Industrial space heating and cooling from stored spent nuclear power plant fuel

International Nuclear Information System (INIS)

Shaver, B.O.; Doman, J.W.

1980-01-01

Projections by the Department of Energy indicate that some 5800 metric tons of spent fuel from nuclear power reactors are now in storage and that some 33000 metric tons are expected to be in storage in 1990. The bulk of the spent fuel is currently stored in water-filled basins at the reactor sites from which the material was discharged. The thermal energy in the fuel is dissipated to atmospheres via a pumped water-to-air heat exchanger system. This paper describes a feasibility study of potential methods for the use of the heat. Also, potential applications of heat recovery systems at larger AFR storage facilities were investigated

Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg

International Nuclear Information System (INIS)

2017-04-01

The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

Radiation heat transfer model in a spent fuel pool by TRACE code

International Nuclear Information System (INIS)

Sanchez-Saez, F.; Carlos, S.; Villanueva, J.F.; Martorell, S.

2014-01-01

Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. The work presented consists of the thermohydraulic simulation of spent fuel pool behavior after a loss of coolant throughout transfer channel with loss of cooling transient is produced. The simulation is done with the TRACE code. One of the most important variables that define the behavior of the pool is cladding temperature, which evolution depends on the heat emission. In this work convection and radiation heat transfer is considered. When both heat transfer models are considered, a clear delay in achieving the maximum peak cladding temperature (1477 K) is observed compared with the simulation in which only convection heat transfer is considered. (authors)

Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

Method of processing spent ion exchange resins

International Nuclear Information System (INIS)

Mori, Kazuhide; Tamada, Shin; Kikuchi, Makoto; Matsuda, Masami; Aoyama, Yoshiyuki.

1985-01-01

Purpose: To decrease the amount of radioactive spent ion exchange resins generated from nuclear power plants, etc and process them into stable inorganic compounds through heat decomposition. Method: Spent ion exchange resins are heat-decomposed in an inert atmosphere to selectively decompose only ion exchange groups in the preceeding step while high molecular skeltons are completely heat-decomposed in an oxidizing atmosphere in the succeeding step. In this way, gaseous sulfur oxides and nitrogen oxides are generated in the preceeding step, while gaseous carbon dioxide and hydrogen requiring no discharge gas procession are generated in the succeeding step. Accordingly, the amount of discharged gases requiring procession can significantly be reduced, as well as the residues can be converted into stable inorganic compounds. Further, if transition metals are ionically adsorbed as the catalyst to the ion exchange resins, the ion exchange groups are decomposed at 130 - 300 0 C, while the high molecular skeltons are thermally decomposed at 240 - 300 0 C. Thus, the temperature for the heat decomposition can be lowered to prevent the degradation of the reactor materials. (Kawakami, Y.)

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

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

International Nuclear Information System (INIS)

Kim, In Young; Lee, Un Chul

2011-01-01

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

Combined Thermal Management and Power Generation Concept for the Spent Fuel Dry Storage Cask

International Nuclear Information System (INIS)

Kim, In Guk; Bang, In Cheol

2017-01-01

The management of the spent nuclear fuel generated by nuclear power plants is a major issue in Korea due to insufficient capacity of the wet storage pools. Therefore, it is considered that dry storage system is the one possible solution for storing spent fuel. A dual-purpose metal cask (transportation and storage) is currently developing in Korea. This cask has 21 of fuel assemblies and 16.8 kW of maximum decay heat. To evaluate the critical safety in normal/off normal and accident conditions, critical stabilities were conducted by using CSAS 6.0. The experimental investigation of heat removal of a concrete storage cask was also conducted under normal, off normal and accident conditions. The results of the evaluation showed a good safety of the dry storage cask. The results showed the enhanced thermal performance according to modification of flow rate. To verify combined thermal management and power generation concept, a new type of test facility for dry storage cask was designed in 1/10 scale of concrete dry storage cask. The experimental study involved the cooling methods that are an integrated system on the top of the dry cask and air flow path on the canister wall. The results showed the temperature distribution of the wall and inside of the dry cask at the normal condition. The influence of the change of the heat load and cooling system were investigated. The heat removal by the integrated system is approximately 20% of the total heat removal of the dry cask with reduced wall temperature. In these tests, economic analysis is conducted by applying the concept of the cost and efficiency. Under different decay power cases, the energy efficiency of the heat pipe and Stirling engine are determined and compared based on experimental results. The average efficiencies of the Stirling engine were the range of 2.375 to 3.247% under the power range of 35– 65W. These results showed that advanced dry storage concept had a better cooling performance in comparison with

Allowable peak heat-up cladding temperature for spent fuel integrity during interim-dry storage

Directory of Open Access Journals (Sweden)

Ki-Nam Jang

2017-12-01

Full Text Available To investigate allowable peak cladding temperature and hoop stress for maintenance of cladding integrity during interim-dry storage and subsequent transport, zirconium alloy cladding tubes were hydrogen-charged to generate 250 ppm and 500 ppm hydrogen contents, simulating spent nuclear fuel degradation. The hydrogen-charged specimens were heated to four peak temperatures of 250°C, 300°C, 350°C, and 400°C, and then cooled to room temperature at cooling rates of 0.3 °C/min under three tensile hoop stresses of 80 MPa, 100 MPa, and 120 MPa. The cool-down specimens showed that high peak heat-up temperature led to lower hydrogen content and that larger tensile hoop stress generated larger radial hydride fraction and consequently lower plastic elongation. Based on these out-of-pile cladding tube test results only, it may be said that peak cladding temperature should be limited to a level < 250°C, regardless of the cladding hoop stress, to ensure cladding integrity during interim-dry storage and subsequent transport.

CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

International Nuclear Information System (INIS)

In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

2014-01-01

A dry storage system is used for the interim storage of spent fuel prior to permanent depository and/or recycling. The spent fuel is initially stored in a water pool for more than 5 years at least after dispatch from the reactor core and is transported to dry storage. The dry cask contains a multiple number of spent fuel assemblies, which are cooled down in the spent fuel pool. The dry cask is usually filled up with helium gas for increasing the heat transfer to the environment outside the cask. The dry storage system has been used for more than a decade in United States of America (USA) and the European Union (EU). Korea is also developing a dry storage system since its spent fuel pool is anticipated to be full within 10 years. The spent fuel will be stored in a dry cask for more than 40 years. The integrity and safety of spent fuel are important for long-term dry storage. The long-term storage will experience the degradation of spent fuel such as the embrittlement of fuel cladding, thermal creep and hydride reorientation. High burn-up fuel may expedite the material degradation. It is known that the cladding temperature has a strong influence on the material degradation. Hence, it is necessary to accurately predict the local distribution of the cladding temperature using the Computational Fluid Dynamics (CFD) approach. The objective of this study is to apply the CFD method for predicting the three-dimensional distribution of fuel temperature in a dry cask. This CFD study simulated the dry cask for containing the 21 fuel assemblies under development in Korea. This paper presents the fluid velocity and temperature distribution as well as the fuel temperature. A two-step CFD approach was applied to simulate the heat and fluid flow in a dry storage of 21 spent fuel assemblies. The first CFD analysis predicted the helium flow and temperature in a dry cask by a assuming porous body of the spent fuel. The second CFD analysis was to simulate a spent fuel assembly in the

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

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

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

Near-field heat transfer at the spent fuel test-climax: a comparison of measurements and calculations

International Nuclear Information System (INIS)

Patrick, W.C.; Montan, D.N.; Ballou, L.B.

1981-01-01

The Spent Fuel Test in the Climax granitic stock at the DOE Nevada Test Site is a test of the feasibility of storage and retrieval of spent nuclear reactor fuel in a deep geologic environment. Eleven spent fuel elements, together with six thermally identical electrical resistance heaters and 20 peripheral guard heaters, are emplaced 420 m below surface in a three-drift test array. This array was designed to simulate the near-field effects of thousands of canisters of nuclear waste and to evaluate the effects of heat alone, and heat plus ionizing radiation on the rock. Thermal calculations and measurements are conducted to determine thermal transport from the spent fuel and electrical resistance heaters. Calculations associated with the as-built Spent Fuel Test geometry and thermal source histories are presented and compared with thermocouple measurements made throughout the test array. Comparisons in space begin at the spent fuel canister and include the first few metres outside the test array. Comparisons in time begin at emplacement and progress through the first year of thermal loading in this multi-year test

Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

International Nuclear Information System (INIS)

Allen, G.K.

1980-08-01

Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects

Energy Technology Data Exchange (ETDEWEB)

Yavuz, U. [Turkish Atomic Energy Authority, Ankara (Turkey). Nuclear Safety Dept.; Zabunoolu, O.H. [Hacettepe Univ., Ankara (Turkey). Dept. of Nuclear Engineering

2006-08-15

Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33 000, 45 000, and 55 000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards. (orig.)

COBRA-SFS predictions of single assembly spent fuel heat transfer data

International Nuclear Information System (INIS)

Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.; Rector, D.R.

1986-04-01

The study reported here is one of several efforts to evaluate and qualify the COBRA-SFS computer code for use in spent fuel storage system thermal analysis. The ability of COBRA-SFS to predict the thermal response of two single assembly spent fuel heat transfer tests was investigated through comparisons of predictions with experimental test data. From these comparisons, conclusions regarding the computational treatment of the physical phenomena occurring within a storage system can be made. This objective was successfully accomplished as reasonable agreement between predictions and data were obtained for the 21 individual test cases of the two experiments

Thermal-hydraulic analysis of spent fuel storage systems

International Nuclear Information System (INIS)

Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

1987-01-01

This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

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

Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Bang, In Cheol

2016-01-01

Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

Non-linear thermal and structural analysis of a typical spent fuel silo

International Nuclear Information System (INIS)

Alvarez, L.M.; Mancini, G.R.; Spina, O.A.F.; Sala, G.; Paglia, F.

1993-01-01

A numerical method for the non-linear structural analysis of a typical reinforced concrete spent fuel silo under thermal loads is proposed. The numerical time integration was performed by means of a time explicit axisymmetric finite-difference numerical operator. An analysis was made of influences by heat, viscoelasticity and cracking upon the concrete behaviour between concrete pouring stage and the first period of the silo's normal operation. The following parameters were considered for the heat generation and transmission process: Heat generated during the concrete's hardening stage, Solar radiation effects, Natural convection, Spent-fuel heat generation. For the modelling of the reinforced concrete behaviour, use was made of a simplified formulation of: Visco-elastic effects, Thermal cracking, Steel reinforcement. A comparison between some experimental temperature characteristic values obtained from the numerical integration process and empirical data obtained from a 1:1 scaled prototype was also carried out. (author)

Numerical calculation and analysis of natural convection removal of the spent fuel residual heat of 10 MW high temperature gas cooled reactor

International Nuclear Information System (INIS)

Wang Jinhua; Huang Yifan; Wu Bin

2013-01-01

The spent fuel of 10 MW High Temperature Gas Cooled Reactor (HTR-10) could be stored in the shielded tank, and the tank is stored in the concrete shielded canister in spent fuel storage room, the residual heat of the spent fuel could be removed by the air. The ability of residual heat removal is analyzed in the paper, and the temperature field is numerically calculated through FEA program ANSYS, the analysis and the calculation are used to validate the safety of the spent fuel and the tank, the ultimate temperature of the spent fuel and the tank should below the safety limit. The calculation shows that the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank, and the temperature decreases gradually with radial distance, the temperature in the tank body is evenly distributed, and the temperature in the concrete shielded canister decreases gradually with radial distance. It is feasible to remove the residual heat of the spent fuel storage tank by natural ventilation, in natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method. (authors)

Thermomechanical modeling of the Spent Fuel Test-Climax

Energy Technology Data Exchange (ETDEWEB)

Butkovich, T.R.; Patrick, W.C.

1986-02-01

The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.

Thermomechanical modeling of the Spent Fuel Test-Climax

International Nuclear Information System (INIS)

Butkovich, T.R.; Patrick, W.C.

1986-02-01

The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C

Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

2015-10-15

The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

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

Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

International Nuclear Information System (INIS)

Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

1976-01-01

A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

Thermal model of spent fuel transport cask

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

1996-01-01

The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg; Bericht ueber die Entsorgung von radioaktiven Abfaellen und abgebrannten Brennelementen aus Baden-Wuerttemberg

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-04-15

The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

Spent grain as energy source for breweries

Energy Technology Data Exchange (ETDEWEB)

Scharf, P; Meyer-Pittroff, R [Technical Univ. of Munich, Freising (West Germany)

1990-01-01

About 1.7 million tons of spent grain are produced from brewing operations each year in the Federal Republic of Germany. Because there is more spent grain than can be sold as cattle fodder, other uses for this product are being sought. These include composting, methane generation, and direct combustion. Researchers are using anaerobic fermentation as a means of producing biogas and using waste energy from the brewery to heat the fermenter. Combustion of spent grain can give a net energy of 12.8 MJ per kg dry grain. With biogas production, about 15-20% of the whole primary energy or 50% of the primary energy used in the boiling house could be substituted by the biogas.

Container for spent fuel assembly

International Nuclear Information System (INIS)

Sawai, Takeshi.

1996-01-01

The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

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.

Thermal characteristics of spent activated carbon generated from air cleaning units in Korean nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Cho, Hang Rae; So, Ji Yang [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2017-06-15

To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

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

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

Thermal characteristics of spent activated carbon generated from air cleaning units in korean nuclear power plants

Directory of Open Access Journals (Sweden)

Ji-Yang So

2017-06-01

Full Text Available To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

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.

Neutron radiation characteristics of the IVth generation reactor spent fuel

Science.gov (United States)

Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

2018-03-01

Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

Examination of spent fuel radiation energy conversion for electricity generation

Energy Technology Data Exchange (ETDEWEB)

Lee, Haneol; Yim, Man-Sung, E-mail: msyim@kaist.ac.kr

2016-04-15

Highlights: • Utilizing conversion of radiation energy of spent fuel to electric energy. • MCNPX modeling and experiment were used to estimate energy conversion. • The converted energy may be useful for nuclear security applications. • The converted energy may be utilized for safety applications through energy storage. - Abstract: Supply of electricity inside nuclear power plant is one of the most important considerations for nuclear safety and security. In this study, generation of electric energy by converting radiation energy of spent nuclear fuel was investigated. Computational modeling work by using MCNPX 2.7.0 code along with experiment was performed to estimate the amount of electric energy generation. The calculation using the developed modeling work was validated through comparison with an integrated experiment. The amount of electric energy generation based on a conceptual design of an energy conversion module was estimated to be low. But the amount may be useful for nuclear security applications. An alternative way of utilizing the produced electric energy could be considered for nuclear safety application through energy storage. Further studies are needed to improve the efficiency of the proposed energy conversion concept and to examine the issue of radiation damage and economic feasibility.

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...

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

Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data

International Nuclear Information System (INIS)

Roddy, J.W.; Mailen, J.C.

1987-12-01

This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs

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

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.

Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

Energy Technology Data Exchange (ETDEWEB)

Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-11-01

The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

International Nuclear Information System (INIS)

De Wet, G. J.; Dent, C.

2008-01-01

A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

International Nuclear Information System (INIS)

Fukaya, Y.; Okubo, T.; Uchikawa, S.

2008-01-01

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

2008-07-15

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

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

International Nuclear Information System (INIS)

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

1986-02-01

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

Technique for recovering rare-earth metals from spent sintered Nd-Fe-B magnets without external heating

Directory of Open Access Journals (Sweden)

Ryo Sasai

2016-06-01

Full Text Available To selectively recover rare-earth metals with higher purity from spent sintered Nd-Fe-B magnets without external heating, we investigated the mechano-chemical treatment of spent sintered Nd-Fe-B magnet powder with a reaction solution of HCl and (COOH2 at room temperature. The results of various experiments showed that the mechano-chemical treatment with HCl and (COOH2 is very effective for recovering the rare-earth metals contained in spent sintered Nd-Fe-B magnet powder; the recovery rate and purity of the rare-earth metals were 95.3 and 95.0 mass%, respectively, under optimal conditions ([HCl] = 0.2 mol/dm3 and [(COOH2] = 0.25 mol/dm3.

