generated by material deformation near the pin and the tool shoulder region. Numerical results obtained in this work indicate a more prominent influence from the friction-generated heat. The slip rate of the tool relative to the workpiece material is related to this portion of heat. The material velocity, on the other hand, is related to the heat generated by plastic deformation. Increasing the plunging speed of the tool decreases the friction-generated heat and increases the amount of deformation-generated heat, while increasing the tool rotating speed has the opposite influence on both heat portions. Numerical results are compared with the experimental ones, in order to validate the numerical model, and a good agreement is obtained.

Study of entropy generation in a slab with non-uniform internal heat generation

Directory of Open Access Journals (Sweden)

El Haj Assad Mamdouh

2013-01-01

Full Text Available Analysis of entropy generation in a rectangular slab with a nonuniform internal heat generation is presented. Dimensionless local and total entropy generation during steady state heat conduction through the slab are obtained. Two different boundary conditions have been considered in the analysis, the first with asymmetric convection and the second with constant slab surface temperature. Temperature distribution within the slab is obtained analytically. The study investigates the effect of some relevant dimensionless heat transfer parameters on entropy generation. The results show that there exists a minimum local entropy generation but there does not exist a minimum total entropy generation for certain combinations of the heat transfer parameters. The results of calculations are presented graphically.

Susy Les Houches accord: Interfacing SUSY spectrum calculators, decay packages, and event generators

International Nuclear Information System (INIS)

Skands, P.; Allanach, B.C.; Baer, H.

2003-11-01

An accord specifying generic file structures for 1) supersymmetric model specifications and input parameters, 2) electroweak scale supersymmetric mass and coupling spectra, and 3) decay tables is defined, to provide a universal interface between spectrum calculation programs, decay packages, and high energy physics event generators. (orig.)

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

International Nuclear Information System (INIS)

Winterhagen, D.

1986-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Heat transfer enhancement in heat exchangers by longitudinal vortex generators

International Nuclear Information System (INIS)

Guntermann, T.; Fiebig, M.; Mitra, N.K.

1990-01-01

In this paper heat transfer enhancement and flow losses are computed for the interaction of a laminar channel flow with a pair of counterrotating longitudinal vortices generated by a pair of delta-winglets punched out of the channel wall. The geometry simulates an element of a fin-plate or fin-tube heat exchanger. The structure of the vortex flow and temperature distribution, the local heat transfer coefficients and the local flow losses are discussed for a sample case. For a Reynolds number of Re d = 1000 and a vortex generator angle of attack of β = 25 degrees heat transfer is enhanced locally by more than 300% and in the mean by 50%. These values increase further with Re and β

Analysis of Heat Generation Mechanism in Ultrasound Infrared Thermography

International Nuclear Information System (INIS)

Choi, Man Yong; Lee, Seung Seok; Park, Jeong Hak; Kang, Ki Soo; Kim, Won Tae

2009-01-01

Heat generation mechanism of ultrasound infrared thermography is still not well understood, yet and there are two reliable assumptions of heat generation, friction and thermo-mechanical effect. This paper investigates the principal cause of heat generation at fatigue crack with experimental and numerical approach. Our results show most of heat generation is contributed by friction between crack interface and thermo-mechanical effect is a negligible quantity

Generator-absorber-heat exchange heat transfer apparatus and method and use thereof in a heat pump

Science.gov (United States)

Phillips, B.A.; Zawacki, T.S.

1998-07-21

Numerous embodiments and related methods for generator-absorber heat exchange (GAX) are disclosed, particularly for absorption heat pump systems. Such embodiments and related methods use, as the heat transfer medium, the working fluid of the absorption system taken from the generator at a location where the working fluid has a rich liquor concentration. 5 figs.

Simple model for decay of laser generated shock waves

International Nuclear Information System (INIS)

Trainor, R.J.

1980-01-01

A simple model is derived to calculate the hydrodynamic decay of laser-generated shock waves. Comparison with detailed hydrocode simulations shows good agreement between calculated time evolution of shock pressure, position, and instantaneous pressure profile. Reliability of the model decreases in regions of the target where superthermal-electron preheat effects become comparable to shock effects

Heat transfer of liquid-metal magnetohydrodynamic flow with internal heat generation

International Nuclear Information System (INIS)

Kumamaru, Hiroshige; Kurita, Kazuhisa; Kodama, Satoshi

2000-01-01

Numerical calculations on heat transfer of a magnetohydrodynamic (MHD) flow with internal heat generation in a rectangular channel have been performed for the cases of very-large Hartmann numbers, finite wall conductivities and small aspect ratio (i.e. small length ratios of the channel side perpendicular to the applied magnetic field and the side parallel to the field), simulating typical conditions for a fusion-reactor blanket. The Nusselt numbers of the MHD flow in rectangular channels with aspect ratios of 1/10 to 1/40 for Hartmann numbers of ∼5 x 10 5 become ∼10 times higher than those for the corresponding flow under no magnetic field. The Nusselt number becomes higher as the internal heat generation rate increases as far as the heat generation rates in a fusion reactor blanket are considered. (author)

Possible design of PBR for passive decay heat removal

International Nuclear Information System (INIS)

Sambuu, Odmaa; Obara, Toru

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Passive flow heat exchanger simulation for power generation from solar pond using thermoelectric generators

Science.gov (United States)

Baharin, Nuraida'Aadilia; Arzami, Amir Afiq; Singh, Baljit; Remeli, Muhammad Fairuz; Tan, Lippong; Oberoi, Amandeep

2017-04-01

In this study, a thermoelectric generator heat exchanger system was designed and simulated for electricity generation from solar pond. A thermoelectric generator heat exchanger was studied by using Computational Fluid Dynamics to simulate flow and heat transfer. A thermoelectric generator heat exchanger designed for passive in-pond flow used in solar pond for electrical power generation. A simple analysis simulation was developed to obtain the amount of electricity generated at different conditions for hot temperatures of a solar pond at different flow rates. Results indicated that the system is capable of producing electricity. This study and design provides an alternative way to generate electricity from solar pond in tropical countries like Malaysia for possible renewable energy applications.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

Schenter, G.K.; Schmittroth, F.

1979-01-01

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

Flavor-changing Z decay in the one generation technicolor models

International Nuclear Information System (INIS)

Wang, X.; Lu, G.; Xiao, Z.

1995-01-01

The flavor-changing decay Z→b bar s(bar bs) induced through pseudo Goldstone bosons (PGB's) is calculated in two kinds of one generation technicolor models (OGTM's). (a) For model I, we find that an interesting branching ratio B(Z→b bar s+bar bs)∼10 -6 can be obtained for particular choices of the parameters; such a magnitude of the order of the branching ratio is at the border of being detectable. (b) For model II, the PGB contributions can strongly enhance the branching ratio B(Z→b bar s+bar bs). With the current experimental limit on the branching ratios of rare Z decay, the model-dependent bounds on the mass of the color octet pseudo Goldstone bosons can be derived. As will be seen, the decay Z→b bar s(bar bs) may provide a unique window to study the TC theory

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

International Nuclear Information System (INIS)

Loimann, E.; Reutler, H.

1977-01-01

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

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

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2014-01-01

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

The ratio between the decay heat output and activity content of discharged magnox fuel

International Nuclear Information System (INIS)

Davies, B.S.J.

1977-01-01

Values of the ratio between activity and heat production rate have been calculated for magnox fuel irradiated to 3500 and 8000 MWd.Te -1 and for cooling times of 100, 200 and 500 days. Results are expressed in terms of both MeV.decay -1 and MCi.KW -1 . The results indicate that: for these irradiation and cooling conditions 21 nuclides account for over 99% of the total activity; the calculated values show only small variations with burn-up and cooling time, although the mean energy per decay does fall slightly at 500 days cooling: so for many purposes a median value of 0.63 MeV.decay -1 (0.27 MCi.MW -1 ) may be used; the calculated values have standard deviations ranging from 2.6% at 100 days cooling to 9% at 500 days cooling. (author)

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

International Nuclear Information System (INIS)

Roughley, R.; Jones, N.

1975-01-01

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

Safety provision during heating of coal downcast shafts with gas heat generators using degassed methane

Directory of Open Access Journals (Sweden)

В. Р. Алабьев

2017-06-01

Together with heat generators of mixed type the article also describes a working principle of heat generator of indirect action type, which to the fullest extent possible meets requirements of Russian Federation legislation and regulation for application of this heat generators in coal mines conditions. The article has a principal working scheme of heat unit layout using this type of generator. It is shown that after development of corresponding normative documents regulating processes of design, construction and operation of heating units using heaters of indirect action, their application in Russian coal mines will be possible without breaking Safety standards and rules.

A value/impact assessment for alternative decay heat removal systems

International Nuclear Information System (INIS)

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

1984-01-01

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

4th Generation District Heating (4GDH)

DEFF Research Database (Denmark)

Lund, Henrik; Werner, Sven; Wiltshire, Robin

2014-01-01

This paper defines the concept of 4th Generation District Heating (4GDH) including the relations to District Cooling and the concepts of smart energy and smart thermal grids. The motive is to identify the future challenges of reaching a future renewable non-fossil heat supply as part...... of the implementation of overall sustainable energy systems. The basic assumption is that district heating and cooling has an important role to play in future sustainable energy systems – including 100 percent renewable energy systems – but the present generation of district heating and cooling technologies will have...

The optimization of longitudinal convective fins with internal heat generation

International Nuclear Information System (INIS)

Razelos, P.

1979-01-01

The solution of the optimization problem for longitudinal convective fins of constant thickness, triangular or parabolic profile, and uniform internal heat generation, is presented. The cases considered are those of a given heat generation density, total heat generation and heat generation per unit width of the fin, when either the heat dissipation or the width of the fin is prescribed. The results are set forth in a nondimensional form, which are presented graphically. The effect of the fin's thermal conductivity upon the optimum dimensions is discussed, and limiting values for the heat generation and the heat dissipation, which may be imposed on the fin for a feasible optimization, are also obtained. (Auth.)

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

Directory of Open Access Journals (Sweden)

M. Ali Asgarian

2018-04-01

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

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

Science.gov (United States)

Ali Asgarian, M.; Abbasi, M.

2018-04-01

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

Thermal mechanical analysis of applications with internal heat generation

Science.gov (United States)

Govindarajan, Srisharan Garg

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

Heat operated cryogenic electrical generator

International Nuclear Information System (INIS)

Fletcher, J.C.; Wang, T.C.; Saffren, M.M.; Elleman, D.D.

1975-01-01

An electrical generator useful for providing electrical power in deep space, is disclosed. The subject electrical generator utilizes the unusual hydrodynamic property exhibited by liquid helium as it is converted to and from a superfluid state to cause opposite directions of rotary motion for a rotor cell thereof. The physical motion of said rotor cell is employed to move a magnetic field provided by a charged superconductive coil mounted on the exterior of said cell. An electrical conductor is placed in surrounding proximity to said cell to interact with the moving magnetic field provided by the superconductive coil and thereby generate electrical energy. A heat control arrangement is provided for the purpose of causing the liquid helium to be partially converted to and from a superfluid state by being cooled and heated, respectively. (U.S.)

Entropy and heat generation of lithium cells/batteries

International Nuclear Information System (INIS)

Wang Songrui

2016-01-01

The methods and techniques commonly used in investigating the change of entropy and heat generation in Li cells/batteries are introduced, as are the measurements, calculations and purposes. The changes of entropy and heat generation are concomitant with the use of Li cells/batteries. In order to improve the management and the application of Li cells/batteries, especially for large scale power batteries, the quantitative investigations of the change of entropy and heat generating are necessary. (topical review)

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

Science.gov (United States)

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

2009-01-01

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

District heating and co-generation in Slovenia

International Nuclear Information System (INIS)

Hrovatin, Franc; Pecaric, Marko; Perovic, Olgica

2000-01-01

Recent development of district heating systems, gasification and co-generation processes in local communities in Slovenia as well as current status, potentials, possibilities and plans for further development in this sphere are presented. The current status presents energy production, distribution and use in district heating systems and in local gas distribution networks. An analysis of the energy and power generated and distributed in district power systems, made with regard to the size of the system, fuel used, type of consumers and the way of production, is given. Growth in different areas of local power systems in the period of last years is included. Potentials in the sphere of electrical energy and heat co-generation were assessed. Some possibilities and experience in heat energy storage are given and trends and plans for further development are introduced. (Authors)

Heat Generation by Irradiated Complex Composite Nanostructures

DEFF Research Database (Denmark)

Ma, Haiyan; Tian, Pengfei; Pello, Josselin

2014-01-01

Heating of irradiated metallic e-beam generated nanostructures was quantified through direct measurements paralleled by novel model-based numerical calculations. By comparing discs, triangles, and stars we showed how particle shape and composition determines the heating. Importantly, our results...... revealed that substantial heat is generated in the titanium adhesive layer between gold and glass. Even when the Ti layer is as thin as 2 nm it absorbs as much as a 30 nm Au layer and hence should not be ignored....

Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators

Energy Technology Data Exchange (ETDEWEB)

Yanagihara, J.I.; Rodriques, R. Jr. [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering

1996-12-31

Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)

Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators

Energy Technology Data Exchange (ETDEWEB)

Yanagihara, J I; Rodriques, R Jr [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering

1997-12-31

Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)

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

Science.gov (United States)

Li, Shengjie

1997-09-01

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

Nuclear heat generating plants - technical concepts and market potentials. Chapter 8

International Nuclear Information System (INIS)

Thoene, E.