A study on the expulsion of iodine from spent-fuel solutions

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1995-02-01

During dissolution of spent nuclear fuels, some radioiodine remains in spent-fuel solutions. Its expulsion to dissolver off-gas is important to minimize iodine escape to the environment. In our current work, the iodine remaining in spent-fuel solutions varied from 0 to 10% after dissolution of spent PWR-fuel specimens (approximately 3 g each). The amount remaining probably was dependent upon the dissolution time required. The cause is ascribable to the increased nitrous acid concentration that results from NOx generated during dissolution. The presence of nitrous acid was confirmed spectrophotometrically in an NO-HNO{sub 3} system at 100{degrees}C. Experiments examining NOx concentration versus the quantity of iodine in a simulated spent-fuel solution indicate that iodine (I{minus}) in spent fuels is subjected to the following three reactions: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid arising from NOx, and (3) formation of colloidal iodine (AgI, PdI{sub 2}), the major iodine species in a spent-fuel solution. Reaction (2) competes with reaction (3) to control the quantity of iodine remaining in solution. The following two-step expulsion process to remove iodine from a spent-fuel solution was derived from these experiments: Step One - Heat spent-fuel solutions without NOx sparging. When aged colloidal iodine is present, an excess amount of iodate should be added to the solution. Step Two - Sparge the fuel solution with NOx while heating. Effect of this new method was confirmed by use of a spent PWR-fuel solution.

Development of Experimental Facilities for Advanced Spent Fuel Management Technology

Energy Technology Data Exchange (ETDEWEB)

You, G. S.; Jung, W. M.; Ku, J. H. [and others

2004-07-01

The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

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.

Investigation of Electricity Generation by Using Gamma Radiation from Spent Fuels

International Nuclear Information System (INIS)

Lee, Haneol; Yim, Man-Sung Yim

2015-01-01

The purpose of this study is to investigate the electric power generation with scale spent fuel. OrigenArp has analyzed gamma radiation environment of spent fuel assembly, MCNPX has analyzed the scintillator behavior, and experimental work has analyzed the electric output of photovoltaic cell. Gamma radiation environment analysis result indicates gamma source rapidly decreases for the early storage period. Scintillator analysis result calculates the photon flux distribution which enters photovoltaic cell. Photovoltaic cell experiment calculates electric current, voltage current generation per each system unit. Generated electric power can be used to cope with existing safety system (i.e. storage monitoring system) under severe accident or to operate security system under external invasion situation (i.e. passive physical barrier system). Several researchers have shown that converting radiation energy into electric energy is possible. Karl Scharf studied the direct electric conversion of radiation energy by using photovoltaic cells. Researchers in University of Massachusetts Lowell have studied radiation-electric energy conversion by using gadolinium oxide scintillator and dye sensitized solar cell (DSSCs) and N. Horuichi et al. studied radiation-electric energy conversion by using inorganic scintillators and amorphous and crystal photovoltaic cells

Gamma irradiation and heat treatment for sterilization of Mushroom Spent Compost (MSC)

International Nuclear Information System (INIS)

Rosnani Abdul Rashid; Hassan Hamdani Mutaat; Foziah Ali; Zainab Harun; Hoe, P.C.K.; Mohd Meswan Maskom; Khairuddin Abdul Rahim

2016-01-01

Mushroom spent compost (MSC) refers to substrate residue from mushroom cultivation. MSC can be used as a material to improve soil properties for farming. MSC is also suitable as a carrier or substrate to hold beneficial microorganisms in bio fertilizer products. The carrier for bio fertilizer products must be sterilized completely before inoculating with bacteria inoculum. In this study, the effects of gamma irradiation and heat on sterilization of MSC were evaluated by microbial enumeration technique. The MSC was packed into polyethylene bags. The packed MSC were irradiated at different doses of 0 (control), 10, 20, 30, 40, 50 kGy, and heat sterilized at different autoclaving time of 0 (control), 15, 30, 45 and 60 minutes, separately. The irradiated and autoclaved samples were enumerated for viable colonies. The results showed that MSC was completely sterilized by gamma irradiation at the dose of 30 kGy whilst sterilization by heat required 30 minute autoclaving time. (author)

Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

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)

Estimation of the heat generation in vitrified waste product and shield thickness of the cask for the transportation of vitrified waste product using Monte Carlo technique

International Nuclear Information System (INIS)

Deepa, A.K.; Jakhete, A.P.; Mehta, D.; Kaushik, C.P.

2011-01-01

High Level Liquid waste (HLW) generated during reprocessing of spent fuel contains most of the radioactivity present in the spent fuel resulting in the need for isolation and surveillance for extended period of time. Major components in HLW are the corrosion products, fission products such as 137 Cs, 90 Sr, 106 Ru, 144 Ce, 125 Sb etc, actinides and various chemicals used during reprocessing of spent fuel. Fresh HLW having an activity concentration of around 100Ci/l is to be vitrified into borosilicate glass and packed in canisters which are placed in S.S overpacks for better confinement. These overpacks contain around 0.7 Million Curies of activity. Characterisation of activity in HLW and activity profile of radionuclides for various cooling periods sets the base for the study. For transporting the vitrified waste product (VWP), two most important parameters is the shield thickness of the transportation cask and the heat generation in the waste product. This paper describes the methodology used in the estimation of lead thickness for the transportation cask using the Monte Carlo Technique. Heat generation due to decay of fission products results in the increase in temperature of the vitrified waste product during interim storage and disposal. Glass being the material, not having very high thermal conductivity, temperature difference between the canister and surrounding bears significance in view of the possibility of temperature based devitrification of VWP. The heat generation in the canister and the overpack containing vitrified glass is also estimated using MCNP. (author)

Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

International Nuclear Information System (INIS)

St John, C.M.

1977-04-01

An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

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.

An assessment of temperature history on concrete silo dry storage system for CANDU spent fuel

International Nuclear Information System (INIS)

Lee, Dong-Gyu; Sung, Nak-Hoon; Park, Jea-Ho; Chung, Sung-Hwan

2016-01-01

Highlights: • We performed thermal analysis to predict the temperature distribution in the concrete silo. • Thermal analysis of the concrete silo was based on CFD code. • Temperature distribution and history for storage period was presented. • Thermal analysis results and test results agreed well. • The correlations can predict the maximum fuel temperature over storage period. - Abstract: Concrete silo is a dry storage system for spent fuel generated from CANDU reactors. The silo is designed to remove passively the decay heat from spent fuel, as well as to secure the integrity of spent fuel during storage period. Dominant heat transfer mechanisms must be characterized and validated for the thermal analysis model of the silo, and the temperature history along storage period could be determined by using the validated thermal analysis model. Heat transfer characteristics on the interior and exterior of fuel basket in the silo were assessed to determine the temperature history of silo, which is necessary for evaluating the long-term degradation behavior of CANDU spent fuel stored in the silo. Also a methodology to evaluate the temperature history during dry storage period was proposed in this study. A CFD model of fuel basket including fuel bundles was suggested and temperature difference correlation between fuel bundles and silo’s internal member, as a function of decay heat of fuel basket considering natural convection and radiation heat transfer, was deduced. Temperature difference between silo’s internal cavity and ambient air was determined by using a concept of thermal resistance, which was validated by CFD analysis. Fuel temperature was expressed as a function of ambient temperature and decay heat of fuel basket in the correlation, and fuel temperature along storage period was determined. Therefore, it could be used to assess the degradation behavior of spent fuel by applying the degradation mechanism expressed as a function of spent fuel

Criteria of reference radionuclides for safety analysis of spent fuel waste disposal

International Nuclear Information System (INIS)

Suryanto

1998-01-01

Study on the criteria for reference radionuclides selection for assessment on spent fuel disposal have done. The reference radionuclides in this study means radionuclides are predicted to contribute of the most radiological effect for man if spent fuel waste are discharged on deep geology formation. The research was done by investigate critically of parameters were used on evaluation a kind of radionuclide. Especially, this research study of parameter which relevant disposal case and or spent fuel waste on deep geology formation . The research assumed that spent fuel discharged on deep geology by depth 500-1000 meters from surface of the land. The migration scenario Radionuclides from waste form to man was assumed particularly for normal release in which Radionuclides discharge from waste form in a series thorough container, buffer, geological, rock, to fracture(fault) and move together with ground water go to biosphere and than go into human body. On this scenario, the parameter such as radionuclides inventory, half life, heat generation, hazard index based on maximum permissible concentration (MPC) or annual limit on intake (ALI) was developed as criteria of reference radionuclides selection. The research concluded that radionuclides inventory, half live, heat generated, hazard index base on MPC or ALI can be used as criteria for selection of reference Radionuclide. The research obtained that the main radionuclides are predicted give the most radiological effect to human are as Cs-137, Sr-90, I-129, Am-243, Cm-244, Pu-238, Pu-239, Pu-240. The radionuclides reasonable to be used as reference radionuclides in safety analysis at spent fuel disposal. (author)

Preliminary estimates of the impacts of alternative spent fuel acceptance rates

International Nuclear Information System (INIS)

Clark, L.L.; McKee, R.W.; Short, S.M.

1986-02-01

The rate at which spent fuel is accepted by the federal waste management system is an important interface between the private nuclear power sector and the federal government, which will assume responsibility for spent fuel disposal. An analysis of alternative rates based on minimum age criteria indicates substantial incentives to limit acceptance rates so as to result in minimum 10 to 15 year fuel ages for repository acceptance [1500 to 2500 metric tons of uranium (MTU) per year]. These incentives include lower heat generation rates and systems costs

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.

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

Directory of Open Access Journals (Sweden)

Algirdas Kaliatka

2013-01-01

Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

Neutron Generators for Spent Fuel Assay

International Nuclear Information System (INIS)

Ludewigt, Bernhard A.

2010-01-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

Examination of the Properties of a Spent Fuel based Electricity Generation System - Scintillator Performance Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Haneol; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

2016-05-15

Gammavoltaic was proposed by Karl Scharf in 1960. The low efficiency resulted in gammavoltaic being used as a radiation detector. In the 1990s the efficiency of gammavoltaic increased by the use of a scintillator. Gammavoltaic was further studied as a power source for spent fuel transportation and a nuclear battery in the 2000s Haneol Lee and Man-Sung Yim also suggested electricity generation system based on spent fuel stored inside the fuel pool of a nuclear power plant. This study proposed the systematic design of an electricity conversion system using CsI(Tl) scintillator and a-Si photovoltaic cell. As such, this study is selected to be a reference paper. The results of this paper indicate a self-absorption effect from the reference model. This effect is negligible while the irradiation degradation has to be considered. Two main ways to reduce radiation induced degradation are scintillator shielding and replacing scintillator material with a material having higher radiation resistance. The analysis of the scintillator used in the 'electricity generation system using gamma radiation from spent fuel' was performed to evaluate the ideal electricity generation in the reference research.

Examination of the Properties of a Spent Fuel based Electricity Generation System - Scintillator Performance Analysis

International Nuclear Information System (INIS)

Lee, Haneol; Yim, Man-Sung

2016-01-01

Gammavoltaic was proposed by Karl Scharf in 1960. The low efficiency resulted in gammavoltaic being used as a radiation detector. In the 1990s the efficiency of gammavoltaic increased by the use of a scintillator. Gammavoltaic was further studied as a power source for spent fuel transportation and a nuclear battery in the 2000s Haneol Lee and Man-Sung Yim also suggested electricity generation system based on spent fuel stored inside the fuel pool of a nuclear power plant. This study proposed the systematic design of an electricity conversion system using CsI(Tl) scintillator and a-Si photovoltaic cell. As such, this study is selected to be a reference paper. The results of this paper indicate a self-absorption effect from the reference model. This effect is negligible while the irradiation degradation has to be considered. Two main ways to reduce radiation induced degradation are scintillator shielding and replacing scintillator material with a material having higher radiation resistance. The analysis of the scintillator used in the 'electricity generation system using gamma radiation from spent fuel' was performed to evaluate the ideal electricity generation in the reference research

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)

Gravity-driven flow and heat transfer in a spent nuclear fuel storage pool

International Nuclear Information System (INIS)

Gay, R.R.

1983-01-01

The GFLOW code analyzes a three-dimensional rectangular porous medium by dividing the porous medium into a number of nodes or cells specified by the user. The finite difference form of the fluid conservation equations is solved for each node by application of a modified ''marker and cell'' numerical technique. The existence of spent nuclear fuel in any node is modeled by using a porosity value less than unity in that node and by including a surface heat transfer term in the fluid energy equation. In addition, local pressure losses due to grid spaces or other planar flow obstructions can be modeled by local loss coefficients. Heat conduction in the fuel is simulated by a fast running implicit finite difference model of the fuel, gap, and clad regions of the fuel rod

NAC's Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system: new generation multipurpose spent fuel storage for global application

International Nuclear Information System (INIS)

Pennington, C.W.

2004-01-01

Multipurpose canister systems (MCS) have been designed, licensed, fabricated, constructed, and loaded over the last decade within the U.S. These systems are characterized as concrete-based storage overpacks containing transportable canisters utilizing redundantly welded closures. Canisters are designed and intended to be transferred into transport packagings for shipment off-site, and canister designs do not preclude their use in waste disposal overpacks. NAC has learned a number of significant lessons in the deployment of its first generation MCS. During this period prior to the next procurement phase, NAC has developed a new generation MCS, incorporating the lessons learned from the first generation while considering the capabilities of the plants populating the next phase. The system is identified as the Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system, and this paper addresses its unique design, fabrication, and operations features. Among these are: a unique developed cell basket design, under patent review, that increases spent fuel capacities and simplifies fabrication while providing high strength and heat removal efficiency: a significantly enhanced canister closure design that improves welding time, personnel dose, and drying performance: a low profile vertical concrete cask design that improves on-site handling and site dose rates, offers tangible threat limitations for beyond-design-basis events, and maintains proven and simple construction/operation features: a simple, proven transfer system that facilitates transfer without excessive dose or handling: a new approach to water removal and canister drying, using a moisture entrainment, gas absorption vacuum (MEGAVAC) system. The paper includes design and licensing status of the MAGNASTOR system, and prototyping development that NAC has performed to date

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)

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.

Generation projection of solid and liquid radioactive wastes and spent radioactive sources in Mexico

International Nuclear Information System (INIS)

Garcia A, E.; Hernandez F, I. Y.; Fernandez R, E.; Monroy G, F.; Lizcano C, D.

2014-10-01

This work is focused to project the volumes of radioactive aqueous liquid wastes and spent radioactive sources that will be generated in our country in next 15 years, solids compaction and radioactive organic liquids in 10 years starting from the 2014; with the purpose of knowing the technological needs that will be required for their administration. The methodology involves six aspects to develop: the definition of general objectives, to specify the temporary horizon of projection, data collection, selection of the prospecting model and the model application. This approach was applied to the inventory of aqueous liquid wastes, as well as radioactive compaction organic and solids generated in Mexico by non energy applications from the 2001 to 2014, and of the year 1997 at 2014 for spent sources. The applied projection models were: Double exponential smoothing associating the tendency, Simple Smoothing and Lineal Regression. For this study was elected the first forecast model and its application suggests that: the volume of the compaction solid wastes, aqueous liquids and spent radioactive sources will increase respectively in 152%, 49.8% and 55.7%, while the radioactive organic liquid wastes will diminish in 13.15%. (Author)

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Directory of Open Access Journals (Sweden)

Ternovykh Mikhail

2017-01-01

Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

Science.gov (United States)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Economic Viability of Brewery Spent Grain as a Biofuel

Energy Technology Data Exchange (ETDEWEB)

Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-01-01

This report summarizes an investigation into the technical feasibility and economic viability of use grain wastes from the beer brewing process as fuel to generate the heat needed in subsequent brewing process. The study finds that while use of spent grain as a biofuel is technically feasible, the economics are not attractive. Economic viability is limited by the underuse of capital equipment. The investment in heating equipment requires a higher utilization that the client brewer currently anticipates. It may be possible in the future that changing factors may swing the decision to a more positive one.

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.

Post-test thermomechanical calulations and preliminary data analysis for the Spent Fuel Test: Climax

International Nuclear Information System (INIS)

Butkovich, T.R.; Patrick, W.C.

1985-09-01

The Spent Fuel Test - Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated, spent nuclear-reactor fuel assemblies. Thermomechanical response of the SFT-C was calculated before the test began using the finite-element structural analysis code ADINA and its companion heat transfer code ADINAT. While we found that the level of agreement between measured and calculated rock displacements was quite good, we needed to revise certain aspects of the heat transfer calculation, material properties, and in situ stresses to incorporate information obtained during and after the heated phase of the test. The post-test calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels that were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares those results with selected measurements made during the 3-year heating phase and 6-month cooling phase of the SFT-C

Spent fuel test. Climax data acquisition system integration report

International Nuclear Information System (INIS)

Nyholm, R.A.; Brough, W.G.; Rector, N.L.

1982-06-01

The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

Geomechanics of the Spent Fuel Test: Climax

International Nuclear Information System (INIS)

Wilder, D.G.; Yow, J.L. Jr.

1987-07-01

Three years of geomechanical measurements were made at the Spent Fuel Test-Climax (SFT-C) 1400 feet underground in fractured granitic rock. Heating of the rock mass resulted from emplacement of spent fuel as well as the heating by electrical heaters. Cooldown of the rock occurred after the spent fuel was removed and the heaters were turned off. The measurements program examines both gross and localized responses of the rock mass to thermal loading, to evaluate the thermomechanical response of sheared and fractured rock with that of relatively unfractured rock, to compare the magnitudes of displacements during mining with those induced by extensive heating of the rock mass, and to check assumptions regarding symmetry and damaged zones made in numerical modeling of the SFT-C. 28 refs., 113 figs., 10 tabs

A cooling concept of spent fuels in lag storage system

International Nuclear Information System (INIS)

Park, Jeong-Hwa; Yoo, Jae-Hyung; Park, Hyun-Soo

1991-01-01

A cooling concept of spent fuels by natural convection of hot cell air in storage pits was developed. Each storage pit was considered to be located below the hot cell floor and to accommodate only one spent fuel assembly. The aim of this study is to apply an appropriate cooling system to the design of a hot cell where considerable heat-generating fuels are handled. In such operations as disassembling, rod consolidation and packaging of spent fuels, a number of assemblies are on stand-by in the cell before and/or after the operations. A lag storage system can be used for temporary storage of spent fuels in nuclear facilities. Since the air in contact with bare fuel assemblies is potentially contaminated, it must be exhausted through high-efficiency particulate air (HEPA) filters. If the storage pit is completely isolated from the hot cell space, then it will require another separate ventilation system by forced convection of air, which will result in additional cost for the construction. In this work, however, a cooling system was proposed where natural convection of hot cell air itself is achieved by thermo-syphon. The cold air from the hot cell is supplied to the inlet provided at the bottom of each pit through the gap between the concrete pit wall and the interior thermal shield. This thermal shield is needed to form flow channels for cold and heated air, and to prevent the concrete from over-heating. The heated air exhausts from the outlet located at the top of cell wall. No additional HEPA filters are needed in this system because the heated air is routed back to the hot cell due to buoyancy-induced flow. The technical feasibility of this concept was validated by thermal analyses. As the key design constraints are the surface temperature of fuel cladding and the concrete temperature of the storage pit, the thermal analyses were focused on these parameters whether they follow within allowable limits or not. (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.

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)

Long-term storage of spent nuclear fuel

International Nuclear Information System (INIS)

Kempe, T.F.; Martin, A.; Thorne, M.C.

1980-06-01

This report presents the results of a study on the storage of spent nuclear fuel, with particular reference to the options which would be available for long-term storage. Two reference programmes of nuclear power generation in the UK are defined and these are used as a basis for the projection of arisings of spent fuel and the storage capacity which might be needed. The characteristics of spent fuel which are relevant to long-term storage include the dimensions, materials and physical construction of the elements, their radioactive inventory and the associated decay heating as a function of time after removal from the reactor. Information on the behaviour of spent fuel in storage ponds is reviewed with particular reference to the corrosion of the cladding. The review indicates that, for long-term storage, both Magnox and AGR fuel would need to be packaged because of the high rate of cladding corrosion and the resulting radiological problems. The position on PWR fuel is less certain. Experience of dry storage is less extensive but it appears that the rate of corrosion of cladding is much lower than in water. Unit costs are discussed. Consideration is given to the radiological impact of fuel storage. (author)

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

International Nuclear Information System (INIS)

Bergelson, B.; Gerasimov, A.; Tikhomirov, G.

2005-01-01

Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)

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.

Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

Energy Technology Data Exchange (ETDEWEB)

Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

Sensitivity analysis of characteristics of spent mixed oxide fuel

International Nuclear Information System (INIS)

Hagura, Naoto; Yoshida, Tadashi

2008-01-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)

Sensitivity analysis of characteristics of spent mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Hagura, Naoto; Yoshida, Tadashi [Musashi Institute of Technology, 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan)

2008-07-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 sigma basis. (authors)

Powerful electrostatic FEL: Regime of operation, recovery of the spent electron beam and high voltage generator

Energy Technology Data Exchange (ETDEWEB)

Boscolo, I. [Univ. and INFN, Milan (Italy); Gong, J. [Southwest Jiaotong Univ., Chengdu (China)

1995-02-01

FEL, driven by a Cockcroft-Walton electrostatic accelerator with the recovery of the spent electron beam, is proposed as powerful radiation source for plasma heating. The low gain and high gain regimes are compared in view of the recovery problem and the high gain regime is shown to be much more favourable. A new design of the onion Cockcroft-Walton is presented.

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.

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.)

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.)

NAC's Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system: new generation multipurpose spent fuel storage for global application

Energy Technology Data Exchange (ETDEWEB)

Pennington, C.W. [NAC International, Norcross, GA (United States)

2004-07-01

Multipurpose canister systems (MCS) have been designed, licensed, fabricated, constructed, and loaded over the last decade within the U.S. These systems are characterized as concrete-based storage overpacks containing transportable canisters utilizing redundantly welded closures. Canisters are designed and intended to be transferred into transport packagings for shipment off-site, and canister designs do not preclude their use in waste disposal overpacks. NAC has learned a number of significant lessons in the deployment of its first generation MCS. During this period prior to the next procurement phase, NAC has developed a new generation MCS, incorporating the lessons learned from the first generation while considering the capabilities of the plants populating the next phase. The system is identified as the Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system, and this paper addresses its unique design, fabrication, and operations features. Among these are: a unique developed cell basket design, under patent review, that increases spent fuel capacities and simplifies fabrication while providing high strength and heat removal efficiency: a significantly enhanced canister closure design that improves welding time, personnel dose, and drying performance: a low profile vertical concrete cask design that improves on-site handling and site dose rates, offers tangible threat limitations for beyond-design-basis events, and maintains proven and simple construction/operation features: a simple, proven transfer system that facilitates transfer without excessive dose or handling: a new approach to water removal and canister drying, using a moisture entrainment, gas absorption vacuum (MEGAVAC) system. The paper includes design and licensing status of the MAGNASTOR system, and prototyping development that NAC has performed to date.

High temperature dielectric properties of spent adsorbent with zinc sulfate by cavity perturbation technique

Energy Technology Data Exchange (ETDEWEB)

Lin, Guo [State Key Laboratory of Complex Nonferrous Metal Resources Clean Utilization, Kunming University of Science and Technology, Kunming, Yunnan 650093 (China); Faculty of Metallurgical and Energy Engineering, Kunming University of Science and Technology, Kunming, Yunnan 650093 (China); Key Laboratory of Unconventional Metallurgy, Ministry of Education, Kunming, Yunnan 650093 (China); National Local Joint Laboratory of Engineering Application of Microwave Energy and Equipment Technology, Kunming, Yunnan 650093 (China); Liu, Chenhui [State Key Laboratory of Complex Nonferrous Metal Resources Clean Utilization, Kunming University of Science and Technology, Kunming, Yunnan 650093 (China); Key Laboratory of Unconventional Metallurgy, Ministry of Education, Kunming, Yunnan 650093 (China); National Local Joint Laboratory of Engineering Application of Microwave Energy and Equipment Technology, Kunming, Yunnan 650093 (China); Faculty of Chemistry and Environment, Yunnan Minzu University, Kunming, Yunnan 650093 (China); Zhang, Libo, E-mail: libozhang77@163.com [State Key Laboratory of Complex Nonferrous Metal Resources Clean Utilization, Kunming University of Science and Technology, Kunming, Yunnan 650093 (China); Faculty of Metallurgical and Energy Engineering, Kunming University of Science and Technology, Kunming, Yunnan 650093 (China); Key Laboratory of Unconventional Metallurgy, Ministry of Education, Kunming, Yunnan 650093 (China); National Local Joint Laboratory of Engineering Application of Microwave Energy and Equipment Technology, Kunming, Yunnan 650093 (China); and others

2017-05-15

Highlights: • Cavity perturbation technique is employed to measure the dielectric properties. • Microwave absorption capability of ZnO is poor from 20 °C to 850 °C. • Dielectric properties of spent absorbent and zinc sulfate are influenced by temperature especially in high temperature stage. • Penetration depths and heating curve indicate spent adsorbent and ZnO·2ZnSO{sub 4}, ZnSO{sub 4} are excellent microwave absorber. • The pore structures of spent adsorbent are improved significantly by microwave-regeneration directly. - Abstract: Dielectric properties of spent adsorbent with zinc sulfate are investigated by cavity perturbation technique at 2450 MHz from 20 °C to approximately 1000 °C. Two weight loss stages are observed for spent adsorbent by thermogravimetric-differential scanning calorimeter (TG-DSC) analysis, and zinc sulfate is decomposed to ZnO·2ZnSO{sub 4} and ZnO at about 750 °C and 860 °C. Microwave absorption capability of ZnSO{sub 4} increases with increasing temperature and declines after ZnO generation on account of the poor dielectric properties. Dielectric properties of spent adsorbent are dependent on apparent density and noticed an interestingly linearly relationship at room temperature. The three parameters increase gently from 20 °C to 400 °C, but a sharp increase both in real part and imaginary part are found subsequently due to the volatiles release and regeneration of carbon. And material conductivity is improved, which contributes to the π-electron conduction appearance. Relationship between penetration depth and temperature further elaborate spent adsorbent is an excellent microwave absorber and the microwave absorption capability order of zinc compounds is ZnO·2ZnSO{sub 4}, ZnSO{sub 4} and ZnO. Heating characteristics suggest that heating rate is related with dielectric properties of materials. The pore structures of spent adsorbent are improved significantly and the surface is smoother after microwave-regeneration.

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)

Safety aspects of dry spent fuel storage and spent fuel management

International Nuclear Information System (INIS)

Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

2014-01-01

The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)

German Approach to Spent Fuel Management

International Nuclear Information System (INIS)

Jussofie, A.; Graf, R.; Filbert, W.

2010-01-01

in the pilot conditioning facility. The conditioning mode was taken over for the subsequent BSK3 concept which varies in the canister for the fuel rods and the emplacement technology. To prove the principle feasibility of this concept aboveground demonstration tests have been performed by 2009. Investigations on a further concept intended for the direct disposal of spent fuel assemblies without conditioning have just begun. Conceptual differences may have an impact on the safeguards approach. With the end of the moratorium the political situation will change and further steps related to the final disposal of heat-generating waste could be decided. (author)

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....

Spent fuel management

International Nuclear Information System (INIS)

2005-01-01

The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

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)

Generation of PHB from Spent Sulfite Liquor Using Halophilic Microorganisms.

Science.gov (United States)

Weissgram, Michaela; Gstöttner, Janina; Lorantfy, Bettina; Tenhaken, Raimund; Herwig, Christoph; Weber, Hedda K

2015-06-08

Halophilic microorganisms thrive at elevated concentrations of sodium chloride up to saturation and are capable of growing on a wide variety of carbon sources like various organic acids, hexose and also pentose sugars. Hence, the biotechnological application of these microorganisms can cover many aspects, such as the treatment of hypersaline waste streams of different origin. Due to the fact that the high osmotic pressure of hypersaline environments reduces the risk of contamination, the capacity for cost-effective non-sterile cultivation can make extreme halophilic microorganisms potentially valuable organisms for biotechnological applications. In this contribution, the stepwise use of screening approaches, employing design of experiment (DoE) on model media and subsequently using industrial waste as substrate have been implemented to investigate the applicability of halophiles to generate PHB from the industrial waste stream spent sulfite liquor (SSL). The production of PHB on model media as well as dilutions of industrial substrate in a complex medium has been screened for by fluorescence microscopy using Nile Blue staining. Screening was used to investigate the ability of halophilic microorganisms to withstand the inhibiting substances of the waste stream without negatively affecting PHB production. It could be shown that neither single inhibiting substances nor a mixture thereof inhibited growth in the investigated range, hence, leaving the question on the inhibiting mechanisms open. However, it could be demonstrated that some haloarchaea and halophilic bacteria are able to produce PHB when cultivated on 3.3% w/w dry matter spent sulfite liquor, whereas H. halophila was even able to thrive on 6.6% w/w dry matter spent sulfite liquor and still produce PHB.

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.

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

Hydrothermal Carbonization of Spent Osmotic Solution (SOS Generated from Osmotic Dehydration of Blueberries

Directory of Open Access Journals (Sweden)

Kaushlendra Singh

2014-09-01

Full Text Available Hydrothermal carbonization of spent osmotic solution (SOS, a waste generated from osmotic dehydration of fruits, has the potential of transformation into hydrochars, a value-added product, while reducing cost and overall greenhouse gas emissions associated with waste disposal. Osmotic solution (OS and spent osmotic solution (SOS generated from the osmotic dehydration of blueberries were compared for their thermo-chemical decomposition behavior and hydrothermal carbonization. OS and SOS samples were characterized for total solids, elemental composition, and thermo-gravimetric analysis (TGA. In addition, hydrothermal carbonization was performed at 250 °C and for 30 min to produce hydrochars. The hydrochars were characterized for elemental composition, Brunauer-Emmett-Teller (BET surface area, particle shape and surface morphology. TGA results show that the SOS sample loses more weight in the lower temperature range than the OS sample. Both samples produced, approximately, 40%–42% (wet-feed basis hydrochar during hydrothermal carbonization but with different properties. The OS sample produced hydrochar, which had spherical particles of 1.79 ± 1.30 μm diameter with a very smooth surface. In contrast, the SOS sample produced hydrochar with no definite particle shape but with a raspberry-like surface.

Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

International Nuclear Information System (INIS)

Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

2002-01-01

Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

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.

Second Generation Ethanol Production from Brewers’ Spent Grain

Directory of Open Access Journals (Sweden)

Rossana Liguori

2015-03-01

Full Text Available Ethanol production from lignocellulosic biomasses raises a global interest because it represents a good alternative to petroleum-derived energies and reduces the food versus fuel conflict generated by first generation ethanol. In this study, alkaline-acid pretreated brewers’ spent grain (BSG was evaluated for ethanol production after enzymatic hydrolysis with commercial enzymes. The obtained hydrolysate containing a glucose concentration of 75 g/L was adopted, after dilution up to 50 g/L, for fermentation by the strain Saccharomyces cerevisiae NRRL YB 2293 selected as the best producer among five ethanologenic microorganims. When the hydrolysate was supplemented with yeast extract, 12.79 g/L of ethanol, corresponding to 0.28 g of ethanol per grams of glucose consumed (55% efficiency, was obtained within 24 h, while in the non-supplemented hydrolysate, a similar concentration was reached within 48 h. The volumetric productivity increased from 0.25 g/L·h in the un-supplemented hydrolysate to 0.53 g/L h in the yeast extract supplemented hydrolysate. In conclusion, the strain S. cerevisiae NRRL YB 2293 was shown able to produce ethanol from BSG. Although an equal amount of ethanol was reached in both BSG hydrolysate media, the nitrogen source supplementation reduced the ethanol fermentation time and promoted glucose uptake and cell growth.