1988-01-01

To determine the advantages and disadvantages of different heat generating systems, a comparison is made between nuclear heat generating plants and competing heat generating systems. Nuclear heat generating plant concepts in practice have to compete with a wide range of existing and new fossil heat generating technologies of the most different capacities, ranging from combined heat and power generation to individual heating in one-family houses. Heat generation costs are calculated by means of a dynamic annuity method from an economic point of view. The development of real prices of fossil energy sources is based on two scenarios characterized as follows: scenario I - insignificant price increase by the year 2000, then stagnant; scenario II - moderate price increase by the year 2010, then stagnant. As a result of that systems comparison it can be stated that the considered nuclear heat generating plants may be an interesting competitive heat generation option, provided the assumptions on which the study is based can be implemented. This applies especially to investment costs. At the same time those plants contribute to a diversification of energy source options on the heat market. Their use leads to a reduction of fossil fuel imports, increasing at the same time short- and long-term supply guarantees. If nuclear heat generating plants substitute fossil heat generating plants, or render the construction of new ones superfluous, they contribute to avoiding chemical air pollutants. (orig./UA) [de

Control system for fluid heated steam generator

Science.gov (United States)

Boland, J.F.; Koenig, J.F.

1984-05-29

A control system for controlling the location of the nucleate-boiling region in a fluid heated steam generator comprises means for measuring the temperature gradient (change in temperature per unit length) of the heating fluid along the steam generator; means for determining a control variable in accordance with a predetermined function of temperature gradients and for generating a control signal in response thereto; and means for adjusting the feedwater flow rate in accordance with the control signal.

Update and evaluation of decay data for spent nuclear fuel analyses

Science.gov (United States)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

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

Update and evaluation of decay data for spent nuclear fuel analyses

Directory of Open Access Journals (Sweden)

Simeonov Teodosi

2017-01-01

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

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

International Nuclear Information System (INIS)

Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-02-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Heat-Pipe-Associated Localized Thermoelectric Power Generation System

Science.gov (United States)

Kim, Pan-Jo; Rhi, Seok-Ho; Lee, Kye-Bock; Hwang, Hyun-Chang; Lee, Ji-Su; Jang, Ju-Chan; Lee, Wook-Hyun; Lee, Ki-Woo

2014-06-01

The present study focused on how to improve the maximum power output of a thermoelectric generator (TEG) system and move heat to any suitable space using a TEG associated with a loop thermosyphon (loop-type heat pipe). An experimental study was carried out to investigate the power output, the temperature difference of the thermoelectric module (TEM), and the heat transfer performance associated with the characteristic of the researched heat pipe. Currently, internal combustion engines lose more than 35% of their fuel energy as recyclable heat in the exhaust gas, but it is not easy to recycle waste heat using TEGs because of the limited space in vehicles. There are various advantages to use of TEGs over other power sources, such as the absence of moving parts, a long lifetime, and a compact system configuration. The present study presents a novel TEG concept to transfer heat from the heat source to the sink. This technology can transfer waste heat to any location. This simple and novel design for a TEG can be applied to future hybrid cars. The present TEG system with a heat pipe can transfer heat and generate power of around 1.8 V with T TEM = 58°C. The heat transfer performance of a loop-type heat pipe with various working fluids was investigated, with water at high heat flux (90 W) and 0.05% TiO2 nanofluid at low heat flux (30 W to 70 W) showing the best performance in terms of power generation. The heat pipe can transfer the heat to any location where the TEM is installed.

RESEARCH OF HYDRODYNAMICS OF HEAT GENERATORS FOR MECHANICAL SYSTEMS AUTONOMOUS HEATING

Directory of Open Access Journals (Sweden)

E. M. Derbasova

2014-01-01

Full Text Available A design of mechanical heat source, allows direct conversion of mechanical energy of the wind flow into thermal energy due to friction forces in a highly viscous fluid. Obtained theoretical dependence for calculating the heat generated by converting mechanical energy into heat. For laminar flow of a highly viscous, fluid in the gap between the stationary and rotating disk heat source. Based on experimental studies to determine the average thickness of the boundary layer between the rotating and fixed disks. The dependences to identify key structural dimensions of mechanical heat sources for heating systems. 

Heat Pipe-Assisted Thermoelectric Power Generation Technology for Waste Heat Recovery

Science.gov (United States)

Jang, Ju-Chan; Chi, Ri-Guang; Rhi, Seok-Ho; Lee, Kye-Bock; Hwang, Hyun-Chang; Lee, Ji-Su; Lee, Wook-Hyun

2015-06-01

Currently, large amounts of thermal energy dissipated from automobiles are emitted through hot exhaust pipes. This has resulted in the need for a new efficient recycling method to recover energy from waste hot exhaust gas. The present experimental study investigated how to improve the power output of a thermoelectric generator (TEG) system assisted by a wickless loop heat pipe (loop thermosyphon) under the limited space of the exhaust gas pipeline. The present study shows a novel loop-type heat pipe-assisted TEG concept to be applied to hybrid vehicles. The operating temperature of a TEG's hot side surface should be as high as possible to maximize the Seebeck effect. The present study shows a novel TEG concept of transferring heat from the source to the sink. This technology can transfer waste heat to any local place with a loop-type heat pipe. The present TEG system with a heat pipe can transfer heat and generate an electromotive force power of around 1.3 V in the case of 170°C hot exhaust gas. Two thermoelectric modules (TEMs) for a conductive block model and four Bi2Te3 TEMs with a heat pipe-assisted model were installed in the condenser section. Heat flows to the condenser section from the evaporator section connected to the exhaust pipe. This novel TEG system with a heat pipe can be placed in any location on an automobile.

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

International Nuclear Information System (INIS)

1975-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-07-01

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

Optimization of the Heat Exchangers of a Thermoelectric Generation System

Science.gov (United States)

Martínez, A.; Vián, J. G.; Astrain, D.; Rodríguez, A.; Berrio, I.

2010-09-01

The thermal resistances of the heat exchangers have a strong influence on the electric power produced by a thermoelectric generator. In this work, the heat exchangers of a thermoelectric generator have been optimized in order to maximize the electric power generated. This thermoelectric generator harnesses heat from the exhaust gas of a domestic gas boiler. Statistical design of experiments was used to assess the influence of five factors on both the electric power generated and the pressure drop in the chimney: height of the generator, number of modules per meter of generator height, length of the fins of the hot-side heat exchanger (HSHE), length of the gap between fins of the HSHE, and base thickness of the HSHE. The electric power has been calculated using a computational model, whereas Fluent computational fluid dynamics (CFD) has been used to obtain the thermal resistances of the heat exchangers and the pressure drop. Finally, the thermoelectric generator has been optimized, taking into account the restrictions on the pressure drop.

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

International Nuclear Information System (INIS)

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

1989-04-01

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

Theoretical investigation of the doubly stratified flow of an Eyring-Powell nanomaterial via heat generation/absorption

Science.gov (United States)

Khan, M. Ijaz; Waqas, M.; Alsaedi, A.; Hayat, T.; Khan, M. Imran

2017-11-01

The mixed convective flow of an Eyring-Powell nanomaterial in a doubly stratified medium is addressed in this paper. The stretching surface has varying thickness. The nanofluid model given by Buongiorno is utilized in the formulation of energy and concentration expressions. Heat generation is also retained. Ordinary differential systems are obtained by utilizing the transformations procedure. Homotopy series solutions containing exponentially functions are developed. Significant characteristics of influential variables for velocity, temperature, nanoparticle concentration, skin friction coefficient and Nusselt and Sherwood numbers are reported through graphs and tables. It is found that stratification phenomenon leads to a decay in temperature and nanoparticle concentration.

Searches for rare top processes and decay and third generation SUSY

CERN Document Server

Zambito, Stefano; The ATLAS collaboration

2018-01-01

Many models of new physics, including natural models of Supersymmetry and vector-like quarks, predict processes that would be visible at the LHC in events with third generation quarks. These include searches for stop and sbottom particles in a variety of decay modes, as well as for flavour-changing neutral current top-quark decays. The modeling of the Standard Model backgrounds to these processes is difficult and vital to the success of the searches. This talk presents recent results from ATLAS and CMS on searches for rare top processes and stop and sbottom pair production, using the data collected during the LHC Run 2.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Sorption heat engines: simple inanimate negative entropy generators

OpenAIRE

Muller, Anthonie W. J.; Schulze-Makuch, Dirk

2005-01-01

The name 'sorption heat engines' is proposed for simple negative entropy generators that are driven by thermal cycling and work on alternating adsorption and desorption. These generators are in general not explicitly recognized as heat engines. Their mechanism is applicable to the fields of engineering, physics, chemistry, geology, and biology, in particular the origin of life. Four kinds of sorption heat engines are distinguished depending on the occurrence of changes in the adsorbent or ads...

Thermal resistance of a convectively cooled plate with applied heat flux and variable internal heat generation

International Nuclear Information System (INIS)

Venkataraman, N.S.; Cardoso, H.P.; Oliveira Filho, O.B. de

1981-01-01

The conductive heat transfer in a rectangular plate with nonuniform internal heat generation, with one end convectively cooled and a part of the opposite end subjected to external heat flux is considered. The remaining part of this end as well as the other two sides are thermally insulated. The governing differential equation is solved by a finite difference scheme. The variation of the thermal resistance with Biot modulus, the plate geometry, the internal heat generation parameter and the type of profile of internal heat generation is discussed. (author) [pt

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

CERN Document Server

Cepraga, D G

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1976-09-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Composite electric generator equipped with steam generator for heating reactor coolant

International Nuclear Information System (INIS)

Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

1997-01-01

The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

Thermodynamic analysis of waste heat power generation system

International Nuclear Information System (INIS)

Guo, Jiangfeng; Xu, Mingtian; Cheng, Lin

2010-01-01

In the present work, a waste heat power generation system is analyzed based on the criteria with and without considering the heat/exergy loss to the environment. For the criteria without considering the heat/exergy loss to the environment, the first- and second-law efficiencies display different tendencies with the variations of some system parameters. When the heat/exergy loss to the environment is taken into consideration, the first and second law efficiencies display the same tendency. Thus, choosing the appropriate expressions for the performance criteria is crucial for the optimization design of the waste heat power generation system. It is found that there are two approaches to improving the system performance: one is to improve the heat/exergy input; the other is to enhance the heat-work conversion ability of the system. The former would deteriorate the environment if the heat-work conversion ability of the system remains unchanged; the latter could reduce the environmental impact but it's restricted by the heat/exergy input. Therefore, the optimal operation condition should be achieved at the trade-off between the heat/exergy input and the heat-work conversion ability of the system.

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

International Nuclear Information System (INIS)

Apperson, C.E. Jr.

1981-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-01-15

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

Parametric Decay during HHFW on NSTX

International Nuclear Information System (INIS)

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

2005-01-01

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

Magnetoresistance effect of heat generation in a single-molecular spin-valve

International Nuclear Information System (INIS)

Jiang, Feng; Yan, Yonghong; Wang, Shikuan; Yan, Yijing

2016-01-01

Based on non-equilibrium Green's functions' theory and small polaron transformation's technology, we study the heat generation by current through a single-molecular spin-valve. Numerical results indicate that the variation of spin polarization degree can change heat generation effectively, the spin-valve effect happens not only in electrical current but also in heat generation when Coulomb repulsion in quantum dot is smaller than phonon frequency and interestingly, when Coulomb repulsion is larger than phonon frequency, the inverse spin-valve effect appears by sweeping gate voltage and is enlarged with bias increasing. The inverse spin-valve effect will induce the unique heat magnetoresistance effect, which can be modulated from heat-resistance to heat-gain by gate voltage easily. - Highlights: • Spin-valve effect of heat generation happens when Coulomb repulsion in quantum dot is less than phonon frequency. • When Coulomb repulsion is larger than phonon frequency, inverse spin-valve effect appears and is enlarged with bias increasing. • The variation of spin polarization degree can change heat generation effectively. • The heat magnetoresistance can be modulated from heat-resistance to heat-gain by gate voltage easily.

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

Directory of Open Access Journals (Sweden)

Maryam Ghorbani

2014-11-01

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

Heat Recovery From Tail Gas Incineration To Generate Power

Energy Technology Data Exchange (ETDEWEB)

Tawfik, Tarek

2010-09-15

Many industrial processes result in tail gas wastes that must be flared or incinerated to abide with environmental guidelines. Tail gas incineration occurs in several chemical processes resulting in high-temperature exhaust gas that simply go to the stack, thus wasting all that valuable heat! This paper discusses useful heat recovery and electric power generation utilizing available heat in exhaust gas from tail gas incinerators. This heat will be recovered in a waste-heat recovery boiler that will produce superheated steam to expand in a steam turbine to generate power. A detailed cost estimate is presented.

Passive decay heat removal by natural air convection after severe accidents

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Heating tokamaks by parametric decay of intense extraordinary mode radiation

International Nuclear Information System (INIS)

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

1979-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Effective thermal conductivity of a heat generating rod bundle dissipating heat by natural convection and radiation

International Nuclear Information System (INIS)

Senve, Vinay; Narasimham, G.S.V.L.

2011-01-01

Highlights: → Transport processes in isothermal hexagonal sheath with 19 heat generating rods is studied. → Correlation is given to predict the maximum temperature considering all transport processes. → Effective thermal conductivity of rod bundle can be obtained using max temperature. → Data on the critical Rayleigh numbers for p/d ratios of 1.1-2.0 is presented. → Radiative heat transfer contributes to heat dissipation of 38-65% of total heat. - Abstract: A numerical study of conjugate natural convection and surface radiation in a horizontal hexagonal sheath housing 19 solid heat generating rods with cladding and argon as the fill gas, is performed. The natural convection in the sheath is driven by the volumetric heat generation in the solid rods. The problem is solved using the FLUENT CFD code. A correlation is obtained to predict the maximum temperature in the rod bundle for different pitch-to-diameter ratios and heat generating rates. The effective thermal conductivity is related to the heat generation rate, maximum temperature and the sheath temperature. Results are presented for the dimensionless maximum temperature, Rayleigh number and the contribution of radiation with changing emissivity, total wattage and the pitch-to-diameter ratio. In the simulation of a larger system that contains a rod bundle, the effective thermal conductivity facilitates simplified modelling of the rod bundle by treating it as a solid of effective thermal conductivity. The parametric studies revealed that the contribution of radiation can be 38-65% of the total heat generation, for the parameter ranges chosen. Data for critical Rayleigh number above which natural convection comes into effect is also presented.

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

Energy Technology Data Exchange (ETDEWEB)

Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

1998-03-01

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

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

KAUST Repository

Huang, Bin-Juine

2015-09-05

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

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

International Nuclear Information System (INIS)

Kim, Kyung Mo; Bang, In Cheol

2017-01-01

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

Hyperon decays and spectrum generating SU(3)

International Nuclear Information System (INIS)

Teese, R.B.; Boehm, A.

1976-02-01

The research program described in this review is aimed at describing the properties of relativistic one-hadron systems by an algebra of observables, in analogy to the nonrelativistic description of atoms. This formalism has recently been applied to the leptonic and semi-leptonic decays of pseudoscalar mesons, and was shown to be capable of predicting both the suppression of strangeness changing decays and the value of the form factor ratio xi in K/sub l 3 / decay. A preliminary description of the leptonic decays of hyperons indicates that second class matrix elements are predicted as a consequence of a precise formulation of SU(3) symmetry breaking. A chi 2 -fit to the experimental data indicates that this preliminary model is an improvement over the usual Cabibbo model, and points the way for further theoretical work. It is hoped that this program will lead to a model for the leptonic decays of hadrons which improves upon the results of the Cabibbo model and which explains some of the assumptions of that model

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

1994-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

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

OpenAIRE

Stankunas Gediminas; Tidikas Andrius

2017-01-01

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

Cogeneration using a nuclear reactor to generate process heat

International Nuclear Information System (INIS)

Alonso, Gustavo; Ramirez, Ramon

2009-01-01

Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

Sustainability assessment of renewable power and heat generation technologies

International Nuclear Information System (INIS)

Dombi, Mihály; Kuti, István; Balogh, Péter

2014-01-01

Rationalisation of consumption, more efficient energy usage and a new energy structure are needed to be achieved in order to shift the structure of energy system towards sustainability. The required energy system is among others characterised by intensive utilisation of renewable energy sources (RES). RES technologies have their own advantages and disadvantages. Nevertheless, for the strategic planning there is a great demand for the comparison of RES technologies. Furthermore, there are additional functions of RES utilisation expected beyond climate change mitigation, e.g. increment of employment, economic growth and rural development. The aim of the study was to reveal the most beneficial RES technologies with special respect to sustainability. Ten technologies of power generation and seven technologies of heat supply were examined in a multi-criteria sustainability assessment frame of seven attributes which were evaluated based on a choice experiment (CE) survey. According to experts the most important characteristics of RES utilisation technologies are land demand and social impacts i.e. increase in employment and local income generation. Concentrated solar power (CSP), hydropower and geothermal power plants are favourable technologies for power generation, while geothermal district heating, pellet-based non-grid heating and solar thermal heating can offer significant advantages in case of heat supply. - highlights: • We used choice experiment to estimate the weights of criteria for the sustainability assessment of RES technologies. • The most important attributes of RES technologies according to experts are land demand and social impacts. • Concentrated solar power (CSP), hydropower and geothermal power plants are advantageous technologies for power generation. • Geothermal district heating, pellet-based non-grid heating and solar thermal heating are favourable in case of heat supply

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

International Nuclear Information System (INIS)

Rasmussen, D.A.

1981-01-01

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

EVENT GENERATION OF STANDARD MODEL HIGGS DECAY TO DIMUON PAIRS USING PYTHIA SOFTWARE

CERN Document Server

Yusof, Adib

2015-01-01

My project for CERN Summer Student Programme 2015 is on Event Generation of Standard Model Higgs Decay to Dimuon Pairs using Pythia Software. Briefly, Pythia or specifically, Pythia 8.1 is a program for the generation of high-energy Physics events that is able to describe the collisions at any given energies between elementary particles such as Electron, Positron, Proton as well as anti-Proton. It contains theory and models for a number of Physics aspects, including hard and soft interactions, parton distributions, initial-state and final-state parton showers, multiparton interactions, fragmentation and decay. All programming code is to be written in C++ language for this version (the previous version uses FORTRAN) and can be linked to ROOT software for displaying output in form of histogram. For my project, I need to generate events for standard model Higgs Boson into Muon and anti-Muon pairs (H→μ+ μ) to study the expected significance value for this particular process at centre-of-mass energy of 13 TeV...

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

JNDC FP decay data file

International Nuclear Information System (INIS)

Yamamoto, Tohru; Akiyama, Masatsugu

1981-02-01

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

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

Directory of Open Access Journals (Sweden)

Frisoni Manuela

2016-01-01

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

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

International Nuclear Information System (INIS)

Rudstam, G.

1979-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Transient thermal stresses in circular cylinder under intermittently sudden heat generation

International Nuclear Information System (INIS)

Sugano, Y.; Saito, K.; Takeuti, Y.

1975-01-01

The thermal stresses associated with the transient temperature distribution arising in a circular cylinder under intermittently changing sudden heat generation over a finite band and with heat loss to a surrounding medium on the remainder of the cylinder surface are exactly analysed. For the first time the temperature field in a circular cylinder under sudden heat generation over a finite band of the cylinder surface is determined by combined use of Fourier cosine, Laplace transforms in axial position and time, respectively. Secondly it is assumed that the temperature fields in a circular cylinder subjected to heat generation Qsub(i) (i=0, 1, 2, ...) independently over a finite band are given by T 0 (r,z,t), T 1 (r,z,t), T 2 (r,z,t),... respectively. Tsub(i)(r,z,t) indicates the temperature field before the i-th heat generation Qsub(i). The thermal stresses associated with the temperature field described above are analysed by using the Hoyle stress functions. Numerical calculations are carried out for the extensive case of the ratio of the heat-generating length to the diameter of cylinder. It is found that the time in which the maximum stresses occur on the cylinder surface does not depend on the heat-generating length-to-diameter ratio

Rare decays of the Z and the standard model, 4th generation, and beyond

International Nuclear Information System (INIS)

Weiler, T.J.

1989-01-01

Several issues in rare decays of the Z are addressed. The rate for flavor-changing Z decay grows as the fourth power of the fermion masses internal to the quantum loop, and so offers a window to the existence of ultraheavy (m > M W ) fermions. In the standard model, with three generations, BR(Z → bs) -7 and BR(Z→tc) -13 . With four generations, BR(Z → bb 4 ) may be as large as 10 -5 if m b4 Z ; and similarly for BR(Z → N 4 v), where N 4 is the possibly heavy fourth generation neutrino. In supersymmetric and other two Higgs doublet models, BR(Z → tc) may be as large as 5 x 10 -6 in the three generation scheme. With minimal supersymmetry, the reaction Z → Hγ is guaranteed to go, with a parameter-dependent branching ratio of 10 -6 ± 3 . With mirror fermions or exotic E 6 fermions, the branching ratios for Z → ct (70 GeV), Z → μτ, and Z → bb 4 (70 GeV) are typically 10 -4 , 10 -4 , and 10 -3 respectively, clearly measurable at LEP. Depending on unknown quark masses, the Z may mix with vector (b 4 bar b 4 ) and the W may mix with vector (t bar b) or (t bar s). CP violating asymmetries in flavor-changing Z decay are immeasurably small in the standard model, but may be large in supersymmetric and other nonstandard models. 28 refs

Next Generation Nuclear Plant Intermediate Heat Exchanger Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C to 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium cooled, prismatic or pebble-bed reactor, and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Intermediate Heat Exchanger (IHX).This component will be operated in flowing, impure helium on the primary and secondary side at temperatures up to 950°C. There are major high temperature design, materials availability, and fabrication issues that need to be addressed. The prospective materials are Alloys 617, 230, 800H and X, with Alloy 617 being the leading candidate for the use at 950°C. The material delivery schedule for these materials does not pose a problem for a 2018 start up as the vendors can quote reasonable delivery times at the moment. The product forms and amount needed must be finalized as soon as possible. An

Heat Generation Effects on U-Mo/Al through ABAQUS FEM Simulation

Energy Technology Data Exchange (ETDEWEB)

Cho, Taewon; Jeong, Gwan Yoon; Lee, Cheol Min; Sohn Dongseong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

U-Mo/Al dispersion fuels have been considered a most promising candidate for a replacement of Highly Enriched Uranium (HEU) fuel in many research reactors. Coulson developed a FEM model which show the fuel meat realistically and compared the thermal conductivity results of two and three dimensional model. Williams also developed a FEM model which are different from the former in that it use regularly meshed unit cells. He showed a heat generation effects through FEM simulation and the effective thermal conductivity of the fuel with heat generated in the fuel particles is a little lower than that of the fuel with no heat generated. In the current work, the heat generation effects are analyzed and discussed in a wider range of volume fraction with more realistic models by using ABAQUS finite element package. The FEM model is used to determine the effective thermal conductivity of U-Mo/Al and to simulate the heat generation effects in the study. This model reflected the microscopic morphology of the fuel very well by making random distribution particles although the particle shape is considered as sphere. All simulation results show the heat generation effects although the effects are small when the volume fraction of fuels are high. When the particles are surrounded with interaction layers, the heat transfer from the particle to matrix is disturbed by interaction layers due to the low thermal conductivity of interaction layers. However this effects decreases when the sum of the volume fraction of fuels and interaction layers exceeds 40-50 vol% because a great portion of the heat must pass through fuels and interaction layers although the heat is applied on the surface. Therefore particle size and initial particle volume fractions will be the important factors for the heat generation effects when interaction layers grow during irradiations.

Experimental evaluation of sodium to air heat exchanger performance

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Science.gov (United States)

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

2017-04-01

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

Rare B decays, rare τ decays, and grand unification

International Nuclear Information System (INIS)

Sher, M.; Yuan, Y.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Current generation by minority-species heating

International Nuclear Information System (INIS)

Fisch, N.J.

1981-01-01

It is proposed that electric currents be generated from the preferential heating of ions travelling in one direction but with no net momentum injected into the system. This can be accomplished with, for example, travelling waves in a two-ion-species plasma. The current can be generated efficiently enough for the scheme to be of interest in maintaining steady-state toroidal currents in a reactor. (author)

Current generation by minority species heating

International Nuclear Information System (INIS)

Fisch, N.J.

1980-07-01

It is proposed that electric currents be generated from the preferential heating of ions traveling in one direction but with no net momentum injected into the system. This can be accomplished with, for example, traveling waves in a two-ion-species plasma. The current can be generated efficiently enough for the scheme to be of interest in maintaining steady-state toroidal currents in a reactor

Smart energy systems and 4th generation district heating

DEFF Research Database (Denmark)

Lund, Henrik; Duic, Neven; Østergaard, Poul Alberg

2016-01-01

scientific understanding on how we can design and implement a suitable and least-cost transformation into a sustainable energy future. The concept of Smart Energy Systems emphasizes the importance of being coherent and cross-sectoral when the best solutions are to be found and how this also calls......This editorial gives an introduction to the important relationship between Smart Energy Systems and 4th Generation District Heating and presents a number of selected papers from the 1st International Conference on the topic. All of the papers elaborate on or otherwise contribute to the theoretical...... for the active inclusion of the heating and cooling sectors. The concept of 4th Generation District Heating emphasizes that district heating and cooling are both important elements but also technologies that have to be developed further into a 4th generation version to be able to fulfil their roles in future...

Power generation from residual industrial heat

International Nuclear Information System (INIS)

Nguyen, T.Q.; Slawnwhite, J.D.; Boulama, K.Goni

2010-01-01

Industrial plants continuously reject large amounts of thermal energy through warm liquid or gaseous effluents during normal operation. These energy losses contribute to an inflation of production costs and also threaten the environment. This paper investigates methods of recovering the residual low grade thermal energy and converting it into higher quality mechanical energy using the thermodynamic Rankine cycle principle. For the temperature range of the available thermal energy, water was shown to be a poor working fluid for the conversion system, thus several potential working fluids, including ammonia, synthetic refrigerants, and organic compounds have been considered as alternatives. A comparative analysis led to the identification of different performance evaluation criteria. For example, the water-based Rankine cycle and, to a lesser extent, the ammonia-based Rankine cycle proved to be interesting when the power generation potential per unit working fluid mass flow rate was considered. On the other hand, Rankine-like cycles using dry hydrocarbon working fluids proved much more interesting in terms of energy conversion efficiency, as well as in terms of the net mechanical power generation potential for a given heat source. All performance indicators were low at low temperatures, and improved as the primary heat source was available at higher temperatures. This paper also discusses the influence of various external and internal operating parameters, such as heat source and heat sink temperatures, turbine and pump isentropic efficiencies and the addition of an internal heat exchanger on the overall performance of the energy recovery and conversion system.