Source Term Characteristics Analysis for Structural Components in PWR spent fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Kook, Dong Hak; Choi, Heui Joo; Cho, Dong Keun [KAERI, Daejeon (Korea, Republic of)

2010-12-15

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core under different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.32x1015 Bequerels, 238 Watts, 4.32x109 m3 water, respectively, at 10 years after discharge. Those values correspond to 0.6 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 25{approx}50 % and 35{approx}40 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important

Investigations of physicochemical properties of dusts generated in mechanical reclamation process of spent moulding sands with alkaline resins

Directory of Open Access Journals (Sweden)

R. Dańko

2014-03-01

Full Text Available Mechanical reclamation processes of spent moulding sands generate large amounts of post-reclamation dusts mainly containing rubbed spent binding agents and quartz dusts. The amount of post-reclamation dusts, depending in the reclamation system efficiency and the reclaim dedusting system, can reach 5%-10% in relation to the total reclaimed spent moulding sand. The proper utilization of such material is a big problem facing foundries these days. This study presents the results of investigations of physicochemical properties of post- reclamation dusts. All tested dusts originated from various Polish cast steel plants applying the mechanical reclamation process of moulding sands with alkaline resins, obtained from different producers. Different dusts, delivered from foundries, were tested to determine their chemical composition, granular characterization, physicochemical and energetic properties. Presented results confirmed assumptions that it is possible to utilize dusts generated during mechanical reclamation of used sands with organic resins as a source of energy.

Numerically predicting horizontally oriented spent fuel rod surface temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1992-01-01

A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degree C and 6 degree C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

Spent nuclear fuel assembly storage vessel

International Nuclear Information System (INIS)

Yagishita, Takuya

1998-01-01

The vessel of the present invention promotes an effect of removing after heat of spent nuclear fuel assemblies so as not to give force to the storage vessel caused by expansion of heat removing partitioning plates. Namely, the vessel of the present invention comprises a cylinder body having closed upper and lower portions and a plurality of heat removing partitioning cylinders disposed each at a predetermined interval in the circumferential direction of the above-mentioned cylinder body. The heat removing partitioning cylinders comprises (1) first heat removing partitioning plates extended in the radial direction of the cylinder body and opposed at a predetermined gap in the circumferential direction of the cylinder body, and having the base ends on the side of the inner wall of the cylinder body being secured to the inner wall of the cylinder body and (2) a second heat removing plate for connecting the top ends of both opposed heat removing partitioning plates on the central side of the cylinder body with each other. Spent nuclear fuel assemblies are contained in a plurality of closed spaces surrounded by the first heat removing partitioning plates and the second heat removing partitioning plate. With such constitution, since after heat is partially transferred from the heat removing partitioning plates to the cylindrical body directly by heat conduction, the heat removing effect can be promoted compared with the prior art. (I.S.)

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.

Physical and chemical changes to rock near electrically heated boreholes at Spent Fuel Test-Climax

International Nuclear Information System (INIS)

Beiriger, J.M.; Durham, W.B.; Ryerson, F.J.

1985-01-01

Sections of Climax Stock quartz monzonite taken from the vicinity of two electrically heated boreholes at Spent Fuel Test-Climax (SFT-C) have been studied by scanning electron microscopy and optical microscopy for signs of changes in crack structure and in mineralogy resulting from operations at SFT-C. The crack structure, as measured by density of cracks and average crack lengths was found not to have changed as a result of heating, regardless of distance from the heater hole. However, rock near the heater borehole sampled in the north heater drift was found to be more cracked than rock near the borehole sampled in the south heater drift. Mineralogically, the post-test samples are identical to the pre-test samples. No new phases have been formed as a result of the test. 10 refs., 6 figs., 8 tabs

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

Numerically predicting horizontally oriented spent fuel rod surface temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1993-01-01

A comparison between numerical calculations with use of commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degrees C and 23 degrees C for the low heat dissipation and high dissipation, respectively. The temperature predictions using helium as a fill gas are lower than the experimental data for the low and medium heat dissipation levels. The temperature predictions are 1 degrees C and 6 degrees C lower than the experimental data for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degrees C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include a experimental uncertainity in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This works demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects, such as axial heat transfer through the spent fuel rods, will be increasingly important as the amount of dissipated heat increases

Development of Dynamic Spent Nuclear Fuel Environmental Effect Analysis Model

International Nuclear Information System (INIS)

Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

2010-07-01

The dynamic environmental effect evaluation model for spent nuclear fuel has been developed and incorporated into the system dynamic DANESS code. First, the spent nuclear fuel isotope decay model was modeled. Then, the environmental effects were modeled through short-term decay heat model, short-term radioactivity model, and long-term heat load model. By using the developed model, the Korean once-through nuclear fuel cycles was analyzed. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. If the disposal starts from 2060, the short-term decay heat of Cs-137 and Sr-90 isotopes are W and 1.8x10 6 W in 2100. Also, the total long-term heat load in 2100 will be 4415 MW-y. From the calculation results, it was found that the developed model is very convenient and simple for evaluation of the environmental effect of the spent nuclear fuel

Risk analysis methodology for unreprocessed spent fuel disposal in bedded salt

International Nuclear Information System (INIS)

Pepping, R.E.; Chu, M.S.Y.; Cranwell, R.M.

1982-01-01

In accordance with the decision to defer the reprocessing of commercially generated spent fuel, we are investigating the implications on risk of direct disposal of spent fuel assemblies. To the extent possible, we are using the methodology developed at Sandia for the NRC to evaluate risks from the disposal of wastes from reprocessing of spent fuel. This allows direct comparison of the risks calculated for the two waste forms. A number of differences between the two waste forms with implications on risk have been identified and investigation of their effects has begun. Among these are the presence of gases and additional plutonium and uranium isotopes, the potential for differing leach behavior, and the difference in the decay heat source which determines the overall thermomechanical response of the host media. We have analyzed a number of scenarios for a hypothetical geologic repository that have been identified as important contributors to risk from the disposal of both reprocessed and unreprocessed spent fuel. For each scenario, we employ the Groundwater Transport, Pathways to Man, and Dosimetry and Health Effects models of the High Level Waste Methodology. Risks are compared for the reprocessed and unreprocessed spent fuel wastes and the effects of uncertainty in the parameters of the various models are compared

Cementation of secondary wastes generated from carbonisation of spent organic ion exchange resins from nuclear power plants

International Nuclear Information System (INIS)

Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

2004-07-01

The spent IX resins containing radioactive fission and activation products from power reactors are highly active solid wastes generated during operations of nuclear reactors. Process for carbonization of IX resins to achieve weight and volume reduction has been optimized on 50 dm 3 /batch pilot test rig. The process generates carbonaceous residue, organic liquid condensates (predominantly styrene) and aqueous alkaline scrubber solutions as secondary wastes. The report discusses laboratory tests on leaching of 137 Cs from cement matrix incorporating carbonaceous residues and extrapolation of results to 200 liter matrix block. The cumulative fraction of 137 Cs leached from 200 liter cement matrix was estimated to be 0.0021 in 200 days and 0.0418 over a period of 30 years. Incorporation of organic liquid condensates into cement matrix has been tried out successfully. Thus two types of secondary wastes generated during carbonization of spent IX resins can be immobilized in cement matrix. (author)

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.)

Spent 99Mo/99mTc generator as an economical source of 99Mo

International Nuclear Information System (INIS)

El-Kolaly, M.T.

1990-01-01

An improved method for utilization and purification of 99 Mo from spent 90 Mo/ 99m Tc generators has been described. After washing the generator with saline to remove the generated 99m Tc, followed by 2 mL 5 M NaOH containing a few drops of H 2 O 2 , the 99 Mo was quantitatively eluted from the generator with 5 mL 5 M NaOH. The alkaline eluate containing 99 Mo was contaminated with partially dissolved alumina. In the present method, an anion-exchange resin Dowex 1 x 8 column was used for purification of 99 Mo from the contaminating alumina. The resultant 99 Mo was of high purity and contained 3+ /mL 99 Mo solution, as estimated by atomic absorption. (author)

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.

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.

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.

Spent fuel reprocessing options

International Nuclear Information System (INIS)

2008-08-01

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

Some notes on the Timing of Geological Disposal of CANDU Spent Fuels

International Nuclear Information System (INIS)

Choi, Heui Joo; Kook, Dong Hak; Choi, Jong Won

2010-01-01

CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.

Development of the spent fuel rod cutting device using the blade cutters

International Nuclear Information System (INIS)

Jung, Jae Hoo; Yoon, Ji Sup; Hong, Dong Hee; Kim, Young Hwan; Park, Gee Yong; Kim, Do Woo

2000-11-01

A spent fuel rod cutting device is to cut a spent nuclear fuel rod to optimal size for consequent decladding operation. In this paper, various properties of fuel rod, such as a dimension and material of zircaloy tubes and fuel pellets, are investigated. Also, commercially available cutting method and tools is investigated in terms of its performance. As a result, the blade cutter is selected for the design. In order to fabricate the durable blade cutter, various materials are analyzed in terms of material properties, cutter shape, and heat treatment method, etc. Also, the durability of this tool is tested by cutting the SUS tubes and zircaloy tubes. In the device design, the remote maintainability is considered so that the modularized design is accomplished. Also, the other factors considered in the design are the round shape sustainability at the cut surface, the amount of debris generation, and the fire risk, etc. Considering these design consideration, the spent fuel rod cutting device is fabricated and tested

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.

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.

Novel approach to recover cobalt and lithium from spent lithium-ion battery using oxalic acid.

Science.gov (United States)

Zeng, Xianlai; Li, Jinhui; Shen, Bingyu

2015-09-15

With the booming of consumer electronics (CE) and electric vehicle (EV), a large number of spent lithium-ion battery (LIBs) have been generated worldwide. Resource depletion and environmental concern driven from the sustainable industry of CE and EV have motivated spent LIBs should be recovered urgently. However, the conventional process combined with leaching, precipitating, and filtering was quite complicated to recover cobalt and lithium from spent LIBs. In this work, we developed a novel recovery process, only combined with oxalic acid leaching and filtering. When the optimal parameters for leaching process is controlled at 150 min retention time, 95 °C heating temperature, 15 g L(-1) solid-liquid ratio, and 400 rpm rotation rate, the recovery rate of lithium and cobalt from spent LIBs can reach about 98% and 97%, respectively. Additionally, we also tentatively discovered the leaching mechanism of lithium cobalt oxide (LiCoO2) using oxalic acid, and the leaching order of the sampling LiCoO2 of spent LIBs. All the obtained results can contribute to a short-cut and high-efficiency process of spent LIBs recycling toward a sound closed-loop cycle. Copyright © 2015 Elsevier B.V. All rights reserved.

Source Term Characterization for Structural Components in 17 x 17 KOFA Spent Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Cho, Dong Keun; Kook, Dong Hak; Choi, Heui Joo; Choi, Jong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-12-15

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be 1.40 x 10{sup 15} Bequerels, 236 Watts, 4.34 x 10{sup 9} m{sup 3}-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20 {approx} 45 % and 30 {approx} 45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

Possibility of multiple temperature maxima in geologic repositories for spent fuel from nuclear reactors

International Nuclear Information System (INIS)

Beyerlein, S.W.; Claiborne, H.C.

1980-01-01

Heat transfer studies show that two temperature maxima at the disposal horizon could be experienced in CANDU spent fuel repositories - one at about 60 years and another slightly higher one at 13,000 years. Because CANDU spent fuels display a monotonically decreasing heat generation rate, it is not immediately obvious why this behavior should occur. This report investigates this behavior, confirms the Canadian results, demonstrates that the double peak phenomenon is due to the presence of the right mixture of short- and long-lived nuclides in the fuel, and concludes that the 13,000-year maximum is largely an artifact of the infinite or very large plane source model. When more realistic repository geometries are used, the second peak disappears for repository sizes less than about 1 km 2 . Over the long term, radial and surface heat transfer causes the thermal history of the disposal region to deviate from that predicted by infinite plane (or large finite) source models by reducing the magnitude of the second peak. Beyond a 1000-year time horizon, care should be exercised in modeling spent fuel repositories to include the proper boundary conditions. For the first few centuries after emplacement, however, the infinite source model is consistent with the finite disk source model as well as with arrays of spherical and point sources. The second temperature peak can be avoided by restricting the size of the repository and/or partitioning out the long-lived components of the fuel. When spent fuel from PWRs was examined for multiple temperature maxima, only one peak was found, even for the infinite plane source model

Disposal of spent fuel

International Nuclear Information System (INIS)

Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

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

Castor-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

CASTOR-V/21 PWR spent fuel storage cask performance test

International Nuclear Information System (INIS)

Creer, J.M.; Schoonen, D.H.

1986-01-01

Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

Thermal analysis for a spent reactor fuel storage test in granite

International Nuclear Information System (INIS)

Montan, D.N.

1980-09-01

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy's Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m 3 /s and an ambient rock temperature of 23 0 C, the maximum calculated rock temperature (near the center of the heat source) is about 100 0 C while the maximum air temperature in the drift is around 40 0 C. This ventilation (1 m 3 /s through the main drift and 1/2 m 3 /s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage

Electrochemical generation of fentons reagent to treat spent caustic wastewater

International Nuclear Information System (INIS)

Hansen, H. K.; Nunez, P.; Rodriguez, N.; Guzman, J.

2009-01-01

An important wastewater stream from oil refineries is the spent caustic. Caustic solutions are used as scrubbing agent during the desulphurization process to eliminate sulphur an mercaptans from oil and gasses. Spent caustic is classified as DOO3 (reactive sulphide) hazardous waste under the US Resource Conservation and Recovery Act (RCRA). (Author)

Spent fuel and fuel pool component integrity. Annual report, FY 1979

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

1980-05-01

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

Verification of heat removal capability of a concrete cask system for spent fuel storage

International Nuclear Information System (INIS)

Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatugu

2001-01-01

The reprocessing works comprising of a center of nuclear fuel cycle in Japan is now under construction at Rokkasho-mura in Aomori prefecture, which is to be operated in 2005. However, as reprocessing capacity of the works is under total forming amount of spent nuclear fuels, it has been essential to construct a new facility intermediately to store them at a period before reprocessing them because of prediction to reach limit of pool storage in nuclear power stations. There are some intermediate storage methods, which are water pool method for wet storage, and bolt method, metal cask method, silo method and concrete cask method for dry storage. Among many methods, the dry storage is focussed at a standpoint of its operability and economy, the concrete cask method which has a lot of using results in U.S.A. has been focussed as a method expectable in its cost reduction effect among it. The Ishikawajima-Harima Heavy Industries Co., Ltd. produced, in trial, a concrete cask with real size to confirm productivity when advancing design work on concrete cask. By using the trial product, a heat removal test mainly focussing temperature of concrete in the cask was carried out to confirm heat conductive performances of the cask. And, analysis of heat conductivity was also carried out to verify validity of its analysis model. (G.K.)

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

2014-02-01

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

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

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.

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

Science.gov (United States)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

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...