Passive decay heat removal by sump cooling after core meltdown

International Nuclear Information System (INIS)

Knebel, J.U.; Mueller, U.

1996-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Ocean disposal of heat generating radioactive waste

International Nuclear Information System (INIS)

1986-03-01

The objective of this study was to predict tensile stress levels in thin-walled titanium alloy and thick-walled carbon steel containers designed for the ocean disposal of heat-generating radioactive wastes. Results showed that tensile stresses would be produced in both designs by the expansion of the lead filter, for a temperature rise of 200 0 C. Tensile stress could be reduced if the waste heat output at disposal was reduced. Initial stresses for the titanium-alloy containers could be relieved by heat treatment. (UK)

Development of low grade waste heat thermoelectric power generator

Directory of Open Access Journals (Sweden)

Suvit Punnachaiya

2010-07-01

Full Text Available This research aimed to develop a 50 watt thermoelectric power generator using low grade waste heat as a heat source,in order to recover and utilize the excess heat in cooling systems of industrial processes and high activity radioisotope sources. Electricity generation was based on the reverse operation of a thermoelectric cooling (TEC device. The TEC devices weremodified and assembled into a set of thermal cell modules operating at a temperature less than 100°C. The developed powergenerator consisted of 4 modules, each generating 15 watts. Two cascade modules were connected in parallel. Each modulecomprised of 96 TEC devices, which were connected in series. The hot side of each module was mounted on an aluminumheat transfer pipe with dimensions 12.212.250 cm. Heat sinks were installed on the cold side with cooling fans to provideforced air cooling.To test electricity generation in the experiment, water steam was used as a heat source instead of low grade waste heat.The open-circuit direct current (DC of 250 V and the short-circuit current of 1.2 A was achieved with the following operatingconditions: a hot side temperature of 96°C and a temperature difference between the hot and cold sides of 25°C. The DC poweroutput was inverted to an AC power source of 220 V with 50 Hz frequency, which can continuously supply more than 50 wattsof power to a resistive load as long as the heat source was applied to the system. The system achieved an electrical conversionefficiency of about 0.47 percent with the capital cost of 70 US$/W.

National need for utilizing nuclear energy for process heat generation

International Nuclear Information System (INIS)

Gambill, W.R.; Kasten, P.R.

1984-01-01

Nuclear reactors are potential sources for generating process heat, and their applications for such use economically competitive. They help satisfy national needs by helping conserve and extend oil and natural gas resources, thus reducing energy imports and easing future international energy concerns. Several reactor types can be utilized for generating nuclear process heat; those considered here are light water reactors (LWRs), heavy water reactors (HWRs), gas-cooled reactors (GCRs), and liquid metal reactors (LMRs). LWRs and HWRs can generate process heat up to 280 0 C, LMRs up to 540 0 C, and GCRs up to 950 0 C. Based on the studies considered here, the estimated process heat markets and the associated energy markets which would be supplied by the various reactor types are summarized

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Investigation of Counter-Flow in a Heat Pipe-Thermoelectric Generator (HPTEG)

Science.gov (United States)

Remeli, Muhammad Fairuz; Singh, Baljit; Affandi, Nor Dalila Nor; Ding, Lai Chet; Date, Abhijit; Akbarzadeh, Aliakbar

2017-05-01

This study explores a method of generating electricity while recovering waste heat through the integration of heat pipes and thermoelectric generators (i.e. HPTEG system). The simultaneous waste heat recovery and power generation processes are achieved without the use of any moving parts. The HPTEG system consists of bismuth telluride thermoelectric generators (TEG), which are sandwiched between two finned pipes to achieve a temperature gradient across the TEG for electricity generation. A counter-flow heat exchanger was built using two separate air ducts. The air ducts were thermally coupled using the HPTEG modules. The evaporator section of the heat pipe absorbed the waste heat in a hot air duct. The heat was then transferred across the TEG surfaces. The condenser section of the HPTEG collected the excess heat from the TEG cold side before releasing it to the cold air duct. A 2-kW electrical heater was installed in the hot air duct to simulate the exhaust gas. An air blower was installed at the inlet of each duct to direct the flow of air into the ducts. A theoretical model was developed for predicting the performance of the HPTEG system using the effectiveness-number of transfer units method. The developed model was able to predict the thermal and electrical output of the HPTEG, along with the rate of heat transfer. The results showed that by increasing the cold air velocity, the effectiveness of the heat exchanger was able to be increased from approximately 52% to 58%. As a consequence of the improved heat transfer, maximum power output of 4.3 W was obtained.

Passive heat removal system with injector-condenser

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Generation of heat on fuel rod in cosine pattern by using induction heating

International Nuclear Information System (INIS)

Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

2017-01-01

Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

Thermoelectric as recovery and harvesting of waste heat from portable generator

Science.gov (United States)

Mustafa, S. N.; Kamarrudin, N. S.; Hashim, M. S. M.; Bakar, S. A.; Razlan, Z. M.; Harun, A.; Ibrahim, I.; Faizi, M. K.; Saad, M. A. M.; Zunaidi, I.; Wan, W. K.; Desa, H.

2017-10-01

Generation of waste heat was ineluctable especially during energy producing process. Waste heat falls into low temperature grade make it complicated to utilize. Thermoelectric generator (TEG) offers opportunity to harvest any temperature grade heat into useful electricity. This project is covered about recovery and utilizing waste heat from portable electric generator by using a TEG which placed at exhaust surface. Temperature difference at both surfaces of TEG was enhanced with supplying cold air from a wind blower. It is found that, even at low air speed, the TEG was successfully produced electricity with aid from DC-DC booster. Results shows possibility to harvest low temperature grade heat and still exist areas for continual improvement.

Rare decays of the Z and the standard model, 4th generation, and beyond

Energy Technology Data Exchange (ETDEWEB)

Weiler, T.J.

1989-01-01

Several issues in rare decays of the Z are addressed. The rate for flavor-changing Z decay grows as the fourth power of the fermion masses internal to the quantum loop, and so offers a window to the existence of ultraheavy (m > M{sub W}) fermions. In the standard model, with three generations, BR(Z {yields} bs) < 10{sup -7} and BR(Z{yields}tc)<10{sup -13}. With four generations, BR(Z {yields} bb{sub 4}) may be as large as 10{sup -5} if m{sub b4} < M{sub Z}; and similarly for BR(Z {yields} N{sub 4}v), where N{sub 4} is the possibly heavy fourth generation neutrino. In supersymmetric and other two Higgs doublet models, BR(Z {yields} tc) may be as large as 5 {times} 10{sup -6} in the three generation scheme. With minimal supersymmetry, the reaction Z {yields} H{gamma} is guaranteed to go, with a parameter-dependent branching ratio of 10{sup -6 {plus minus} 3}. With mirror fermions or exotic E{sub 6} fermions, the branching ratios for Z {yields} ct (70 GeV), Z {yields} {mu}{tau}, and Z {yields} bb{sub 4} (70 GeV) are typically 10{sup -4}, 10{sup -4}, and 10{sup -3} respectively, clearly measurable at LEP. Depending on unknown quark masses, the Z may mix with vector (b{sub 4}{bar b}{sub 4}) and the W may mix with vector (t{bar b}) or (t{bar s}). CP violating asymmetries in flavor-changing Z decay are immeasurably small in the standard model, but may be large in supersymmetric and other nonstandard models. 28 refs.

Flow visualization in heat-generating porous media

International Nuclear Information System (INIS)

Lee, D.O.; Nilson, R.H.

1977-11-01

The work reported is in support of the Sandia Post-Accident Heat Removal Program, in which simulated LMFBR beds will be subjected to in-pile heating in the ACPR (Annular Core Pulsed Reactor). Flow visualization experiments were performed to gain some insight into the flow patterns and temperature distributions in a fluid-saturated heat-generating porous medium. Although much of the information presented is of a qualitative nature, it is useful in the recognition of the controlling transport process and in the formulation of analytic and numerical models

Power generation using sugar cane bagasse: A heat recovery analysis

Science.gov (United States)

Seguro, Jean Vittorio

The sugar industry is facing the need to improve its performance by increasing efficiency and developing profitable by-products. An important possibility is the production of electrical power for sale. Co-generation has been practiced in the sugar industry for a long time in a very inefficient way with the main purpose of getting rid of the bagasse. The goal of this research was to develop a software tool that could be used to improve the way that bagasse is used to generate power. Special focus was given to the heat recovery components of the co-generation plant (economizer, air pre-heater and bagasse dryer) to determine if one, or a combination, of them led to a more efficient co-generation cycle. An extensive review of the state of the art of power generation in the sugar industry was conducted and is summarized in this dissertation. Based on this models were developed. After testing the models and comparing the results with the data collected from the literature, a software application that integrated all these models was developed to simulate the complete co-generation plant. Seven different cycles, three different pressures, and sixty-eight distributions of the flue gas through the heat recovery components can be simulated. The software includes an economic analysis tool that can help the designer determine the economic feasibility of different options. Results from running the simulation are presented that demonstrate its effectiveness in evaluating and comparing the different heat recovery components and power generation cycles. These results indicate that the economizer is the most beneficial option for heat recovery and that the use of waste heat in a bagasse dryer is the least desirable option. Quantitative comparisons of several possible cycle options with the widely-used traditional back-pressure turbine cycle are given. These indicate that a double extraction condensing cycle is best for co-generation purposes. Power generation gains between 40 and

Heating unit of Berovo by co-generation (Macedonia)

International Nuclear Information System (INIS)

Armenski, Slave; Dimitrov, Konstantin; Tashevski, Done

1999-01-01

A plant for combined heat and electric power production, for central heating of the town Berovo (Macedonia) is proposed. The common reason to use a co-generation unit is the energy efficiency and a significant reduction of environmental pollution. The heat consumption of town Berovo is analyzed and determined. Based on the energy consumption of a whole power plant, e. i. the plant for combined and simultaneous production of power is proposed. The quantity of annually heat and electrical production and annually coal consumption are estimated. (Author)

Heat savings in energy systems with substantial distributed generation

DEFF Research Database (Denmark)

Østergaard, PA

2003-01-01

In Denmark, the integration of wind power is affected by a large amount of cogeneration of heat and power. With ancillary services supplied by large-scale condensation and combined heat and power (CHP) plants, a certain degree of large-scale generation is required regardless of momentary wind input......, if a certain production is required regardless of whether over-all electricity generation is sufficient. This article analyses this and although heat savings do have a negative impact on the amount of wind power the system may integrate a given moment in certain cases, associated fuel savings are notable...

Residual heat use generated by a 12 kW fuel cell in an electric vehicle heating system

International Nuclear Information System (INIS)

Colmenar-Santos, Antonio; Alberdi-Jiménez, Lucía; Nasarre-Cortés, Lorenzo; Mora-Larramona, Joaquín

2014-01-01

A diesel or gasoline vehicle heating is produced by the heat of the engine coolant liquid. Nevertheless, electric vehicles, due to the fact that electric motor transform directly electricity into mechanical energy through electromagnetic interactions, do not generate this heat so other method of providing it has to be developed. This study introduces the system developed in a fuel cell electric vehicle (lithium-ion battery – fuel cell) with residual heat use. The fuel cell electric vehicle is driven by a 12 kW PEM (proton exchange membrane) fuel cell. This fuel cell has an operating temperature around 50 °C. The residual heat generated was originally wasted by interaction with the environment. The new developed heating system designed integrates the heat generated by the fuel cell into the heating system of the vehicle, reducing the global energy consumption and improving the global efficiency as well. - Highlights: • Modification of heating system was done by introducing the residual heat from fuel cell. • Maximum heat achieved by the heating radiator of 9.27 kW. • Reduction of the heat dissipation by the fuel cell cooling system 1.5 kW. • Total efficiency improvement of 20% with an autonomy increase of 21 km

Efficient thermo-mechanical generation of electricity from the heat of radioisotopes

International Nuclear Information System (INIS)

Cooke-Yarborough, E.H.; Yeats, F.W.

1975-01-01

The thermomechanical generator uses a thermomechanical oscillator to convert heat efficiently into a mechanical oscillation which in turn excites a suitable transducer to generate alternating electricity. The thermomechanical oscillator used is based on the Stirling cycle, but avoids the need for rotary motion and for sliding pistons by having a mechanically-resonant, spring-suspended displacer, and by using an oscillating metal diaphragm to provide the mechanical output. The diaphragm drives an alternator consisting of a spring-suspended permanent magnet oscillating between fixed pole pieces which carry the electrical power output windings. Because a thermomechanical generator is much more efficient than a thermo-electric generator at comparable temperatures, it is particularly suitable for use with a radioisotope heat source. The amounts of radioisotope and of shielding required are both greatly reduced. A machine heated by radioisotopes and delivering 10.7W ac at 80Hz began operating in October, 1974. Operating experience with this machine is reported, and these results, together with those obtained with higher-powered machines heated by other means, are used to calculate characteristics and performance of thermo-mechanical radioisotope generators capable of using heat sources such as the waste-management 90 Sr radioisotope sources becoming available from the US nuclear waste management programme. A design to use one of these heat sources in a 52-W underwater generator is described

Thermodynamic analysis of heat recovery steam generator in combined cycle power plant

Directory of Open Access Journals (Sweden)

Ravi Kumar Naradasu

2007-01-01

Full Text Available Combined cycle power plants play an important role in the present energy sector. The main challenge in designing a combined cycle power plant is proper utilization of gas turbine exhaust heat in the steam cycle in order to achieve optimum steam turbine output. Most of the combined cycle developers focused on the gas turbine output and neglected the role of the heat recovery steam generator which strongly affects the overall performance of the combined cycle power plant. The present paper is aimed at optimal utilization of the flue gas recovery heat with different heat recovery steam generator configurations of single pressure and dual pressure. The combined cycle efficiency with different heat recovery steam generator configurations have been analyzed parametrically by using first law and second law of thermodynamics. It is observed that in the dual cycle high pressure steam turbine pressure must be high and low pressure steam turbine pressure must be low for better heat recovery from heat recovery steam generator.

An empirical test of pseudo random number generators by means of an exponential decaying process

International Nuclear Information System (INIS)

Coronel B, H.F.; Hernandez M, A.R.; Jimenez M, M.A.; Mora F, L.E.

2007-01-01

Empirical tests for pseudo random number generators based on the use of processes or physical models have been successfully used and are considered as complementary to theoretical tests of randomness. In this work a statistical methodology for evaluating the quality of pseudo random number generators is presented. The method is illustrated in the context of the so-called exponential decay process, using some pseudo random number generators commonly used in physics. (Author)

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Modelling of decay heat removal using large water pools

International Nuclear Information System (INIS)

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

1992-01-01

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

Heat wave experiments on the W7-AS stellarator

International Nuclear Information System (INIS)

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

1993-01-01

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

Dryout in sodium-heated helically-coiled steam generator tubes

International Nuclear Information System (INIS)

Tomita, Y.; Kosugi, T.; Kubota, J.; Nakajima, K.; Tsuchiya, T.

1984-01-01

Experimental research on the dryout phenomenon in sodium heated, helically coiled steam generator tubes was carried out. The fluctuation of the tube wall temperature caused by dryout was measured with thermocouples installed in the center of the tube wall. Empirical correlations of dryout quality were developed as functions of critical heat flux, water mass velocity and saturation pressure. These correlations confirmed that the design criterion of the MONJU steam generator was reasonable. (author)

The effect of heat generation in inclined slats on the natural convective heat transfer from an isothermal heated vertical plate

International Nuclear Information System (INIS)

Oosthuizen, P.H.; Sun, L.; Naylor, D.

2003-01-01

Natural convective heat transfer from a wide heated vertical isothermal plate with adiabatic surfaces above and below the heated surface has been considered. There are a series of equally spaced vertical thin, flat surfaces (termed 'slats') near the heated surface, these surfaces being, in general, inclined to the heated surface. There is, in general, a uniform heat generation in the slats. The slats are pivoted about their centre-point and thus as their angle is changed, the distance of the tip of the slat from the plate changes. The situation considered is an approximate model of a window with a vertical blind, the particular case where the window is hotter than the room air being considered. The heat generation in the slats in this situation is the result of solar radiation passing through the window and falling on and being absorbed by the slats of the blind. The flow has been assumed to be laminar and steady. Fluid properties have been assumed constant except for the density change with temperature that gives rise to the buoyancy forces. The governing equations have been written in dimensionless form and the resulting dimensionless equations have been solved using a commercial finite-element package. The solution has the following parameters: (1) the Rayleigh number (2) the Prandtl number (3) the dimensionless heat generation rate in the slats per unit frontal area (4) the dimensionless distance of the slat center point (the pivot point) from the surface (5) the dimensionless slat size (6) the dimensionless slat spacing (7) the angle of inclination of the slats. Because of the application that motivated the study, results have only been obtained for a Prandtl number of 0.7. The effect of the other dimensionless variables on the mean dimensionless heat transfer rate from the heated vertical surface has been examined. (author)

Fuel cell - An alternative for power and heat generating

International Nuclear Information System (INIS)

Zubcu, Victor; Ursescu, Gabriel; Zubcu, Dorina Silvia; Miler, Mihai Cristian

2004-01-01

One of the most promising energy generating technologies is the fuel cell (FC) because of its high efficiency and low emissions. There are even zero chemical emissions FC and cogeneration plants based on FC generate low heat emissions too. FC was invented 160 years ago but it was usually used only since 1960 in space missions. A FC farm tractor was tested 40 years ago. FC was again taken into account by power engineering since 1990 and it is now considered a credible alternative to power and heat generating. The thermal power engineers (and not only they) have two problems of cardinal importance for mankind to solve: - Energy saving (by increasing of energy generating efficiency) and - Environmental protection (by reducing chemical and heat emissions). The possibilities to use FC to generate power and heat are practically endless: on the earth, in the air and outer space, by and under water, in numberless areas of human activities. FC are now powering buses, cars, trains, boats, plains, scooters, highway road signs etc. There are already miniature FC for portable electronics. Homes, schools, hospitals, institutes, banks, police stations, etc are using FC to generate power and heat for their facilities. The methane gas produced by wastewater treatment plants and landfills is converted into electricity by using FC. Being less expensive than nuclear and solar source of energy, FC is now generally used in the space missions (in addition FC generates water). In this work an analysis of the possibilities to use FC especially for combined power and heat generating is presented. FC is favourite as energy source in space missions because it is less expensive than nuclear or solar sources. All major automobile companies have FC powered automobiles in testing stage. Mini FC for phone, laptop, and electronics are already on market. FC will be use to pagers, video recorders, small portable tools, miniature robots, special devices as hearing aid various devices, smoke detectors

Heat generation: prices have only a minor influence

International Nuclear Information System (INIS)

Stadelmann, M.

2006-01-01

This article takes a look at long-term trends in the heat generation market. Here, heat-pumps, gas heaters and wood-fired systems, together with their combination with solar collectors, are gaining ground, whereas heating oil is loosing its share of the market. The various influences on the market and, in particular, price increases for oil are discussed. The influence of revised energy legislation is discussed, which calls for 20% of the standardised energy requirements of housing to be met by renewables or increased thermal insulation. Increased sales in the solar sector are discussed, as are future trends in the heating market

Natural convection in porous media with heat generation

International Nuclear Information System (INIS)

Hardee, H.C. Jr.; Nilson, R.H.

1976-12-01

Heat transfer characteristics of a fluid saturated porous media are investigated for the case of uniform internal heat generation with cooling from above. Analytical models of conduction and single phase cellular convection show good agreement with previous Rayleigh number correlations and with experimental data obtained by Joule heating of salt water in a sand bed. An approximate dryout criterion is also derived for two phase boiling heat transfer in a fixed bed which is neither channeled nor fluidized. Correlation of dryout data using this criterion is encouraging, especially considering the analytical rather than correlational basis of the criterion

From Modules to a Generator: An Integrated Heat Exchanger Concept for Car Applications of a Thermoelectric Generator

Science.gov (United States)

Bosch, Henry

2016-03-01

A heat exchanger concept for a thermoelectric generator with integrated planar modules for passenger car applications is introduced. The module housings, made of deep drawn stainless steel sheet metal, are brazed onto the exhaust gas channel to achieve an optimal heat transfer on the hot side of the modules. The cooling side consists of winding fluid channels, which are mounted directly onto the cold side of the modules. Only a thin foil separates the cooling media from the modules for an almost direct heat contact on the cooling side. Thermoelectric generators with up to 20 modules made of PbTe and Bi2Te3, respectively, are manufactured and tested on a hot gas generator to investigate electrical power output and performance of the thermoelectric generator. The proof of concept of the light weight heat exchanger design made of sheet metal with integrated modules is positively accomplished.

Dry out of a fluidized particle bed with internal heat generation

International Nuclear Information System (INIS)

Keowen, R.S.; Catton, I.

1975-03-01

An apparatus was designed to adequately simulate the characteristics of a particle bed formed by nuclear reactor fuel after the reactor has been operable for some length of time at high power. This was accomplished by using a 10 KW, 453 Kc induction heater, coupled through a multi-turn work coil to particle beds of cast steel shot and lead shot in water. The temperature response and dryout condition was determined for various bed levels, particle diameters, and heat fluxes. Analysis of the data retrieved from the bed was used to generate a family of curves to predict the necessary conditions for dryout to occur within a fluidized particle bed with internal heat generation. The results presented here, with internal heat generation, show that previous results with bottom heating and volume heating are conservative. (U.S.)

Impact of squark generation mixing on the search for squarks decaying into fermions at LHC

International Nuclear Information System (INIS)

Bartl, A.; Eberl, H.; Majerotto, W.; Herrmann, B.; Wuerzburg Univ.; Hidaka, K.; Porod, W.; Valencia Univ.

2010-07-01

We study the effect of squark generation mixing on squark production and decays at LHC in the Minimal Supersymmetric Standard Model (MSSM).We show that the effect can be very large despite the very strong constraints on quark flavour violation (QFV) from experimental data on B mesons. We find that the two lightest up-type squarks u 1,2 can have large branching ratios for the decays into cχ 0 1 and tχ 0 1 at the same time due to squark generation mixing, leading to QFV signals 'pp→ c anti t(t anti c)+missing-E T + X' with a significant rate. The observation of this remarkable signature would provide a powerful test of supersymmetric QFV at LHC. This could have a significant impact on the search for squarks and the determination of the underlying MSSM parameters. (orig.)

Experimental studies of parameters affecting the heat generation in friction stir welding process

Directory of Open Access Journals (Sweden)

Mijajlović Miroslav M.

2012-01-01

Full Text Available Heat generation is a complex process of transformation of a specific type of energy into heat. During friction stir welding, one part of mechanical energy delivered to the welding tool is consumed in the welding process, another is used for deformational processes etc., and the rest of the energy is transformed into heat. The analytical procedure for the estimation of heat generated during friction stir welding is very complex because it includes a significant number of variables and parameters, and many of them cannot be fully mathematically explained. Because of that, the analytical model for the estimation of heat generated during friction stir welding defines variables and parameters that dominantly affect heat generation. These parameters are numerous and some of them, e. g. loads, friction coefficient, torque, temperature, are estimated experimentally. Due to the complex geometry of the friction stir welding process and requirements of the measuring equipment, adequate measuring configurations and specific constructional solutions that provide adequate measuring positions are necessary. This paper gives an overview of the process of heat generation during friction stir welding, the most influencing parameters on heat generation, constructional solutions for the measuring equipment needed for these experimental researches and examples of measured values.

Numerical investigation of conjugate heat transfer and flow performance of a fin and tube heat exchanger with vortex generators

DEFF Research Database (Denmark)

Singh, Shobhana; Sørensen, Kim

2017-01-01

Vortex generator is considered as an effective device for augmentation of the thermal-hydraulic performance of a heat exchanger. The aim of present study is to examine the influence of vortex generators on a double fin and tube heat exchanger performance. Vortex generator of rectangular winglet...

Analysis of a sandwich-type generator with self-heating thermoelectric elements

International Nuclear Information System (INIS)

Kim, Mikyung; Yang, Hyein; Wee, Daehyun

2014-01-01

Highlights: • A novel and unique type of thermoelectric generators is proposed. • Heat source is combined in thermoelectric elements, reducing heat transfer problems. • Embedding radioactive isotopes is proposed as a way to implement the new design. • Conversion efficiency and power density are estimated for the proposed design. - Abstract: A novel and unique design of thermoelectric generators, in which a heat source is combined with thermoelectric elements, is proposed. By placing heat-generating radioactive isotopes inside the thermoelectric elements, the heat transfer limitation between the generator and the heat source can be eliminated, ensuring simplicity. The inner electrode is sandwiched between identical thermoelectric elements, which naturally allows the inner core to act as the hot side. Analysis shows that conversion efficiency and power density increase as the heat density inside the thermoelectric elements increases and as the thermoelectric performance of the material improves. The theoretical maximum efficiency is shown to be 50%. However, realistic performance under practical constraint is much worse. In realistic cases, the efficiency would be about 3% at best. The power density of the proposed design exhibits a much more reasonable value as high as 3000 W/m 2 . Although the efficiency is low, the simplicity of the proposed design combined with its reasonable power density may result in some, albeit limited, potential applications. Further investigation must be performed in order to realize such potential

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

International Nuclear Information System (INIS)

Schubert, B.K.

1984-07-01

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

Modification of the collective Thomson scattering radiometer in the search for parametric decay on TEXTOR

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Salewski, Mirko; Bongers, W.

2012-01-01

Strong scattering of high-power millimeter waves at 140 GHz has been shown to take place in heating and current-drive experiments at TEXTOR when a tearing mode is present in the plasma. The scattering signal is at present supposed to be generated by the parametric decay instability. Here we descr...

An analytical model for the heat generation in friction stir welding

DEFF Research Database (Denmark)

Schmidt, Henrik Nikolaj Blich; Hattel, Jesper; Wert, John

2004-01-01

The objective of this work is to establish an analytical model for heat generation by friction stir welding (FSW), based on different assumptions of the contact condition between the rotating tool surface and the weld piece. The material flow and heat generation are characterized by the contact...