Kinetics and thermodynamics of aqueous Cu(II adsorption on heat regenerated spent bleaching earth

Directory of Open Access Journals (Sweden)

Enos W. Wambu

2011-08-01

Full Text Available This study investigated the kinetics and thermodynamics of copper(II removal from aqueous solutions using spent bleaching earth (SBE. The spent bleaching earth, a waste material from edible oil processing industries, was reactivated by heat treatment at 370 oC after residual oil extraction in excess methyl-ethyl ketone. Copper adsorption tests were carried out at room temperature (22±3 oC using 5.4 x 10-3C M metal concentrations. More than 70% metal removal was recorded in the first four hours although adsorption continued to rise to within 90% at 42 hours. The pH, adsorbent dosage and initial concentrations were master variables affecting RSBE adsorption of Cu(II ions. The adsorption equilibrium was adequately described by the Dubinin-Radushkevich (D-R and the Temkin isotherms and the maximum sorption capacity derived from the D-R isotherm was compared with those of some other low cost adsorbents. The adsorption process was found to follow Lagergren Pseudo-second order kinetics complimented by intra-particle diffusion kinetics at prolonged periods of equilibration. Based on the D-R isotherm adsorption energy and the thermodynamic adsorption free energy ∆G, it was suggested that the process is spontaneous and based on electrostatic interactions between the metal ions and exposed active sites in the adsorbent surface.

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.

Spent fuel storage at the Rancho Seco Nuclear Generation Station

International Nuclear Information System (INIS)

Miller, K.R.; Field, J.J.

1995-01-01

The Sacramento Municipal Utility District (SMUD) has developed a strategy for the storage and transport of spent nuclear fuel and is now in the process of licensing and manufacturing a Transportable Storage System (TSS). Staff has also engaged in impact limiter testing, non-fuel bearing component reinsertion, storage and disposal of GTCC waste, and site specific upgrades in support of spent fuel dry storage

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 Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

Energy Technology Data Exchange (ETDEWEB)

Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

Spent nuclear fuel storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2005-01-01

When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

Analysis of scenarios for the direct disposal of spent nuclear fuel disposal conditions as expected in Germany

International Nuclear Information System (INIS)

Ashton, P.; Mehling, O.; Mohn, R.; Wingender, H.J.

1990-01-01

This report contains an investigation of aspects of the waste management of spent light water reactor fuel by direct disposal in a deep geological formation on land. The areas covered are: interim dry storage of spent fuel with three options of pre-conditioning; conditioning of spent fuel for final disposal in a salt dome repository; disposal of spent fuel (heat-generating waste) in a salt dome repository; disposal of medium and low-level radioactive wastes in the Konrad mine. Dose commitments, effluent discharges and potential incidents were not found to vary significantly for the various conditioning options/salt dome repository types. Due to uncertainty in the cost estimates, in particular the disposal cost estimates, the variation between the three conditioning options examined is not considered as being significant. The specific total costs for the direct disposal strategy are estimated to lie in the range ECU 600 to 700 per kg hm (basis 1988)

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

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.

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.

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

Criticality evaluation of long term for spent fuel, using Scale

International Nuclear Information System (INIS)

Esquivel E, J.; Vargas E, S.; Ramirez S, J. R.

2013-10-01

Once carried out the spent fuel discharge, of the reactor core, this continues generating decay heat and diverse fission products, reason why is important to store this fuel inside containers able to dissipate the heat generated by the isotopes decay of the fuel and to maintain the fuels arrangement in subcritical condition. This means that: is necessary to assure the sub-criticality of those fuel assemblies in the time. This work, presents a criticality evaluation of fuel assemblies type PWR in a storage generic container. For this purpose have been used two codes: GeeWiz, to carry out the geometric model of the container with the fuel assemblies, and Keno, with which, the criticality of the full container with fuel is determined until a 10 6 years period. These codes are part of the package Scale. The specifications for each one of the analyzed components are based on a Benchmark document of the Nea/OECD, of where, the results that reports are compared with the obtained results by the realized analysis. (Author)

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

Spent fuel pool spray cooling system for the AP1000 {sup registered}

Energy Technology Data Exchange (ETDEWEB)

Vujic, Zoran; Sassen, Felix; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2013-07-01

The AP1000 {sup registered} plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for Design Basis Events and Beyond Design Basis Accidents (BDBA). The AP1000 {sup registered} plant lines of defense with respect to Spent Fuel Pool (SFP) cooling are as follows: 1. During normal and abnormal conditions, defense-in-depth and duty systems provide highly reliable SFP cooling, supplied by offsite AC power or the onsite Standby Diesel Generators. 2. For unlikely events with extended loss of AC power (i.e. station black-out) and/or loss of heat sink, spent fuel cooling can be still provided indefinitely by: 2a. Passive systems, requiring minimal or no operator actions, sufficient for at least 72 hours under all possible loading conditions. 2b. After 3 days, several different means are provided to continue SFP cooling using installed plant equipment as well as off-site equipment with built-in connections. 3. Even for BDBA with postulated SFP damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 {sup registered} SFP Spray System provides an additional line of defense to prevent spent fuel damage. (orig.)

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)

Safe transport of spent fuels after long-term storage

International Nuclear Information System (INIS)

Aritomi, M.; Takeda, T.; Ozaki, S.

2004-01-01

Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly

Storing the world's spent nuclear fuel

International Nuclear Information System (INIS)

Barkenbus, J.N.; Weinberg, A.M.; Alonso, M.

1985-01-01

Given the world's prodigious future energy requirements and the inevitable depletion of oil and gas, it would be foolhardy consciously to seek limitations on the growth of nuclear power. Indeed, the authors continue to believe that the global nuclear power enterprise, as measured by installed reactor capacity, can become much larger in the future without increasing proliferation risks. To accomplish this objective will require renewed dedication to the non-proliferation regime, and it will require some new initiatives. Foremost among these would be the establishment of a spent fuel take-back service, in which one or a few states would retrieve spent nuclear fuel from nations generating it. The centralized retrieval of spent fuel would remove accessible plutonium from the control of national leaders in non-nuclear-weapons states, thereby eliminating the temptation to use this material for weapons. The Soviets already implement a retrieval policy with the spent fuel generated by East European allies. The authors believe that it is time for the US to reopen the issue of spent-fuel retrieval, and thus to strengthen its non-proliferation policies and the nonproliferation regime in general. 7 references

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...

Shipment of gas generating spent fuel on the U.S. Department of Energy Hanford Site

International Nuclear Information System (INIS)

Edwards, W.S.

1998-01-01

, to the onsite movement of commercial spent fuel from a reactor pool to a dry storage facility. In the case of the K Basin spent gas generating played a significant role in the safety analysis

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

Strategic environmental assessment of the national programme for the safe management of spent fuel and radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Steinhoff, Mathias; Kallenbach-Herbert, Beate; Claus, Manuel [Oeko-Institut e.V. Darmstadt (Germany); and others

2015-03-27

The report on the strategic environmental audit for the national waste disposal program covers the following issues: aim of the study, active factors, environmental objectives; description and evaluation of environmental impact including site selection criteria for final repositories of heat generating radioactive waste, intermediate storage of spent fuel elements and waste from reprocessing plants, disposal of wastes retrieved from Asse II; hypothetical zero variants.

Strategic environmental assessment of the national programme for the safe management of spent fuel and radioactive waste

International Nuclear Information System (INIS)

Steinhoff, Mathias; Kallenbach-Herbert, Beate; Claus, Manuel

2015-01-01

The report on the strategic environmental audit for the national waste disposal program covers the following issues: aim of the study, active factors, environmental objectives; description and evaluation of environmental impact including site selection criteria for final repositories of heat generating radioactive waste, intermediate storage of spent fuel elements and waste from reprocessing plants, disposal of wastes retrieved from Asse II; hypothetical zero variants.

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.

Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Hennings, U.

1987-01-01

Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

Spent fuel management in Japan

International Nuclear Information System (INIS)

Mineo, H.; Nomura, Y.; Sakamoto, K.

1998-01-01

In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

Temperature and neutron dose rate measurements at a spent fuel shipping cask

International Nuclear Information System (INIS)

Krause, F.

1982-01-01

Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

A safety study on the wet storage of spent fuel

International Nuclear Information System (INIS)

Chun, Kwan Sik; Whang, Joo Ho; Lee, Hoo Kun; Choi, Jong Won; Lee, Jong Geun

1989-02-01

This study is to provide data related with a basic design of the spent fuel storage facility in the field of radiation and to establish the safety assessment methodology of away from reactor spent fuel storage facility. This is in progress and continue upto the year of 1991. The mathematical model which predict the quantity of environmental release of fission and corrosion products from spent fuel received and stored in wet storage facility operated in normal conditions was prepared. The decay characteristic of domestic spent fuels are analysed and then the coefficients for the prediction of the decay heat by simple formular was determined. This correlations could predict decay heat of spent fuel with ±10% difference from ORIGEN2 results. The release factor of cobalt out of PWR spent fuel in PIE pool is 7.97 x 10-12∼8.49 x 10-11 Ci/ sec-rod, which appears to be linear without being connected with the types of fuel defects, but that of cesium varies with the defect type and the exposure time in water. In water condition, release factor of uranium out of CANDU fuel pellets appears to be about 5 x 10-8/day, whose tendency is similar to that of cesium of the latter half of the exposure time of water. (Author)

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)

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

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)

Thermohydraulic analysis of BWR and PWR spent fuel assemblies contained within square canisters

International Nuclear Information System (INIS)

Wiles, L.E.; McCann, R.A.

1981-09-01

This report presents the results of several thermohydraulic simulations of spent fuel assembly/canister configurations performed in support of a program investigating the feasibility of storing spent nuclear fuel assemblies in canisters that would be stored in an air environment. Eleven thermohydraulic simulations were performed. Five simulations were performed using a single BWR fuel assembly/canister design. The various cases were defined by changing the canister spacing and the heat generation rate of the fuel assembly. For each simulation a steady-state thermohydraulic solution was achieved for the region inside the canister. Similarly, six simulations were performed for a single PWR fuel assembly/canister design. The square fuel rod arrays were contained in square canisters which would permit closer packing of the canisters in a storage facility. However, closer packing of the canisters would result in higher fuel temperatures which would possibly have an adverse impact on fuel integrity. Thus, the most important aspect of the analysis was to define the peak fuel assembly temperatures for each case. These results are presented along with various temperature profiles, heat flux distributions, and air velocity profiles within the canister. 48 figures, 4 tables

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.)

Radiation Analysis for Skeleton of Spent Nuclear Fuel Assembly

International Nuclear Information System (INIS)

Park, Chang Je; Na, Sang Ho; Yang, Jae Hwan; Kang, Kweon Ho

2010-11-01

ORIGEN-S code was used in order to analyze the radioactive characteristics of skeleton of the spent nuclear fuel assembly. From the results, radioactivity, decay heat for various compositions in skeleton were obtained with a variation of cooling period and axial distribution of radioactivity was calculated, too. These data will be utilized later to process and dispose the skeleton of spent nuclear fuel assembly

Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod temperatures

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1992-01-01

Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. The maximum rod temperature is a limiting factor in the amount of spent fuel that can be loaded in a transportation cask. The scope of this work is to demonstrate that reasonable and conservative spent fuel rod temperature predictions can be made using commercially available thermal analysis codes. The demonstration is accomplished by a comparison between numerical temperature predictions, with a commercially available thermal analysis code, and experimental temperature data for electrical rod heaters simulating a horizontally oriented spent fuel rod bundle

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.

Peak cladding temperature in a spent fuel storage or transportation cask

International Nuclear Information System (INIS)

Li, J.; Murakami, H.; Liu, Y.; Gomez, P.E.A.; Gudipati, M.; Greiner, M.

2007-01-01

From reactor discharge to eventual disposition, spent nuclear fuel assemblies from a commercial light water reactor are typically exposed to a variety of environments under which the peak cladding temperature (PCT) is an important parameter that can affect the characteristics and behavior of the cladding and, thus, the functions of the spent fuel during storage, transportation, and disposal. Three models have been identified to calculate the peak cladding temperature of spent fuel assemblies in a storage or transportation cask: a coupled effective thermal conductivity and edge conductance model developed by Manteufel and Todreas, an effective thermal conductivity model developed by Bahney and Lotz, and a computational fluid dynamics model. These models were used to estimate the PCT for spent fuel assemblies for light water reactors under helium, nitrogen, and vacuum environments with varying decay heat loads and temperature boundary conditions. The results show that the vacuum environment is more challening than the other gas environments in that the PCT limit is exceeded at a lower boundary temperature for a given decay heat load of the spent fuel assembly. This paper will highlight the PCT calculations, including a comparison of the PCTs obtained by different models.

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.)

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

Operation of spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs

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...

Safeguardability of advanced spent fuel conditioning process

Energy Technology Data Exchange (ETDEWEB)

Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

2004-01-01

The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

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

Management of spent fuel in Republic of Korea

International Nuclear Information System (INIS)

Pak, Hyun-Soo; Seo, In-Seok; Pak, Sang-Ki.

1989-01-01

At present in Republic of Korea, 8 PWR and 1 CANDU power plants are in operation or under construction, and the total capacity of power generation facilities has become 7.6 GWe. In addition, two PWRs of more than 900 MWe each are expected to be constructed by mid 1990s. More than 50 % of the electric power demand was supplied by nuclear power generation since 1987, but the spent fuel generated in nuclear power plants is stored in storage water tanks in respective reactor sites. The total capacity of spent fuel to be stored in the AR facilities of 9 nuclear power plants is about 2730 MTU, and the spent fuel released from these reactors since 1980 is about 810 MTU. The present capacity of AR storage pools seems to be used up by mid 1990s. According to the revised Atomic Energy Acts in May, 1986, the government is to take the responsibility of spent fuel management, and the policy of constructing the storage facilities outside reactor sites by the end of 1997 was established by the Atomic Energy Commission. The responsibility of the management of spent fuel that exceeds the present capacity of AR pools is to be taken by KEPCO, therefore the preliminary analysis of the feasible option on the extension of AR facilities and the comprehensive management plan for spent fuel placing emphasis on the research and development of away-from-reactor storage were decided. (Kako, I.)

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.

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.

Steady state transient analysis of spent nuclear fuel bundle exposed to stagnant gaseous atmosphere (Paper No. HMT-56-87)

International Nuclear Information System (INIS)

Pal, G.; Markandeya, S.G.; Venkatraj, V.

1987-01-01

This paper deals with the development of a computer code for the analysis of radiative heat exchange in rod bundles. Nuclear fuel bundles continue to generate heat even after their removal from the reactor core because of decay of fission products. During the transfer of the bundles from the core to storage bay they may pass through gaseous environment. Radiative heat exchange will be the dominant mode within the bundle under this condition. A computer code RIIEINA (Radiative Heat Exchange In Nuclear Assemblies) has been developed and used for predicting the behaviour of the spent fuel subassembly of the proposed Prototype Fast Breeder Reactor exposed to gaseous environment. The analytical model computer code and the results obtained are briefly discussed. (author). 5 refs., 5 figs

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

African Journals Online (AJOL)

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

Multi-purpose container technologies for spent fuel management

International Nuclear Information System (INIS)

2000-12-01

The management of spent nuclear fuel is an integral part of the nuclear fuel cycle. Spent fuel management resides in the back end of the fuel cycle, and is not revenue producing as electric power generation is. It instead results in a cost associated power generation. It is a major consideration in the nuclear power industry today. Because technologies, needs and circumstances vary from country to country, there is no single, standardized approach to spent fuel management. The projected cumulative amount of spent fuel generated worldwide by 2010 will be 330 000 t HM. When reprocessing is accounted for, that amount is likely to be reduced to 215 000 t HM, which is still more than twice as much as the amount now in storage. Considering the limited capacity of at-reactor (AR) storage, various technologies are being developed for increasing storage capacities. At present, many countries are developing away-from-reactor (AFR) storage in the form of pool storage or as dry storage. Further these AFR storage systems may be at-reactor sites or away-from-reactor sites (e.g. centrally located interim storage facilities, serving several reactors). The dry storage technologies being developed are varied and include vaults, horizontal concrete modules, concrete casks, and metal casks. The review of the interim storage plans of several countries indicates that the newest approaches being pursued for spent fuel management use dual-purpose and multi-purpose containers. These containers are envisaged to hold several spent fuel assemblies, and be part of the transport, storage, and possibly geological disposal systems of an integrated spent fuel management system

The feasibility of modelling coupled processes in safety analysis of spent nuclear fuel disposal

Energy Technology Data Exchange (ETDEWEB)

Rasilainen, K. [VTT Energy, Espoo (Finland); Luukkonen, A.; Niemi, A.; Poellae, J. [VTT Communities and Infrastructure, Espoo (Finland); Olin, M. [VTT Chemical Technology, Espoo (Finland)

1999-07-01

The potential of applying coupled modelling in the Finnish safety analysis programme has been reviewed. The study focused on the migration of radionuclides escaping from a spent fuel repository planned to be excavated in fractured bedrock. Two effects that can trigger various couplings in and around a spent fuel repository in Finland were studied in detail; namely heat generation in the spent fuel and the presence of deep, saline groundwaters. The latter have been observed in coastal areas. A systematic survey of the requirements of coupled modelling identified features that render such migration calculations a challenging task. In groundwater flow modelling there appears to be wide ranging uncertainty related to conceptualisation of flow systems and to the corresponding input data. In terms of migration related chemistry there appear to be large gaps in the underlying thermodynamic database for geochemical systems. Rock mechanical predictions are heavily dependent on knowing the location, structure and properties of dominant fractures; information which is extremely difficult to obtain. Conduction and convection of heat is understood well in principle. On the basis of this review, it appears that coupled migration modelling may not yet be at the stage of development that would allow its use as a standard modelling tool in performance assessments. However, a firmer basis for the conclusions reached can only be obtained after a systematic modelling exercise on a relevant and real migration problem has been carried out. (orig.)