Dynamic study of steam generation from low-grade waste heat in a zeolite–water adsorption heat pump

International Nuclear Information System (INIS)

Xue, Bing; Meng, Xiangrui; Wei, Xinli; Nakaso, Koichi; Fukai, Jun

2015-01-01

A novel zeolite–water adsorption heat pump system based on a direct-contact heat exchange method to generate steam from low-grade waste gas and water has been proposed and examined experimentally. Superheated steam (200 °C, 0.1 MPa) is generated from hot water (70–80 °C) and dry air (100–130 °C). A dynamic model for steam generation process is developed to describe local mass and heat transfer. This model features a three-phase calculation and a moving water–gas interface. The calculations are carried out in the zeolite–water and zeolite–gas regions. Model outputs are compared with experimental results for validation. The thermal response inside the reactor and mass of steam generated is well predicted. Numerical results show that preheat process with low-temperature steam is an effective method to achieve local equilibrium quickly, thus generation process is enhanced by prolonging the time and increasing mass of the generated steam. Besides, high-pressure steam generation up to 0.5 MPa is possible from the validated dynamic model. Future work could be emphasized on enhancing high-pressure steam generation with preheat process or mass recovery operation

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

International Nuclear Information System (INIS)

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

2010-01-01

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

THE EFFECTS OF SWIRL GENERATOR HAVING WINGS WITH HOLES ON HEAT TRANSFER AND PRESSURE DROP IN TUBE HEAT EXCHANGER

Directory of Open Access Journals (Sweden)

Zeki ARGUNHAN

2006-02-01

Full Text Available This paper examines the effect of turbulance creators on heat transfer and pressure drop used in concentric heat exchanger experimentaly. Heat exchanger has an inlet tube with 60 mm in diameter. The angle of swirl generators wings is 55º with each wing which has single, double, three and four holes. Swirl generators is designed to easily set to heat exchanger entrance. Air is passing through inner tube of heat exhanger as hot fluid and water is passing outer of inner tube as cool fluid.

Decay assessment through thermographic analysis in architectural and archaeological heritage

Science.gov (United States)

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

2010-05-01

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

Utilization of Aluminum Waste with Hydrogen and Heat Generation

Science.gov (United States)

Buryakovskaya, O. A.; Meshkov, E. A.; Vlaskin, M. S.; Shkolnokov, E. I.; Zhuk, A. Z.

2017-10-01

A concept of energy generation via hydrogen and heat production from aluminum containing wastes is proposed. The hydrogen obtained by oxidation reaction between aluminum waste and aqueous solutions can be supplied to fuel cells and/or infrared heaters for electricity or heat generation in the region of waste recycling. The heat released during the reaction also can be effectively used. The proposed method of aluminum waste recycling may represent a promising and cost-effective solution in cases when waste transportation to recycling plants involves significant financial losses (e.g. remote areas). Experiments with mechanically dispersed aluminum cans demonstrated that the reaction rate in alkaline solution is high enough for practical use of the oxidation process. In theexperiments aluminum oxidation proceeds without any additional aluminum activation.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during stead-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The Semiscale (MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Assessment of capability of models for prediction of pressure drop and dryout heat flux in a heat generating particulate debris bed

International Nuclear Information System (INIS)

Kulkarni, P.P.; Nayak, A.K.; Rashid, M.; Kulenovic, R.

2009-01-01

During a severe accident in a light water reactor, the core can melt and be relocated to the lower plenum of the reactor pressure vessel. There it can form a particulate debris bed due to the possible presence of water. This bed, if not quenched in time, can lead to the failure of the pressure vessel because of the insufficient heat removal of decay heat in the debris bed. Therefore, addressing the issue of coolability behaviour of heat generating particulate debris bed is of prime importance in the framework of severe accident management strategies, particularly in case of above mentioned late phase scenario of an accident. In order to investigate the coolability behaviour of particulate debris bed, experiments were carried out at IKE test facility 'DEBRIS' on particle beds of irregularly shaped particles mixed with spheres under top- and bottom-flooding condition. The pressure drop and dryout heat flux (DHF) were measured for top- and bottom-flooding conditions. For top-flooding conditions, it was found that the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial shear stress of the two-phase flow. For bottom-flooding with a relatively high liquid inflow velocity, the pressure gradient increases consistently with the vapour velocity and the fluid-particle drags become important. Also, with additional forced liquid inflow from the bottom, the DHF increases dramatically. In all the cases, it was found that the DHF is significantly larger with bottom-flooding condition compared to top-flooding condition. Different models such as Lipinski, Reed, Tung and Dhir, Hu and Theophanous, and Schulenberg and Mueller have been used to predict the pressure drop characteristics and the DHF of heat generating particulate debris beds. Comparison is made among above mentioned models and experimental results for DHF and pressure drop characteristics. Considering the overall trend in

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

Science.gov (United States)

Gustafson, Stephen; Roxanas, Dimitrios

2018-05-01

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

Impact of Next Generation District Heating Systems on Distribution Network Heat Losses: A Case Study Approach

Science.gov (United States)

Li, Yu; Rezgui, Yacine

2018-01-01

District heating (DH) is a promising energy pathway to alleviate environmental negative impacts induced by fossil fuels. Improving the performance of DH systems is one of the major challenges facing its wide adoption. This paper discusses the heat losses of the next generation DH based on the constructed Simulink model. Results show that lower distribution temperature and advanced insulation technology greatly reduce network heat losses. Also, the network heat loss can be further minimized by a reduction of heat demand in buildings.

Quantity, Quality, and Availability of Waste Heat from United States Thermal Power Generation.

Science.gov (United States)

Gingerich, Daniel B; Mauter, Meagan S

2015-07-21

Secondary application of unconverted heat produced during electric power generation has the potential to improve the life-cycle fuel efficiency of the electric power industry and the sectors it serves. This work quantifies the residual heat (also known as waste heat) generated by U.S. thermal power plants and assesses the intermittency and transport issues that must be considered when planning to utilize this heat. Combining Energy Information Administration plant-level data with literature-reported process efficiency data, we develop estimates of the unconverted heat flux from individual U.S. thermal power plants in 2012. Together these power plants discharged an estimated 18.9 billion GJ(th) of residual heat in 2012, 4% of which was discharged at temperatures greater than 90 °C. We also characterize the temperature, spatial distribution, and temporal availability of this residual heat at the plant level and model the implications for the technical and economic feasibility of its end use. Increased implementation of flue gas desulfurization technologies at coal-fired facilities and the higher quality heat generated in the exhaust of natural gas fuel cycles are expected to increase the availability of residual heat generated by 10.6% in 2040.

Radioactive wastes with negligible heat generation suitable for disposal

International Nuclear Information System (INIS)

Brennecke, P.; Schumacher, J.; Warnecke, E.

1987-01-01

It is planned to dispose of radioactive wastes with negligible heat generation in the Konrad repository. Preliminary waste acceptance requirements are derived taking the results of site-specific safety assessments as a basis. These requirements must be fulfilled by the waste packages on delivery. The waste amounts which are currently stored and those anticipated up to the year 2000 are discussed. The disposability of these waste packages in the Konrad repository was evaluated. This examination reveals that basically almost all radioactive wastes with negligible heat generation can be accepted. (orig.) [de

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

International Nuclear Information System (INIS)

Natta, M.; Bloch, M.

1991-01-01

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

Underground disposal of UK heat-generating wastes: repository design considerations

International Nuclear Information System (INIS)

Steadman, J.A.

1993-12-01

The report discusses the likely differences in design between a deep repository for disposal of UK heat-generating radioactive wastes and that of the planned Nirex ILW/LLW repository at Sellafield, based on a review of international published information. The main differences arise from the greater heat and radiation outputs of the waste, and in the case of intact PWR spent fuel elements, the greater length and weight of the disposal packages. Published cost estimates for other OECD countries for disposal of heat-generating wastes are considerably lower than that for the UK, partly because in most cases they are for co-disposal with a larger quantity of ILW. (author)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Modeling a Thermoelectric Generator Applied to Diesel Automotive Heat Recovery

Science.gov (United States)

Espinosa, N.; Lazard, M.; Aixala, L.; Scherrer, H.

2010-09-01

Thermoelectric generators (TEGs) are outstanding devices for automotive waste heat recovery. Their packaging, lack of moving parts, and direct heat to electrical conversion are the main benefits. Usually, TEGs are modeled with a constant hot-source temperature. However, energy in exhaust gases is limited, thus leading to a temperature decrease as heat is recovered. Therefore thermoelectric properties change along the TEG, affecting performance. A thermoelectric generator composed of Mg2Si/Zn4Sb3 for high temperatures followed by Bi2Te3 for low temperatures has been modeled using engineering equation solver (EES) software. The model uses the finite-difference method with a strip-fins convective heat transfer coefficient. It has been validated on a commercial module with well-known properties. The thermoelectric connection and the number of thermoelements have been addressed as well as the optimum proportion of high-temperature material for a given thermoelectric heat exchanger. TEG output power has been estimated for a typical commercial vehicle at 90°C coolant temperature.

MHTGR steam generator on-line heat balance, instrumentation and function

International Nuclear Information System (INIS)

Klapka, R.E.; Howard, W.W.; Etzel, K.T.; Basol, M.; Karim, N.U.

1991-09-01

Instrumentation is used to measure the Modular High Temperature Gas-Cooled Reactor (MHTGR) steam generator dissimilar metal weld temperature during start-up testing. Additional instrumentation is used to determine an on-line heat balance which is maintained during the 40 year module life. In the process of calibrating the on-line heat balance, the helium flow is adjusted to yield the optimum boiling level in the steam generator relative to the dissimilar metal weld. After calibration is complete the weld temperature measurement is non longer required. The reduced boiling level range results in less restrictive steam generator design constraints

Quantitative Understanding on the Amplitude Decay Characteristic of the Evanescent Electromagnetic Waves Generated by Seismoelectric Conversion

Science.gov (United States)

Ren, Hengxin; Huang, Qinghua; Chen, Xiaofei

2018-03-01

We conduct numerical simulations and theoretical analyses to quantitatively study the amplitude decay characteristic of the evanescent electromagnetic (EM) waves, which has been neglected in previous studies on the seismoelectric conversion occurring at a porous-porous interface. Time slice snapshots of seismic and EM wave-fields generated by a vertical single force point source in a two-layer porous model show that evanescent EM waves can be induced at a porous-porous interface. The seismic and EM wave-fields computed for a receiver array located in a vertical line nearby the interface are investigated in detail. In addition to the direct and interface-response radiation EM waves, we identify three groups of coseismic EM fields and evanescent EM waves associated with the direct P, refracted SV-P and direct SV waves, respectively. Thereafter, we derive the mathematical expression of the amplitude decay factor of the evanescent EM waves. This mathematical expression is further validated by our numerical simulations. It turns out the amplitude decay of the evanescent EM waves generated by seismoelectric conversion is greatly dependent on the horizontal wavenumber of seismic waves. It is also found the evanescent EM waves have a higher detectability at a lower frequency range. This work provides a better understanding on the EM wave-fields generated by seismoelectric conversion, which probably will help improve the interpretation of the seismoelectric coupling phenomena associated with natural earthquakes or possibly will inspire some new ideas on the application of the seismoelectric coupling effect.

An evaluation of analytical heat transfer area with various boiling heat transfer correlations in steam generator thermal sizing

International Nuclear Information System (INIS)

Jung, B. R.; Park, H. S.; Chung, D. M.; Baik, S. J.

1999-01-01

The computer program SAFE has been used to size and analyze the performance of a steam generator which has two types of heat transfer regions in Korean Standard Nuclear Power Plants (KSNP) and Korean Next Generation Reactor (KNGR) design. The SAFE code calculates the analytical boiling heat transfer area using the modified form of the saturated nucleate pool boiling correlation suggested by Rohsenow. The predicted heat transfer area in the boiling region is multiplied by a constant to obtain a final analytical heat transfer area. The inclusion of the multiplier in the analytical calculation has some disadvantage of loss of complete correlation by the governing heat transfer equation. Several comparative analyses have been performed quantitatively to evaluate the possibility of removing the multiplier in the analytical calculation in the SAFE code. The evaluation shows that the boiling correlation and multiplier used in predicting the boiling region heat transfer area can be replaced with other correlations predicting nearly the same heat transfer area. The removal of multiplier included in the analytical calculation will facilitate a direct use of a set of concerned analytical sizing values that can be exactly correlated by the governing heat transfer equation. In addition this will provide more reasonable basis for the steam generator thermal sizing calculation and enhance the code usability without loss of any validity of the current sizing procedure. (author)

Improvement to the gas cycle energy generating installations with heat recuperation

International Nuclear Information System (INIS)

Tilliette, Zephyr.

1977-01-01

Improvement to the gas cycle energy generating installations with heat recuperation, comprising a heat source, supplying a fluid at high temperature and pressure, an expansion turbine, at least one recuperator fitted to the turbine outlet, a cooler and compressor in series, the compressor returning the high pressure fluid to the source after heat exchange in the recuperator with the low pressure fluid from the turbine. It is characterised in that at least one steam generator is connected to the low pressure end of the recuperator [fr

Power performance of the general-purpose heat source radioisotope thermoelectric generator

International Nuclear Information System (INIS)

Bennett, G.L.; Lombardo, J.J.; Rock, B.J.