The feasibility of modelling coupled processes in safety analysis of spent nuclear fuel disposal

International Nuclear Information System (INIS)

Rasilainen, K.; Luukkonen, A.; Niemi, A.; Poellae, J.; Olin, M.

1999-01-01

The potential of applying coupled modelling in the Finnish safety analysis programme has been reviewed. The study focused on the migration of radionuclides escaping from a spent fuel repository planned to be excavated in fractured bedrock. Two effects that can trigger various couplings in and around a spent fuel repository in Finland were studied in detail; namely heat generation in the spent fuel and the presence of deep, saline groundwaters. The latter have been observed in coastal areas. A systematic survey of the requirements of coupled modelling identified features that render such migration calculations a challenging task. In groundwater flow modelling there appears to be wide ranging uncertainty related to conceptualisation of flow systems and to the corresponding input data. In terms of migration related chemistry there appear to be large gaps in the underlying thermodynamic database for geochemical systems. Rock mechanical predictions are heavily dependent on knowing the location, structure and properties of dominant fractures; information which is extremely difficult to obtain. Conduction and convection of heat is understood well in principle. On the basis of this review, it appears that coupled migration modelling may not yet be at the stage of development that would allow its use as a standard modelling tool in performance assessments. However, a firmer basis for the conclusions reached can only be obtained after a systematic modelling exercise on a relevant and real migration problem has been carried out. (orig.)

A burner for the combustion of spent tall oil soap

Energy Technology Data Exchange (ETDEWEB)

Hughes, P.M.; Wong, J.K.; Moffatt, B.; Belanger, G. [Natural Resources Canada, Ottawa, ON (Canada). CANMET Energy Technology Centre; Soriano, D. [Brais Malouin and Associates, Montreal, PQ (Canada)

2003-07-01

Efficiency in industrial processes applies both to the form of energy involved and the many by-products resulting from the process. Tall oil soap (TOS) is a white frothy substance created during the pulping process. It contains chemicals that can be extracted for use in other industries. The processing of TOS results in a product called spent TOS. This study examined the incineration process to derive process heat from the calorific value in spent TOS. Brais Malouin and Associates (BMA) proposed that an atomizing nozzle should be used for use with this liquid in an incinerating burner. The efficiency of atomization of spent TOS with the BMA nozzle was determined by the Canada Centre for Mineral and Energy Technology (CANMET), which also characterized the combustion in a simulated boiler situation. The combustion tests were performed in the Pilot-Scale Research Boiler at the CANMET Energy Technology Centre (CETC). Pre-heating was done with a number 2 oil flame. Flame stability was determined by observing the flame through sight ports and by measuring the gas in the furnace. The experiments showed that spent TOS could successfully burn with a number 2 oil, in a proportion of 81 spent TOS to 19 oil mass ratio. As the amount of spent TOS was increased, the amount of sulphur dioxide, nitrogen oxide (NOx) and carbon monoxide decreased. The number 2 fuel oil was responsible for the sulphur dioxide in the exhaust. It is believed that the reduction in the carbon monoxide in the exhaust is attributable to the water-gas shift reaction. As the proportion of spent TOS increased, it was shown that the amount of NOx in the exhaust decreased rapidly. A bluish-green molten deposit formed in the furnace near the burner came from copper and manganese found in the ash of the spent TOS. 7 refs., 7 tabs., 16 figs.

Spent fuel storage requirements 1993--2040

International Nuclear Information System (INIS)

1994-09-01

Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges

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

Seismic analysis and design of spent subassembly storage bay (SSSB) pool

International Nuclear Information System (INIS)

Abdul Gani, H.I.; Ramanjaneyulu, K.V.S.; Pillai, C.S.; Chetal, S.C.

2003-01-01

Fuel bundles, after their specified stay in reactor core, are replaced by fresh fuel for sustaining power generation at rated levels. The irradiated fuel subassembly, removed fresh from core, known as spent fuel sub assembly, is radioactive and decay heat generating. It needs to be cooled before it becomes amenable for handling, either for reprocessing or for immobilisation. For this purpose, it is immersed in a pool of water, retained in a concrete structure referred as Spent Subassembly Storage Bay (SSSB) pool. The height of water column above fuel bundles is arrived from shielding considerations. SSSB pool is one of the nuclear safety related structures and warrants rigorous analysis and design. The SSSB pool, in case of PFBR 500 MW(e) is located in fuel building. It is a stainless steel lined. water retaining rectangular R.C.C. open tank of size 7.5 X 29.0 m, with a height of 11.0 m. This structure is analysed for two levels of site specific earthquakes taking in to account liquid structure interactions as per ASCE-4, 1998. The design of walls and bottom slab is carried out satisfying the AERB code for nuclear safety related structures. Analysis and design of SSSB pool of PFBR is presented in the following paper. (author)

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

The states of the art of the nondestructive assay of spent nuclear fuel assemblies. A critical review of the Spent Fuel NDA Project of the U.S. Department of Energy's Next Generation Safeguards Initiative

International Nuclear Information System (INIS)

Bolind, Alan Michael; Seya, Michio

2015-12-01

The state of the art of the nondestructive assay of spent nuclear fuel assemblies is represented by the results of the Spent Fuel Nondestructive Assay Project of the Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy / National Nuclear Security Administration. This report surveys the fourteen advanced nondestructive assay (NDA) techniques that were examined by the NGSI. For each technique, it explains how the technique operates, the NGSI's design of an instrument that uses the technique, how the data are analyzed, and the technique's chief limitations. After this survey of the NDA techniques, the report then discusses and critiques the current paradigm of the practice of NDA of spent fuel assemblies. It shows how the current main problem in the NDA of spent fuel assemblies—namely, an unacceptably large uncertainty in the assay results—is caused primarily by using too few independent NDA measurements. Because the physics of the NDA of spent fuel assemblies is three dimensional, at least three independent NDA measurements are required. Thus, NDA results should be able to be improved dramatically by combining the fourteen advanced NDA techniques plus other existing NDA techniques into appropriate combinations of three techniques. This report evaluates the NGSI's proposed NDA combinations according to these principles. (author)

Novel approach to recover cobalt and lithium from spent lithium-ion battery using oxalic acid

Energy Technology Data Exchange (ETDEWEB)

Zeng, Xianlai; Li, Jinhui, E-mail: jinhui@tsinghua.edu.cn; Shen, Bingyu

2015-09-15

Highlights: • Short-cut recovery of cobalt and lithium was directly obtained using oxalic acid. • Short-cut recovery process was optimized for a high recovery rate. • Leaching process was controlled by chemical reaction. • Leaching order of the sampling LiCoO{sub 2} using oxalic acid was first proposed. - Abstract: With the booming of consumer electronics (CE) and electric vehicle (EV), a large number of spent lithium-ion battery (LIBs) have been generated worldwide. Resource depletion and environmental concern driven from the sustainable industry of CE and EV have motivated spent LIBs should be recovered urgently. However, the conventional process combined with leaching, precipitating, and filtering was quite complicated to recover cobalt and lithium from spent LIBs. In this work, we developed a novel recovery process, only combined with oxalic acid leaching and filtering. When the optimal parameters for leaching process is controlled at 150 min retention time, 95 °C heating temperature, 15 g L{sup −1} solid–liquid ratio, and 400 rpm rotation rate, the recovery rate of lithium and cobalt from spent LIBs can reach about 98% and 97%, respectively. Additionally, we also tentatively discovered the leaching mechanism of lithium cobalt oxide (LiCoO{sub 2}) using oxalic acid, and the leaching order of the sampling LiCoO{sub 2} of spent LIBs. All the obtained results can contribute to a short-cut and high-efficiency process of spent LIBs recycling toward a sound closed-loop cycle.

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.

Spent fuel treatment in Japan

International Nuclear Information System (INIS)

Takahashi, K.

1999-01-01

In Japan, 52 nuclear power reactors are operating with a total power generation capacity of 45 GWe. The cumulative amount of spent fuel arising, as of March 1998, is about 14,700 W. Spent fuel is reprocessed and recovered nuclear materials are to be recycled in LWRs and FBRs. Pu utilization in LWRs will commence in 1999. In January 1997, short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of the reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, back-end measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away-from-reactor sites, considering the increasing amount of spent fuel arising. Valuable experience was been accumulated at the Tokai Reprocessing Plant (TRP), from the start of hot operation in 1977 up to now. The role of the TRP will be changed from an operation-oriented to a more R and D oriented facility, when PNC is reorganized into the new organization JNC. The Rokkasho reprocessing plant is under construction and is expected to commence operation in 2003. R and D of future recycling technologies is also continued for the establishment of a nuclear fuel cycle based on FBRs and LWRs. (author)

Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

International Nuclear Information System (INIS)

Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

2012-01-01

After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

Status of spent fuel shipping cask development

International Nuclear Information System (INIS)

Hall, I.K.; Hinschberger, S.T.

1989-01-01

This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

Spent nuclear fuel storage - Basic concept

International Nuclear Information System (INIS)

Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

2009-01-01

According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)

Loss of spent fuel pool cooling PRA: Model and results

International Nuclear Information System (INIS)

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

Energy Technology Data Exchange (ETDEWEB)

Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

Innovative application of ionic liquid to separate Al and cathode materials from spent high-power lithium-ion batteries

Energy Technology Data Exchange (ETDEWEB)

Zeng, Xianlai; Li, Jinhui, E-mail: jinhui@tsinghua.edu.cn

2014-04-01

Highlights: • Manual dismantling is superior in spent high-power LiBs recycling. • Heated ionic liquid can effectively separate Al and cathode materials. • Fourier’s law was adopted to determine the heat transfer mechanism. • The process of spent LiBs recycling with heated ionic liquid dismantling was proposed. - Abstract: Because of the increasing number of electric vehicles, there is an urgent need for effective recycling technologies to recapture the significant amount of valuable metals contained in spent lithium-ion batteries (LiBs). Previous studies have indicated, however, that Al and cathode materials were quite difficult to separate due to the strong binding force supplied by the polyvinylidene fluoride (PVDF), which was employed to bind cathode materials and Al foil. This research devoted to seek a new method of melting the PVDF binder with heated ionic liquid (IL) to separate Al foil and cathode materials from the spent high-power LiBs. Theoretical analysis based on Fourier’s law was adopted to determine the heat transfer mechanism of cathode material and to examine the relationship between heating temperature and retention time. All the experimental and theoretic results show that peel-off rate of cathode materials from Al foil could reach 99% when major process parameters were controlled at 180 °C heating temperature, 300 rpm agitator rotation, and 25 min retention time. The results further imply that the application of IL for recycling Al foil and cathode materials from spent high-power LiBs is highly efficient, regardless of the application source of the LiBs or the types of cathode material. This study endeavors to make a contribution to an environmentally sound and economically viable solution to the challenge of spent LiB recycling.

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

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)

Spent fuel generated by the Kozloduy nuclear power plant within the period 1974 - 1994

International Nuclear Information System (INIS)

Peev, P.

1994-01-01

The spent fuel management during the 20-year operation of Kozloduy NPP is described. Formally this period is divided into two stages. The first one covers 1977 - 1988 when the spent fuel after short-term (3 years) storage in the reactor building was dispatched to the former Soviet Union. Within this period 21 shipments of spent fuel (about 3086 fuel assemblies) with various level of enrichment and burnup were performed. The second stage covers the period 1988 - 1994. In that period the authorities responsible for the safe operation of Kozloduy NPP faced a number of problems related to necessity of on-site spent-fuel storage commissioning. A reassessment of the seismic risk after Vrancha earthquake and the Russia's attitude towards the former Soviet Union policy of spent fuel storage was discussed

Spent fuel generated by the Kozloduy nuclear power plant within the period 1974 - 1994

Energy Technology Data Exchange (ETDEWEB)

Peev, P [National Electricity Company, Sofia (Bulgaria)

1994-12-31

The spent fuel management during the 20-year operation of Kozloduy NPP is described. Formally this period is divided into two stages. The first one covers 1977 - 1988 when the spent fuel after short-term (3 years) storage in the reactor building was dispatched to the former Soviet Union. Within this period 21 shipments of spent fuel (about 3086 fuel assemblies) with various level of enrichment and burnup were performed. The second stage covers the period 1988 - 1994. In that period the authorities responsible for the safe operation of Kozloduy NPP faced a number of problems related to necessity of on-site spent-fuel storage commissioning. A reassessment of the seismic risk after Vrancha earthquake and the Russia`s attitude towards the former Soviet Union policy of spent fuel storage was discussed.

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

A central spent fuel storage in Sweden

International Nuclear Information System (INIS)

Gustafsson, B.; Hagberth, R.

1978-01-01

A planned central spent fuel storage facility in Sweden is described. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed. (author)

Spent Pot Lining Characterization Framework

Science.gov (United States)

Ospina, Gustavo; Hassan, Mohamed I.

2017-09-01

Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.

Spent fuel storage requirements, 1988

International Nuclear Information System (INIS)

1988-10-01

Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs

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.

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

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)

Study on characteristics of spent PWR cladding hull for categorizing into Non-TRU waste

International Nuclear Information System (INIS)

Jung, In Ha; Kim, Jong Ho; Park, Jang Jin; Shin, Jin Myeong; Lee, Ho Hee; Yang, Myung Seung

2005-01-01

AFCI and GEN-IV programs aim for decreasing the high level radioactive wastes to be disposed. They also try to get valuable materials to recycle as resources such as uranium and plutonium. On the other hand, cladding hull expected to be one-thirds in volume of spent fuel assembly has not studied so much in the point view of recycling to reuse. Since traditional process of reprocessing was wet process, cladding hull generating through the reprocessing process was unavoidably contaminated with TRU by acid solvent during the process. Therefore, cladding hull has been classified into TRU wastes or high level wastes. According to the strategy for TRU high level radioactive wastes of USA as well as Korea, it regulates in two respects. One is activity and the other is heat generation. In respect of activity, TRU waste contains more than 100 nCi/kg of alpha emits with longer half life than 20 years and higher than 92 in atomic number. Also, wastes are categorized into TRU waste when it generates higher than 2kW/m3, in the respect of heat generation. Our results as well as literatures, almost all of TRU nuclides in the cladding hull are responsible for remained uranium and plutonium owing to pellet-cladding interaction. In addition, recoiled fission products on the surface of the cladding hull serve as heat generator. Up to now, decontamination of the cladding hull generating from the reprocessing of wet process is regarded as valueless and un-economic works owing to the amount of second waste produced

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

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

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

Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

2006-01-01

Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

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...