1986-01-01

The General-Purpose Heat Source Radioisotope Thermoelectric Generator (GRHS-RTG) has been developed under the sponsorship of the Department of Energy (DOE) to provide electrical power for the National Aeronautics and Space Administration (NASA) Galileo mission to Jupiter and the joint NASA/European Space Agency (ESA) Ulysses mission to study the polar regions of the sun. A total of five nuclear-heated generators and one electrically heated generator have been built and tested, proving out the design concept and meeting the specification requirements. The GPHS-RTG design is built upon the successful-technology used in the RTGs flown on the two NASA Voyager spacecraft and two US Air Force communications satellites. THe GPHS-RTG converts about 4400 W(t) from the nuclear heat source into at least 285 W(e) at beginning of mission (BOM). The GPHS-RTG consists of two major components: the General-Purpose Heat Source (GPHS) and the Converter. A conceptual drawing of the GPHs-RTG is presented and its design and performance are described

Thermal analysis and entropy generation of pulsating heat pipes using nanofluids

International Nuclear Information System (INIS)

Jafarmadar, Samad; Azizinia, Nazli; Razmara, Nayyer; Mobadersani, Farrokh

2016-01-01

Highlights: • Performance of PHP containing 0.5% Al_2O_3, CuO and silver nanofluids is reported. • The rate of entropy generation of PHP is investigated for different nanofluids. • The effects of particle volume concentration on the entropy generation of PHP are studied. • The appropriate volume concentration for the best thermal efficiency is 0.5–1%. • Al_2O_3 and CuO nanofluids show approximately same rate of entropy generation. - Abstract: Demanding of high-performance cooling systems is one of the most challenging and virtual issues in the industry and pulsating heat pipes (PHPs) are effective solutions for this concern. Nanofluids also have attracted attentions, due to its superior heat transfer properties in recent years. In the present study, the flow, heat transfer and entropy generation based on the second law of thermodynamics have been investigated and compared with the flow of Al_2O_3, CuO, Ag nanofluid and pure water through PHPs. The results show that, silver nanofluid provides the highest entropy generation. Also, the effects of different particle volume concentrations on the heat and flow characteristics of Al_2O_3 nanofluid have been studied. It is indicated that the optimal volume concentration of nanoparticles is about 0.5–1% to minimize the entropy generation and appropriate thermal operation.

Measurements of Critical Heat Flux using Mass Transfer System

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Chromatographic generator systems for the actinides and natural decay series elements

International Nuclear Information System (INIS)

McAlister, D.R.; Horwitz, E.P.

2011-01-01

This work describes chromatographic radionuclide generator systems for the production of actinides and natural decay series elements. The generator systems begin with alpha emitting parent radioisotopes with half-lives (T 1/2 ) of greater than one year and produce alpha or beta emitting radioisotopes with half-lives of hours to days. Chromatographic systems were chosen to minimize radiolytic damage to chromatographic supports, preserve the parent activity for repeated use, provide high purity daughter radionuclide tracers, and to minimize or eliminate the need for evaporation of solutions of the parent or daughter nuclides. Useful secondary separations involving the daughters of the initial parent radionuclide are also described. Separation systems for 210 Bi, 210 Po, 211 Pb, 212 Pb, 223 Ra, 224 Ra, 225 Ra, 225 Ac, 227 Th, 228 Th, 231 Th, 234 Th, and 239 Np are outlined in detail. (orig.)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Economic aspects of electricity and industrial heat generating reactors

International Nuclear Information System (INIS)

Gaussens, J.; Moulle, N.; Dutheil, F.

1964-01-01

The economic advantage of electricity-generating nuclear stations decreases when their size decreases. However, when a counter-pressure turbine is joined on to a reactor and the residual heat can be properly used, it can be shown that fairly low capacity nuclear equipment may compete with conventional equipment under certain realistic enough conditions. The aim of this paper is to define these special conditions under which nuclear energy can be profitable. They are connected with the location and the general economic environment of the station, the pattern of the electricity and heat demands it must meet, the level of fuel and specific capital costs, nuclear and conventional. These conditions entail certain technical and economic specifications for the reactors used in this way otherwise they are unlikely to be competitive. In addition, these results are referred to the potential steam and electricity market, which leads us to examine certain uses for the heat generated by double purpose power stations; for example, to supply combined industrial plants, various types of town heating and for removal of salt from sea water. (authors) [fr

Numerical investigation of passive heat removal system via steam generator in VVER 1200

International Nuclear Information System (INIS)

Dinh Anh Tuan; Duong Thanh Tung; Tran Chi Thanh; Nguyen Van Thai

2015-01-01

Passive heat removal system (PHRS) via Steam Generator is an important part in VVER design. In case of Design Basic Accidents such as blackout, failure of feed water supply to steam generator or coolant leakage with failure of emergency core cooling at high pressure. PHRS is designed to remove the residual heat from reactor core through steam generator to heat exchanger which is placed outside reactor vessel. In order to evaluate the passive system, a numerical investigation using a CFD code is performed. However, PHRS has complex geometry for using CFD simulation. Thus, RELAP5 is applied to provide the wall heat flux of tube in the heat exchanger tank. The natural convection in the heat exchanger tank is investigated in this report. Numerical results show temperature and velocity distribution in the heat exchanger tank are calculated with different wall heat flux corresponding to various transient conditions. The calculated results contribute to the capacity analysis of passive heat removal system and giving valuable information for safe operation of VVER 1200. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Heat shrink formation of a corrugated thin film thermoelectric generator

International Nuclear Information System (INIS)

Sun, Tianlei; Peavey, Jennifer L.; David Shelby, M.; Ferguson, Scott; O’Connor, Brendan T.

2015-01-01

Highlights: • Demonstrate and characterize a thermoelectric generator with a corrugated geometry. • Employ a novel heat shrink fabrication approach compatible with low-cost processing. • Use thermal impedance modeling to explore design potential. • Corrugated design shown to be advantageous for low heat-flux density applications. - Abstract: A thin film thermoelectric (TE) generator with a corrugated architecture is demonstrated formed using a heat-shrink fabrication approach. Fabrication of the corrugated TE structure consists of depositing thin film thermoelectric elements onto a planar non-shrink polyimide substrate that is then sandwiched between two uniaxial stretch-oriented co-polyester (PET) films. The heat shrink PET films are adhered to the polyimide in select locations, such that when the structure is placed in a high temperature environment, the outer films shrink resulting in a corrugated core film and thermoelectric elements spanning between the outer PET films. The module has a cross-plane heat transfer architecture similar to a conventional bulk TE module, but with heat transfer in the plane of the thin film thermoelectric elements, which assists in maintaining a significant temperature difference across the thermoelectric junctions. In this demonstration, Ag and Ni films are used as the thermoelectric elements and a Seebeck coefficient of 14 μV K −1 is measured with a maximum power output of 0.22 nW per couple at a temperature difference of 7.0 K. We then theoretically consider the performance of this device architecture with high performance thermoelectric materials in the heat sink limited regime. The results show that the heat-shrink approach is a simple fabrication method that may be advantageous in large-area, low power density applications. The fabrication method is also compatible with simple geometric modification to achieve various form factors and power densities to customize the TE generator for a range of applications

Modelling and verification of once-through subcritical heat recovery steam generator

International Nuclear Information System (INIS)

Lee, Chae Soo; Choi, Young Jun; Kim, Hyun Gee; Yang, Ok Chul; Chong Chae Hon

2004-01-01

The once-through heat recovery steam generator is ideally matched to very high temperature and pressure, well into the supercritical range. Moreover this type of boiler is structurally simpler than drum type boiler. In drum type boiler, each tube play a well-defined role: water preheating, vaporization, superheating. Empirical equations are available to predict the average heat transfer coefficient for each regime. For once-through heat recovery steam generator, this is no more the case and mathematical models have to be adapted to account for the disappearance of drum type economizer, boiler, superheater. General equations have to be used for each tube of boiler, and actual heat transfer condition in each tube has to be identified

Heat production / host rock compatibility; Waermeentwicklung / Gesteinsvertraeglichkeit

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Family symmetries and proton decay

International Nuclear Information System (INIS)

Murayama, Hitoshi; Kaplan, D.B.

1994-01-01

The proton decay modes p → K 0 e + and p → K 0 μ + may be visible in certain supersymmetric theories, and if seen would provide evidence for new flavor physics at extremely short distances. These decay modes can arise from the dimension five operator (Q 1 Q 1 Q 2 L 1,2 ), where Q i and L i are i th generation quark and lepton superfields respectively. Such an operator is not generated at observable levels due to gauge or Higgs boson exchange in a minimal GUT. However in theories that explain the fermion mass hierarchy, it may be generated at the Planck scale with a strength such that the decays p → K 0 ell + are both compatible with the proton lifetime and visible at Super-Kamiokande. Observable proton decay can even occur in theories without unification

Testing for cross-subsidisation in the combined heat and power generation sector

DEFF Research Database (Denmark)

Amundsen, Eirik S; Andersen, Per; Jensen, Frank

2011-01-01

In this paper we examine cross-subsidisation among combined heat and power producers in Denmark.Information on stand-alone costs for heat generation allows us to empirically compare the Faulhaber tests,tests with an upper bound on stand-alone costs (the Palmer tests) and the fully distributed cos...... test (FDC). All tests indicate a substantial amount of cross-subsidisation from heat generation to power generation. It is shown that the FDC test is closer to that of the Faulhaber tests in its results than the Palmer tests. Thus as the Faulhaber tests are considered in the literature...

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Research and Development Supporting a Next Generation Germanium Double Beta Decay Experiment

Science.gov (United States)

Rielage, Keith; Elliott, Steve; Chu, Pinghan; Goett, Johnny; Massarczyk, Ralph; Xu, Wenqin

2015-10-01

To improve the search for neutrinoless double beta decay, the next-generation experiments will increase in source mass and continue to reduce backgrounds in the region of interest. A promising technology for the next generation experiment is large arrays of Germanium p-type point contact detectors enriched in 76-Ge. The experience, expertise and lessons learned from the MAJORANA DEMONSTRATOR and GERDA experiments naturally lead to a number of research and development activities that will be useful in guiding a future experiment utilizing Germanium. We will discuss some R&D activities including a hybrid cryostat design, background reduction in cabling, connectors and electronics, and modifications to reduce assembly time. We acknowledge the support of the U.S. Department of Energy through the LANL/LDRD Program.

Search for Pair Produced Heavy Quarks Decaying into a Z Boson and a Light Generation Jet

CERN Document Server

AUTHOR|(SzGeCERN)672562

We present here our results from the search for the pair production of new massive quarks using $20\\ifb$ data collected by the ATLAS detector in 2012 proton-proton collisions at $\\rts=$8TeV centre of mass energy. There are many models extending beyond the well established Standard Model (SM) predicting additional massive quarks on top of the existing quark list. One of them is the grand unified theory (GUT) having the exceptional $E_{6}$ as its symmetry group. This model predicts an additional iso-singlet down type quark for each existing SM family with possible decay channels involving $\\Wpm,\\Zboson$ and $H$. Assuming a similar mass structure, the lightest of these quarks denoted by the letter $D$ would be the first one to be discovered at ATLAS. Our search focuses on the decay signature of $D$ quark decays via a $\\Zboson$ boson which would further decay into two leptons. The other heavy quark searches at ATLAS are carried out with the assumption that the new quarks should couple to the heavy generations $t$...

Large-Scale Combined Heat and Power (CHP) Generation at Loviisa Nuclear Power Plant Unit 3

International Nuclear Information System (INIS)

Bergroth, N.

2010-01-01

Fortum has applied for a Decision in Principle concerning the construction of a new nuclear power plant unit (Loviisa 3) ranging from 2800-4600 MWth at its site located at the southern coast of Finland. An attractive alternative investigated is a co-generation plant designed for large-scale district heat generation for the Helsinki metropolitan area that is located approximately 75 km west of the site. The starting point is that the district heat generation capacity of 3 unit would be around 1 000 MWth.The possibility of generating district heat for the metropolitan area by Loviisa's two existing nuclear power plant units was investigated back in the 1980s, but it proved unpractical at the time. With the growing concern of the climate change and the subsequent requirements on heat and power generation, the idea is much more attractive today, when recognising its potential to decrease Finland's carbon dioxide emissions significantly. Currently the district heat generation in metropolitan area is based on coal and natural gas, producing some five to seven million tonnes of carbon dioxide emissions annually. Large-scale combined heat and power (CHP) generation at the 3 unit could cut this figure by up to four million tonnes. This would decrease carbon dioxide emissions by as much as six percent. In addition, large-scale CHP generation would increase the overall efficiency of the new unit significantly and hence, reduce the environmental impact on the local marine environment by cutting heat discharges into the Gulf of Nuclear energy has been used for district heating in several countries both in dedicated nuclear heating plants and in CHP generation plants. However, the heat generation capacity is usually rather limited, maximum being around 250 MWth per unit. Set against this, the 3 CHP concept is much more ambitious, not only because of the much larger heat generation output envisaged, but also because the district heating water would have to be transported over a

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Science.gov (United States)

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

2017-05-01

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

Exergetic life cycle assessment of cement production process with waste heat power generation

International Nuclear Information System (INIS)

Sui, Xiuwen; Zhang, Yun; Shao, Shuai; Zhang, Shushen

2014-01-01

Highlights: • Exergetic life cycle assessment was performed for the cement production process. • Each system’s efficiency before and after waste heat power generation was analyzed. • The waste heat power generation improved the efficiency of each production system. • It provided technical support for the implementation of energy-saving schemes. - Abstract: The cement industry is an industry that consumes a considerable quantity of resources and energy and has a very large influence on the efficient use of global resources and energy. In this study, exergetic life cycle assessment is performed for the cement production process, and the energy efficiency and exergy efficiency of each system before and after waste heat power generation is investigated. The study indicates that, before carrying out a waste heat power generation project, the objective energy efficiencies of the raw material preparation system, pulverized coal preparation system and rotary kiln system are 39.4%, 10.8% and 50.2%, respectively, and the objective exergy efficiencies are 4.5%, 1.4% and 33.7%, respectively; after carrying out a waste heat power generation project, the objective energy efficiencies are 45.8%, 15.5% and 55.1%, respectively, and the objective exergy efficiencies are 7.8%, 2.8% and 38.1%, respectively. The waste heat power generation project can recover 3.7% of the total input exergy of a rotary kiln system and improve the objective exergy efficiencies of the above three systems. The study can identify degree of resource and energy utilization and the energy-saving effect of a waste heat power generation project on each system, and provide technical support for managers in the implementation of energy-saving schemes

Natural convection heat transfer in a rectangular pool with volumetric heat sources

International Nuclear Information System (INIS)

Lee, Seung Dong; Lee, Kang Hee; Suh, Kune Y.