Spent nuclear fuel disposal liability insurance

International Nuclear Information System (INIS)

Martin, D.W.

1984-01-01

This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

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

Heat transfer performance of heat pipe for passive cooling of spent fuel pool

International Nuclear Information System (INIS)

Wang Minglu; Xiong Zhengqin; Gu Hanyang; Ye Cheng; Cheng Xu

2014-01-01

A large-scale loop heat pipe has no electricity driven component and high efficiency of heat transfer. It can be used for the passive cooling of the SFP after SBO to improve the safety performance of nuclear power plants. In this paper, such a large-scale loop heat pipe is studied experimentally. The heat transfer rate, evaporator average heat transfer coefficient operating temperature, operating pressure and ammonia flow rate have been obtained with the water flow ranging from 0.007 m/s to 0.02 m/s outside the evaporator section, heating water temperature in the range of 50 to 90℃, air velocity outside the condensation section ranging from 0.5 to 2.5 m/s. It is found that the heat transfer rate reaches as high as 20.1 kW. Parametric analysis indicates that, the heat transfer rate and ammonia flow rate are influenced significantly by hot water inlet temperature and velocity, while beyond 1.5 m/s, the effect of air velocity outside the condensation section is minor. (authors)

Spent fuel dry storage in Hungary

International Nuclear Information System (INIS)

Buday, G.; Szabo, B.; Oerdoegh, M.; Takats, F.

1999-01-01

Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. Since 1989, approximately 40-50% of the total annual electricity generation of the country has been supplied by this plant. The fresh fuel is imported from Russia. Most of the spent fuel assemblies have been shipped back to Russia. Difficulties with spent fuel transportation to Russia have begun in 1992. Since that time, some of the shipments were delayed, some of them were completely cancelled, thus creating a backlog of spent fuel filling all storage positions of the plant. To provide assurance of the continued operation, Paks NPPs management decided to implement an independent spent fuel storage facility and chose GEC-Althom's MVDS design. The construction of the facility started in February 1995 and the first spent fuel assembly was placed in the store in September 1997. The paper gives an overview of the situation, describing the conditions leading to the construction of the dry storage facility at Paks and its implementation. Finally, some information is given about the new Public Agency for Radioactive Waste Management established this year and responsible for managing the issues related to spent fuel management. (author)

Development of evaluation method for heat removal design of dry storage facilities. pt. 1. Heat removal test on vault storage system of cross flow type

International Nuclear Information System (INIS)

Sakamoto, Kazuaki; Koga, Tomonari; Wataru, Masumi; Hattori, Yasuo

1997-01-01

The report describes the result of heat removal test of passive cooling vault storage system of cross flow type using 1/5 scale model. Based on a prospect of steady increase in the amount of spent fuel, it is needed to establish large capacity dry storage technologies for spent fuel. Air flow patterns, distributions of air temperature and velocity were measured, by which heat removal characteristics of the system were made clear. Air flow patterns in the storage module depended on the ratio of the buoyant force to the inertial force; the former generated by the difference of air temperatures and the height of the storage module, the latter by the difference of air densities between the outlet of the storage module and ambience and the height of the chimney of the storage facility. A simple method to estimate air flow patterns in the storage module was suggested, where Ri(Richardson) number was applied to represent the ratio. Moreover, heat transfer coefficient from a model of storage tube to cooling air was evaluated, and it was concluded that the generalized expression of heat transfer coefficient for common heat exchangers could be applied to the vault storage system of cross flow type, in which dozens of storage tubes were placed in a storage module. (author)

Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

International Nuclear Information System (INIS)

Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.

2004-01-01

The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349

PWR and BWR spent fuel assembly gamma spectra measurements

Energy Technology Data Exchange (ETDEWEB)

Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-10-11

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

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

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

Investigation of novel spent fuel verification system for safeguard application

Energy Technology Data Exchange (ETDEWEB)

Lee, Haneol; Yim, Man-Sung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source.

Investigation of novel spent fuel verification system for safeguard application

International Nuclear Information System (INIS)

Lee, Haneol; Yim, Man-Sung

2016-01-01

Radioactive waste, especially spent fuel, is generated from the operation of nuclear power plants. The final stage of radioactive waste management is disposal which isolates radioactive waste from the accessible environment and allows it to decay. The safety, security, and safeguard of a spent fuel repository have to be evaluated before its operation. Many researchers have evaluated the safety of a repository. These researchers calculated dose to public after the repository is closed depending on their scenario. Because most spent fuel repositories are non-retrievable, research on security or safeguards of spent fuel repositories have to be performed. Design based security or safeguard have to be developed for future repository designs. This study summarizes the requirements of future spent fuel repositories especially safeguards, and suggests a novel system which meets the safeguard requirements. Applying safeguards to a spent fuel repository is becoming increasingly important. The future requirements for a spent fuel repository are suggested by several expert groups, such as ASTOR in IAEA. The requirements emphasizes surveillance and verification. The surveillance and verification of spent fuel is currently accomplished by using the Cerenkov radiation detector while spent fuel is being stored in a fuel pool. This research investigated an advanced spent fuel verification system using a system which converts spent fuel radiation into electricity. The system generates electricity while it is conveyed from a transportation cask to a disposal cask. The electricity conversion system was verified in a lab scale experiment using an 8.51GBq Cs-137 gamma source

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.

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

Overhaul of the heat exchanger in JRR-3

International Nuclear Information System (INIS)

Ouchi, Yasuhiro; Kawamata, Satoshi; Taguchi, Yuji; Kamiishi, Eigo; Koda, Nobuyuki

2013-01-01

In JRR-3, heat exchangers are installed in the cooling system equipment to remove the heat generated in the nuclear reactor, For the heat exchangers, overhaul inspection based on the JRR-3 reactor facility maintenance plan, as well as the inspection and maintenance based on reactor facility security provisions and JRR-3 operation guidelines are systematically conducted. Considering the results of overhaul inspection, the second overhaul inspection was applied to the primary coolant heat exchanger. The thinning of heat transfer tubes is within judgment standards with little effects of aging, which verified their soundness. From the fact that the effects of corrosion have been confirmed on the inside of the water chamber, repair work through overlay welding or the like is planned in the next overhaul. As for heavy water heat exchanger and the spent fuel pool water heat exchanger, it is planned to conduct the second overhaul inspection in FY2013 to confirm their soundness. (A.O.)

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.)

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

Heat transfer and pressure drop characteristics of the tube bank fin heat exchanger with fin punched with flow redistributors and curved triangular vortex generators

Science.gov (United States)

Liu, Song; Jin, Hua; Song, KeWei; Wang, LiangChen; Wu, Xiang; Wang, LiangBi

2017-10-01

The heat transfer performance of the tube bank fin heat exchanger is limited by the air-side thermal resistance. Thus, enhancing the air-side heat transfer is an effective method to improve the performance of the heat exchanger. A new fin pattern with flow redistributors and curved triangular vortex generators is experimentally studied in this paper. The effects of the flow redistributors located in front of the tube stagnation point and the curved vortex generators located around the tube on the characteristics of heat transfer and pressure drop are discussed in detail. A performance comparison is also carried out between the fins with and without flow redistributors. The experimental results show that the flow redistributors stamped out from the fin in front of the tube stagnation points can decrease the friction factor at the cost of decreasing the heat transfer performance. Whether the combination of the flow redistributors and the curved vortex generators will present a better heat transfer performance depends on the size of the curved vortex generators. As for the studied two sizes of vortex generators, the heat transfer performance is promoted by the flow redistributors for the fin with larger size of vortex generators and the performance is suppressed by the flow redistributors for the fin with smaller vortex generators.

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Energy Technology Data Exchange (ETDEWEB)

Wolf, S.; Holmes, D.H.

1977-04-01

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

An experimental study of the enhanced heating capacity of an electric heat pump (EHP) using the heat recovered from a gas engine generator

Energy Technology Data Exchange (ETDEWEB)

Kim, Cheol Min; Chang, Se Dong [HAC R and D Laboratory, LG Electronics, 327-23 Gasan-Dong, Geumcheon-gu, Seoul 153-802 (Korea); Lee, Jaekeun; Hwang, Yujin [School of Mechanical Engineering, Pusan National University, San 30, Changjeon-Dong, Keumjeong-Ku, Busan 609-735 (Korea)

2009-11-15

This paper is concerned with the effect of recovered heat on the heating capacity of an Electric Heat Pump (EHP), which is supplied with electric power and recovered heat from a gas engine generator system. Two methods of supplying recovery heat are examined: (i) to the refrigerant with the discharge line heat exchanger (HEX), and (ii) to the refrigerant of the evaporator with the sub-evaporator. Heating capacity, input power and coefficient of performance (COP) were investigated and compared for each heat recovery method. Conclusively, we found that the second method was most reasonable to recover wasted heat and increased system COP by 215%. (author)

Effect of heat transfer tube leak on dynamic characteristic of steam generator

International Nuclear Information System (INIS)

Sun Baozhi; Shi Jianxin; Li Na; Zheng Lusong; Liu Shanghua; Lei Yu

2015-01-01

Taking the steam generator of Daya Bay Nuclear Power Station as the research object, one-dimensional dynamic model of the steam generator based on drift flux theory and leak model of heat transfer tube were established. Steady simulation of steam generator under different conditions was carried out. Based on verifying the drift flux model and leak model of heat transfer tube, the effect of leak location and flow rate under different conditions on steam generator's key parameters was studied. The results show that the drift flux model and leak model can reflect the law of key parameter change accurately such as vapor mass fraction and steam pressure under different leak cases. The variation of the parameters is most apparent when the leak is at the entrance of boiling section and vapor mass fraction varies from 0.261 to 0.163 when leakage accounts for 5% of coolant flow rate. The successful prediction of the effect of heat transfer tube leak on dynamic characteristics of the steam generator based on drift flux theory supplies some references for monitoring and taking precautionary measures to prevent heat transfer tube leak accident. (authors)

Condition monitoring of steam generator by estimating the overall heat transfer coefficient

International Nuclear Information System (INIS)

Furusawa, Hiroaki; Gofuku, Akio

2013-01-01

This study develops a technique for monitoring in on-line the state of the steam generator of the fast-breeder reactor (FBR) “Monju”. Because the FBR uses liquid sodium as coolant, it is necessary to handle liquid sodium with caution due to its chemical characteristics. The steam generator generates steam by the heat of secondary sodium coolant. The sodium-water reaction may happen if a pinhole or crack occurs at the thin metal tube wall that separates the secondary sodium coolant and water/steam. Therefore, it is very important to detect an anomaly of the wall of heat transfer tubes at an early stage. This study aims at developing an on-line condition monitoring technique of the steam generator by estimating overall heat transfer coefficient from process signals. This paper describes simplified mathematical models of superheater and evaporator to estimate the overall heat transfer coefficient and a technique to diagnose the state of the steam generator. The applicability of the technique is confirmed by several estimations using simulated process signals with artificial noises. The results of the estimations show that the developed technique can detect the occurrence of an anomaly. (author)

Regional trends in radiogenic heat generation in the Precambrian basement of the Western Canadian Basin

Science.gov (United States)

Jones, F. W.; Majorowicz, J. A.

Radiogenic heat generation values for 381 basement samples from 229 sites in the western Canadian basin exhibit a lognormal frequency distribution. The mean value = 2.06 (S.D. = 1.22) µWm-3 is larger than the radiogenic heat generation values reported for the shield in the Superior (ca. 1.2 µWm-3, Jessop and Lewis, 1978) and Churchill (ca. 0.7 µWm-3, Drury, 1985) provinces. When equal Log A contour intervals are used to map the basement heat generation, three large zones of relatively high heat generation are found. One coincides with the Peace River Arch basement structure and one with the Athabasca axis (Darnley, 1981). There is no apparent indication of increased heat flow through the Paleozoic formations associated with these two zones. The third zone, in southwestern Saskatchewan, coincides with a high heat flow zone in the Swift Current area. The lack of correlation between heat flow and heat generation in Alberta may be due to the disturbance to the heat flow in the Paleozoic formations by water motion, or may indicate that the heat is from uranium, thorium and potassium isotope enrichment near the basement surface rather than enrichment throughout the entire upper crust.

Method and apparatus for enhanced heat recovery from steam generators and water heaters

Science.gov (United States)

Knight, Richard A.; Rabovitser, Iosif K.; Wang, Dexin

2006-06-27

A heating system having a steam generator or water heater, at least one economizer, at least one condenser and at least one oxidant heater arranged in a manner so as to reduce the temperature and humidity of the exhaust gas (flue gas) stream and recover a major portion of the associated sensible and latent heat. The recovered heat is returned to the steam generator or water heater so as to increase the quantity of steam generated or water heated per quantity of fuel consumed. In addition, a portion of the water vapor produced by combustion of fuel is reclaimed for use as feed water, thereby reducing the make-up water requirement for the system.

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

International Nuclear Information System (INIS)

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

Efficient heat generation in large-area graphene films by electromagnetic wave absorption

Science.gov (United States)

Kang, Sangmin; Choi, Haehyun; Lee, Soo Bin; Park, Seong Chae; Park, Jong Bo; Lee, Sangkyu; Kim, Youngsoo; Hong, Byung Hee

2017-06-01

Graphene has been intensively studied due to its outstanding electrical and thermal properties. Recently, it was found that the heat generation by Joule heating of graphene is limited by the conductivity of graphene. Here we suggest an alternative method to generate heat on a large-area graphene film more efficiently by utilizing the unique electromagnetic (EM) wave absorption property of graphene. The EM wave induces an oscillating magnetic moment generated by the orbital motion of moving electrons, which efficiently absorbs the EM energy and dissipate it as a thermal energy. In this case, the mobility of electron is more important than the conductivity, because the EM-induced diamagnetic moment is directly proportional to the speed of electron in an orbital motion. To control the charge carrier mobility of graphene we functionalized substrates with self-assembled monolayers (SAM). As the result, we find that the graphene showing the Dirac voltage close to zero can be more efficiently heated by EM waves. In addition, the temperature gradient also depends on the number of graphene. We expect that the efficient and fast heating of graphene films by EM waves can be utilized for smart heating windows and defogging windshields.

Heat generation in agglomerated ferrite nanoparticles in an alternating magnetic field

International Nuclear Information System (INIS)

Lima, E Jr; De Biasi, E; Mansilla, M Vasquez; Saleta, M E; Granada, M; Troiani, H E; Zysler, R D; Effenberger, F B; Rossi, L M; Rechenberg, H R

2013-01-01

The role of agglomeration and magnetic interparticle interactions in heat generation of magnetic ferrofluids in an ac magnetic field is still unclear, with apparent discrepancy between the results presented in the literature. In this work, we measured the heat generating capability of agglomerated ferrite nanoparticles in a non-invasive ac magnetic field with f = 100 kHz and H 0 = 13 kA m -1 . The nanoparticles were morphologically and magnetically characterized, and the specific absorption rate (SAR) for our ac magnetic field presents a clear dependence on the diameter of the nanoparticles, with a maximum SAR = 48 W g -1 for 15 nm. Our agglomerated nanoparticles have large hydrodynamic diameters, thus the mechanical relaxation can be neglected as a heat generation mechanism. Therefore, we present a model that simulates the SAR dependence of the agglomerated samples on the diameter of the nanoparticles based on the hysteresis losses that is valid for the non-linear region (with H 0 comparable to the anisotropy field). Our model takes into account the magnetic interactions among the nanoparticles in the agglomerate. For comparison, we also measured the SAR of non-agglomerated nanoparticles in a similar diameter range, in which Néel and Brown relaxations dominate the heat generation.