2003-01-01

Natural convection plays an important role in determining the thermal load from debris accumulated in the reactor vessel lower head during a severe accident. The heat transfer within the molten core material can be characterized by buoyancy-induced flows resulting from internal heating due to decay of fission products. The thermo-fluid dynamic characteristics of the molten pool depend strongly on the thermal boundary conditions. The spatial and temporal variation of heat flux on the pool wall boundaries and the pool superheat are mainly characterized by the natural convection flow inside the molten pool. In general, natural convection involving internal heat generation is delineated in terms of the modified Rayleigh number, Ra', which quantifies the internal heat source and hence the strength of buoyancy. The test section is of rectangular cavity whose length, width, and height are 500 mm, 80 mm, and 250 mm, respectively. A total of twenty-four T-type thermocouples were installed in the test loop to measure temperature distribution. Four T-type thermocouples were utilized to measure temperatures on the boundary. A direct heating method was adopted in this test to simulate the uniform heat generation. The experiments covered a range of Rayleigh number, Ra, between 4.87x10 7 and 2.32x10 14 and Prandtl number, Pr, between 0.7 and 3.98. Tests were conducted with water and air as simulant. The upper and lower boundary conditions were maintained at a uniform temperature of 10degC. (author)

Analysis for a PRHRS Condensation Heat Exchanger of the SMART-P Plant

International Nuclear Information System (INIS)

Lee, Kwon-Yeong; Kim, Moo Hwan

2007-01-01

When an emergency such as the unavailability of feedwater or the loss of off-site power arises with SMART-P, the PRHRS passively removes the core decay heat via natural convection. The system is connected to the feedwater and steam pipes and consists of a heat exchanger submerged in a refueling water tank, a compensation tank, and check and isolation valves. The heat exchanger removes the heat from the reactor coolant system through a steam generator via condensation heat transfer to water in the refueling water tank. The compensating tank is pressurized using a nitrogen gas to make up the water volume change in the PRHRS. Before PRHRS operation, nitrogen may be dissolved in the cooling water of the PRHRS. Therefore, during PRHRS operation, nitrogen gas might be generated due to evaporation in the steam generator, which will act as a noncondensable gas in the condensation heat exchanger. The main objective of the present study was to assess the design of a PRHRS condensation heat exchanger (PRHRS HX) by investigating its heat transfer characteristics

Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust

International Nuclear Information System (INIS)

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

2002-01-01

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

Modes of heat removal from a heat-generating debris bed

International Nuclear Information System (INIS)

Squarer, D.; Hochreiter, L.E.; Piecznski, A.T.

1984-01-01

In the worst hypothetical accident in a light water reactor, when all protection systems fail, the core could be converted into a deep particulate bed either in-vessel or ex-vessel. The containment of such an accident depends on the coolability of a heat-generating debris bed. Some recent experimental and analytical studies that are concerned with heat removal from such a particulate bed are reviewed. Studies have indicated that bed dryout flux and, therefore, the heat removal rate from the particulate bed increases with the particle diameter (i.e., the permeability) for pool boiling conditions and can exceed the critical heat flux of a flat plate. Bed dryout in a large particle bed (i.e., a few millimetres) was found to be closely related to the ''flooding'' limit of the bed. Dryout under forced flow conditions was found to be affected by both forced and natural convection for mass flow rate smaller than m /SUB cr/ , whereas above this mass flow rate, bed dryout is proportional to the mass flow rate. Recent analyses were found to be in agreement with experimental data; however, additional research is needed to assess factors not accounted for in previous studies (e.g., effect of pressure, multidimensionality, stratification, etc.). Based on the expected pressure and particle sizes in a postulated severe accident sequence, a debris bed should be coolable, given a sufficient water supply

Multi-boiling Heat Transfer Analysis of a Convective Straight Fin with Temperature-Dependent Thermal Properties and Internal Heat Generation

Directory of Open Access Journals (Sweden)

Gbeminiyi Sobamowo

2017-10-01

Full Text Available In this study, by using the finite volume method, the heat transfer in a convective straight fin with temperature-dependent thermal properties and an internal heat generation under multi-boiling heat transfer modes are analyzed. In this regard, the local heat transfer coefficient is considered to vary within a power-law function of temperature. In the present study, the coexistence of all the boiling modes is taken into consideration. The developed heat transfer models and the corresponding numerical solutions are used to investigate the effects of various thermo-geometric parameters on the thermal performance of the longitudinal rectangular fin. The results shows that the fin temperature distribution, the total heat transfer, and the fin efficiency are significantly affected by the thermo-geometric parameters of the fin and the internal heat generation within the fin. The obtained results can provide a platform for improvements in the design of the fin in the heat transfer equipment.

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Potential ability of zeolite to generate high-temperature vapor using waste heat

Science.gov (United States)

Fukai, Jun; Wijayanta, Agung Tri

2018-02-01

In various material product industries, a large amount of high temperature steam as heat sources are produced from fossil fuel, then thermal energy retained by condensed water at lower than 100°C are wasted. Thermal energies retained by exhaust gases at lower than 200°C are also wasted. Effective utilization of waste heat is believed to be one of important issues to solve global problems of energy and environment. Zeolite/water adsorption systems are introduced to recover such low-temperature waste heats in this study. Firstly, an adsorption steam recovery system was developed to generate high temperature steam from unused hot waste heat. The system used a new principle that adsorption heat of zeolite/water contact was efficiently extracted. A bench-scaled system was constructed, demonstrating contentious generation of saturated steam nearly 150°C from hot water at 80°C. Energy conservation is expected by returning the generated steam to steam lines in the product processes. Secondly, it was demonstrated that superheated steam/vapor at higher than 200°C could be generated from those at nearly 120°C using a laboratory-scaled setup. The maximum temperature and the time variation of output temperature were successfully estimated using macroscopic heat balances. Lastly, the maximum temperatures were estimated whose saturate air at the relative humidity 20-80% were heated by the present system. Theoretically, air at higher than 200°C was generated from saturate air at higher than 70°C. Consequently, zeolite/water adsorption systems have potential ability to regenerate thermal energy of waste water and exhaust gases.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Heat generation and cooling of SSC magnets at high ramp rates

International Nuclear Information System (INIS)

Snitchler, G.; Capone, D.; Kovachev, V.; Schermer, R.

1992-01-01

This presentation will address a summary of AC loss calculations (SSCL), experimental results on cable samples (Westinghouse STC), short model magnets test results (FNAL, KEK-Japan), and recent full length magnets test data on AC losses and quench current ramp rate sensitivity (FNAL, BNL). Possible sources of the observed enhanced heat generation and quench sensitivity for some magnets will be discussed. A model for cooling conditions of magnet coils considering heat generation distribution and specific anisotropy of the heat transfer will be presented. The crossover contact resistance in cables and curing procedure influence on resistivity, currently under study, will be briefly discussed. (author)

Local entropy generation analysis of a rotary magnetic heat pump regenerator

International Nuclear Information System (INIS)

Drost, M.K.; White, M.D.

1990-01-01

The rotary magnetic heat pump has attractive thermodynamic performance but it is strongly influenced by the effectiveness of the regenerator. This paper uses local entropy generation analysis to evaluate the regenerator design and to suggest design improvements. The results show that performance of the proposed design is dominated by heat transfer related entropy generation. This suggests that enhancement concepts that improve heat transfer should be considered, even if the enhancement causes a significant increase in viscous losses (pressure drop). One enhancement technique, the use of flow disruptors, was evaluated and the results showed that flow disruptors can significantly reduce thermodynamic losses

Heat exchanger, particularly liquid sodium heated steam generator

International Nuclear Information System (INIS)

Robin, Marcel; Tillequin, Jean.

1977-01-01

This invention relates to a liquid sodium heated steam generator the characteristic of which is an annular distribution chamber fed by two independent and diametrically opposed manifolds on a common horizontal axis, issuing respectively into two adjacent compartments made in the chambers on both sides of a vertical transversal partition containing the axis of the casing and extending perpendicularly to the manifolds, each compartment being itself divided into a number of adjacent sectors marked by folded metal sheets fixed to the distributor and shaped so as to present in pairs and with the chamber opposite the manifold issuing into a compartment two independent ducts for distributing the sodium flow [fr

Creeping Viscous Flow around a Heat-Generating Solid Sphere

DEFF Research Database (Denmark)

Krenk, Steen

1981-01-01

The velocity field for creeping viscous flow around a solid sphere due to a spherically symmetric thermal field is determined and a simple thermal generalization of Stokes' formula is obtained. The velocity field due to an instantaneous heat source at the center of the sphere is obtained in closed...... form and an application to the storage of heat-generating nuclear waste is discussed....

Layout of an internally heated gas generator for the steam gasification of coal

International Nuclear Information System (INIS)

Feistel, P.P.; Duerrfeld, R.; Heck, K.H. van; Juentgen, H.

1975-01-01

Industrial-scale steam gasification of coal using heat from high temperature reactors requires research and development on allothermal gas generators. Bergbau-Forschung GmbH, Essen, does theoretical and experimental work in this field. The experiments deal with reaction kinetics, heat transfer and material tests. Their significance for the layout of a full-scale gas generator is shown. Including material specifications, the feasibility of a gasifier, characterized by a fluid bed volume of 318 m 3 and a heat transferring area of 4000 m 2 , results. The data, now available, are used to determine the gasification throughput from the heat balance, i.e. the equality of heat consumed and heat transferred. Throughputs of about 50 t/hr of coal are possible for a single gas generator, the helium outlet temperature of the HTR being 950 0 C/ Bergbau-Forschung has commissioned a medium-scale pilot plant (200 kg/hr). (Auth.)

Decay heat measurement of U-235

International Nuclear Information System (INIS)

Baumung, K.

1976-01-01

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

Radiation heat transfer within an open-cycle MHD generator channel

Science.gov (United States)

Delil, A. A. M.

1983-05-01

Radiation heat transfer in an MHD generator was modeled using the Sparrow and Cess model for radiation in an emitting, absorbing and scattering medium. The resulting general equations can be considerably reduced by introducing simplifying approximations for the channel and MHD gas properties. The simplifications lead to an engineering model, which is very useful for one-dimensional channel flow approximation. The model can estimate thermo-optical MHD gas properties, which can be substituted in the energy equation. The model considers the contribution of solid particles in the MHD gas to radiation heat transfer, considerable in coal-fired closed cycle MHD generators. The modeling is applicable also for other types of flow at elevated temperatures, where radiation heat transfer is an important quantity.

Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

International Nuclear Information System (INIS)

Qian Libo; Qiu Suizheng; Zhang Dalin; Su Guanghui; Tian Wenxi

2010-01-01

Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10 3 ∼ 10 7 ) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.

The development of a thermoelectric power generator dedicated to stove-fireplaces with heat accumulation systems

International Nuclear Information System (INIS)

Sornek, Krzysztof; Filipowicz, Mariusz; Rzepka, Kamila

2016-01-01

Highlights: • Application of thermoelectric generators in the stove-fireplace with accumulation. • Construction of the thermoelectric generator is limited by the heat accumulation. • Variants of the heat exchanger’s construction are discussed. • The control method is related on velocity of flue gas and water cooling. • The power limit of 30 W for self-sufficient operation is sufficient. - Abstract: A significant part of the world’s population (about 40%) cooks their meals and provides heating for their homes using wood-burning heating devices. Due to the relatively low cost of fuel and their aesthetic design, solid fuel stoves capable of heat accumulation are convenient and common. The use of dedicated small-scale power generators provides also additional benefits. This paper presents the results of a study conducted to verify the possibility of generating power using stove-fireplaces with heat accumulation systems. In such units, the temperature of the flue gas should be kept at a certain level for the purposes of storing heat, which results from certain limitations of the thermoelectric generators. To verify the possibility of applying thermoelectric modules in such heating devices, a dedicated system with thermoelectric generators was selected from among various microcogeneration systems and implemented. Three types of heat exchangers were studied and the most efficient unit was selected for further testing. Two types of generators, with maximum operating temperatures of 320 and 175 °C, were compared. Subsequently, the characteristics of the latter were determined. The conducted tests allowed to determine the performance and the total efficiency of the generators that were used. It has been demonstrated that the maximum power of the generator would not exceed ca. 30 W e and that there is no economic justification for such a device. However, providing a self-powered and self-sufficient operation of stove-fireplaces with heat accumulation systems