Comparison of spent fuel management fee collection alternatives

International Nuclear Information System (INIS)

White, M.K.; Engel, R.L.

1979-01-01

Five alternative methods for recovering the costs of spent fuel management were evaluated. These alternatives consist of collecting the fee for various components of spent fuel management cost (AFR basin storage, transportation from AFR basin to the repository, packaging, repository, R and D, and government overhead) at times ranging from generation of power to delivery of the spent fuel to the government. The five fee collection mechanisms were analyzed to determine how well they serve the interests of the public and the electricity ratepayer

ECONOMIC COMPARATIVE EVALUATION OF COMBINATION OF ACTIVATED CARBON GENERATION AND SPENT ACTIVATED CARBON REGENERATION PLANTS

Directory of Open Access Journals (Sweden)

TINNABHOP SANTADKHA

2017-12-01

Full Text Available The purpose of this study was to investigate the maximum annual profit of proposed three project plants as follows: (i a generation process of activated carbon (AC prepared from coconut shells; (ii a regeneration process of spent AC obtained from petrochemical industries; and (iii a project combined the AC generation process with the regeneration process. The maximum annual profit obtained from the sole regeneration plant was about 1.2- and 15.4- fold higher than that obtained from the integrated and the generation plants, respectively. The sensitivity of selected variables to net present value (NPV, AC sales price was the most sensitive to NPV while fixed costs of generation and regeneration, and variable cost of regeneration were the least sensitive to NPV. Based on the optimal results of each project plant, the economic indicators namely NPV, return on investment (ROI, internal rate of return (IRR, and simple payback period (SPP were determined. Applying a rule of thumb of 12% IRR and 7-year SPP, the AC sales prices for the generation, regeneration, and integrated plants were 674.31, 514.66 and 536.66 USD/ton of product, respectively. The economic analysis suggested that the sole regeneration project yields more profitable.

Reuse of spent bleaching earth by polymerisation of residual organics.

Science.gov (United States)

Beshara, Abdelhamid; Cheeseman, Christopher R

2014-10-01

Spent bleaching earth (SBE) is a waste generated by the edible oil industry that currently has limited options for beneficial reuse. In excess of ∼2 million tonnes per year of SBE is generated world-wide with major quantities available in the middle-east where significant volumes of edible oils are produced. Low pressure compaction followed by heat treatment at 150°C causes polymerisation of the residual organic components in SBE and this produces monolithic samples with high unconfined compressive strengths (54MPa). SBE can therefore be used to manufacture novel clay blocks for use in construction that are bonded by polymerised vegetable oil. This represents a new, innovative and resource efficient application for SBE. In this research, commercial SBE has been characterised and the effects of key processing variables (temperature and compaction pressure) on the compressive strength, porosity and density of the SBE clay blocks are reported and the mechanisms responsible for strength development are discussed. Copyright © 2014 Elsevier Ltd. All rights reserved.

Isothermal calorimeter for measurements of time-dependent heat generation rate in individual supercapacitor electrodes

Science.gov (United States)

Munteshari, Obaidallah; Lau, Jonathan; Krishnan, Atindra; Dunn, Bruce; Pilon, Laurent

2018-01-01

Heat generation in electric double layer capacitors (EDLCs) may lead to temperature rise and reduce their lifetime and performance. This study aims to measure the time-dependent heat generation rate in individual carbon electrode of EDLCs under various charging conditions. First, the design, fabrication, and validation of an isothermal calorimeter are presented. The calorimeter consisted of two thermoelectric heat flux sensors connected to a data acquisition system, two identical and cold plates fed with a circulating coolant, and an electrochemical test section connected to a potentiostat/galvanostat system. The EDLC cells consisted of two identical activated carbon electrodes and a separator immersed in an electrolyte. Measurements were performed on three cells with different electrolytes under galvanostatic cycling for different current density and polarity. The measured time-averaged irreversible heat generation rate was in excellent agreement with predictions for Joule heating. The reversible heat generation rate in the positive electrode was exothermic during charging and endothermic during discharging. By contrast, the negative electrode featured both exothermic and endothermic heat generation during both charging and discharging. The results of this study can be used to validate existing thermal models, to develop thermal management strategies, and to gain insight into physicochemical phenomena taking place during operation.

Spent-Fuel Test - Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Executive summary of final results

International Nuclear Information System (INIS)

Patrick, W.C.

1986-01-01

This summary volume outlines results that are covered in more detail in the final report of the Spent-Fuel Test - Climate project. The project was conducted between 1978 and 1983 in the granitic Climax stock at the Nevada Test Site. Results indicate that spent fuel can be safely stored for periods of years in this host medium and that nuclear waste so emplaced can be safely retrieved. We also evaluated the effects of heat and radiation (alone and in combination) on emplacement canisters and the surrounding rock mass. Storage of the spent-fuel affected the surrounding rock mass in measurable ways, but did not threaten the stability or safety of the facility at any time

Generating a heated fluid using an electromagnetic radiation-absorbing complex

Science.gov (United States)

Halas, Nancy J.; Nordlander, Peter; Neumann, Oara

2018-01-09

A vessel including a concentrator configured to concentrate electromagnetic (EM) radiation received from an EM radiation source and a complex configured to absorb EM radiation to generate heat. The vessel is configured to receive a cool fluid from the cool fluid source, concentrate the EM radiation using the concentrator, apply the EM radiation to the complex, and transform, using the heat generated by the complex, the cool fluid to the heated fluid. The complex is at least one of consisting of copper nanoparticles, copper oxide nanoparticles, nanoshells, nanorods, carbon moieties, encapsulated nanoshells, encapsulated nanoparticles, and branched nanostructures. Further, the EM radiation is at least one of EM radiation in an ultraviolet region of an electromagnetic spectrum, in a visible region of the electromagnetic spectrum, and in an infrared region of the electromagnetic spectrum.

Thermoelectric Power Generation Utilizing the Waste Heat from a Biomass Boiler

Science.gov (United States)

Brazdil, Marian; Pospisil, Jiri

2013-07-01

The objective of the presented work is to test the possibility of using thermoelectric power to convert flue gas waste heat from a small-scale domestic pellet boiler, and to assess the influence of a thermoelectric generator on its function. A prototype of the generator, able to be connected to an existing device, was designed, constructed, and tested. The performance of the generator as well as the impact of the generator on the operation of the boiler was investigated under various operating conditions. The boiler gained auxiliary power and could become a combined heat and power unit allowing self-sufficient operation. The created unit represents an independent source of electricity with effective use of fuel.

LSDS Development for Isotopic Fissile Assay in Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-07-01

As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel.

LSDS Development for Isotopic Fissile Assay in Spent Fuel

International Nuclear Information System (INIS)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

2011-01-01

As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel

Incineration of spent ion exchange resin

International Nuclear Information System (INIS)

Hasegawa, Chiaki

1990-01-01

It is a pressing need to reduce radioactive waste which is generated from the maintenance and operation of a nuclear power plant. Incineration of low level combustible solid waste such as polyethylene seats, paper and others have been successfully performed since 1984 at the Shimane Nuclear Power Station. Furthermore, for extending incineration treatment to spent ion exchange resin, the incineration test was carried out in 1989. However, as the cation exchange resin contains sulfur and then incineration generates SOx gases, so the components of this facility will be in a corrosive environment. We surveyed incineration conditions to improve the corrosive environment at the exhaust gas treatment system. This paper includes these test results and improved method to incinerate spent ion exchange resin. (author)

Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

International Nuclear Information System (INIS)

Rhee, I.; Cho, D.; Yeon, J.

2003-01-01

Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

Solar hot-water generation and heating - Kombi-Kompakt+

International Nuclear Information System (INIS)

Haller, M.; Vogelsanger, P.

2005-01-01

This final report for the Swiss Federal Office of Energy (SFOE) describes new testing facilities at the Institute for Solar Technology in Rapperswil, Switzerland, that allow the testing of solar systems the whole year through. The systems tested feature the combined generation of heat for hot water storage vessels and heat for space heating. The test method used, the Concise Cycle Test (CCT) is described. The results of tests made on a large number of systems demonstrate that it is especially important to have a test system that allows the solar market to be protected from unsatisfactory systems. Good co-operation with manufactures is noted. As the test method includes tests with secondary energy sources such as oil or gas, certain problems in this area were discovered and corrected. Further tests are to be made with systems using biomass as a secondary source of heat

Status and operational experience report of spent fuel storage facility in Kozloduy NPP for the period 1990 - 1994

Energy Technology Data Exchange (ETDEWEB)

Kalimanov, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1994-12-31

Spent Fuel Storage Facility (SFSF) of Kozloduy NPP is designed for a long-term storage of 4920 spent fuel assemblies which are generated by all units for ten year operational period. The assemblies are stored in SFSF after 3 year storage in the reactor cooling pool. The SFSF operational safety is ensured by a number of strictly followed regulations related to: arrangement of the assemblies and conditions at which they are stored; transportation of the assemblies to the facility; residual heat removal; quality of the water used in the storage pool; water temperature and level control. Two independent groups of experts have carried out investigations to study the building safety. Their reports have been considered and accepted by the council of the Ministry of Environment which was the final step in licensing the SFSF.

Calculations of the Temperature Evolution of a Repository for Spent Fuel in Crystalline and Sedimentary Rocks

International Nuclear Information System (INIS)

Sato, R.; Sasaki, T.; Ando, K.; Smith, P.A.; Schneider, J.W.

1998-08-01

Thermal evolution is a factor influencing repository design, and must be considered in safety assessment, since many of the processes that affect the long-term safety are temperature dependent. This report presents calculations of the thermal evolution of a repository for spent nuclear fuel. The calculations are based on a provisional repository near-field design in which spent fuel is encapsulated in composite copper-steel canisters, which are emplaced centrally along the horizontal axes of repository tunnels, with the space around the canisters backfilled with bentonite. The temperature of these near-field components varies with time, due to the radiogenic heat produced by the spent fuel. The rate of heat production per canister depends on the initial composition of the fuel, its reactor history, the period of intermediate storage before final disposal and the loading of the canisters. The rate decreases with time, as shorter-lived radionuclides decay. The base-case calculation considers spent fuel that is assumed to generate 1000 W per canister, 40 years after unloading of the fuel from the reactor. The results of the base case calculation indicate that the temperatures at the bentonite/host rock interface, at the centre of the bentonite and at the bentonite/canister interface rise to 98 o C, 103 o C and 126 o C, respectively, before declining towards the ambient temperature of the host rock which, in the base case, is taken to be the crystalline basement of Northern Switzerland. In addition to the base case, parameter variations are examined that investigate the sensitivity of thermal evolution to alternative heat output, design specifications and to uncertainties in material properties. Key findings include (i), that an increase in heat generation to 1500 W per canister 40 years after unloading results in a significant increase of repository temperatures (e.g. at the bentonite/host rock interface, an increase of 22 o C is observed), (ii), that a decrease in

Analyses of pebble-bed reactors for the generation of heat for heating purposes

International Nuclear Information System (INIS)

Muehlensiep, J.; Fricke, U.; Inhester, K.H.; Kugeler, K.; Phlippen, P.W.; Schmidtlein, P.; Swatoch, R.; Wagner, U.

1986-10-01

Marginal conditions are described for the use of a nuclear power reactor for long-distance heat supply in densely populated areas. For the design of the high-temperature heat generating reactor, plant components and possible arrangements are analyzed with consideration to safety and costs. System sizes are reasonably chosen on the basis of analyzed parameters, the paramount design goal being to adequately retain the fission products in the coated particles of the fuel elements, in anticipation of probable accidents. With the help of the data record obtained, a system is designed with a cuboid-shaped core as its characteristic feature; the advantage of the core consists in the fact that it quickly discharges the after-heat outwards even in case of a hypothetical accident. Due to the core shape, it is possible to install heat-exchanging components near the core, and to place the safety rods where they can be very effective in reflector borings. (orig./HP) [de

Entropy Generation of Shell and Double Concentric Tubes Heat Exchanger

Directory of Open Access Journals (Sweden)

basma abbas abdulmajeed

2016-06-01

Full Text Available Entropy generation was studied for new type of heat exchanger (shell and double concentric tubes heat exchanger. Parameters of hot oil flow rate, temperature of inlet hot oil and pressure drop were investigated with the concept of entropy generation. The results showed that the value of entropy generation increased with increasing the flow rate of hot oil and when cold water flow rate was doubled from 20 to 40 l/min, these values were larger. On the other hand, entropy generation increased with increasing the hot oil inlet temperature at a certain flow rate of hot oil. Furthermore, at a certain hot oil inlet temperature, the entropy generation increased with the pressure drop at different hot oil inlet flow rates. Finally, in order to keep up with modern technology, infrared thermography camera was used in order to measure the temperatures. The entropy generation was determined with lower values when infrared thermography camera was used to measure the temperatures, compared with the values obtained by using thermocouples.

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

Energy Technology Data Exchange (ETDEWEB)

Aghoyeh, Reza Gholizadeh [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hosein, E-mail: hkhalafi@aeoi.org.i [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

2010-10-15

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

International Nuclear Information System (INIS)

Aghoyeh, Reza Gholizadeh; Khalafi, Hosein

2010-01-01

Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

Potential for increased wind-generated electricity utilization using heat pumps in urban areas

International Nuclear Information System (INIS)

Waite, Michael; Modi, Vijay

2014-01-01

Highlights: • Large-scale wind power and increased electric heat pumps were evaluated. • A deterministic model of wind power and electricity demand was developed. • Sub-models for space heating and domestic hot water demand were developed. • Increased use of heat pumps can improve the viability of large-scale wind power. • Larger wind power capacity can meet a target utilization rate with more heat pumps. - Abstract: The U.S. has substantial wind power potential, but given wind’s intermittent availability and misalignment with electricity demand profiles, large-scale deployment of wind turbines could result in high electricity costs due to energy storage requirements or low utilization rates. While fuel switching and heat pumps have been proposed as greenhouse gas (GHG) emissions and energy reduction strategies at the building scale, this paper shows that heat pump adoption could have additional system-wide benefits by increasing the utilization of wind-generated electricity. A model was developed to evaluate the effects of coupling large-scale wind power installations in New York State with increased use of electric heat pumps to meet a portion of space heating and domestic hot water (DHW) demands in New York City. The analysis showed significant increases in wind-generated electricity utilization with increased use of heat pumps, allowing for higher installed capacity of wind power. One scenario indicates that 78.5% annual wind-generated electricity utilization can be achieved with 3 GW of installed wind power capacity generated electricity equal to 20% of existing NYC annual electricity demand; if 20% of space heating and DHW demands are provided by heat pumps, the 78.5% utilization rate can be achieved with an increase of total wind power capacity to 5 GW. Therefore, this integrated supply–demand approach could provide additional system-wide emissions reductions

Nuclear-enhanced geothermal heat recovery

International Nuclear Information System (INIS)

Clark, W.H. II

1995-01-01

This report proposes the testing of an abandoned drill well for the disposal of spent nuclear fuel rods. The well need not be in a geothermal field, since the downhole assembly takes advantage of only the natural thermal gradient. The water in the immediate vicinity of the fuel will be chemically treated for corrosion resistance. Above this will be a long column of viscous fluid insoluble in water, to act as a fluid barrier. The remainder of the well bore, up to the surface, will be the working fluid for the power turbine at the surface. There will be a low-pressure region in the immediate vicinity of the fuel, encouraging the flashing of steam. Due to the low level of heat emitted by the fuel rods, the radioactive material will be surrounded by a secondary casing that will reduce the water it contacts directly, thus causing it to heat up quickly and to maximize the steam-generating process, and the formation of air nuclides. These will percolate upward through the viscous column where steadily decreasing pressure causes expansion. The nuclear fuel's thermal energy will have been transferred through the high radioactive zone as pressure, then it will flash to steam and heat the water in the top of the wellbore. The drill well, a minimum of 10,000 ft. in depth, will naturally heat any circulating fluid. The fuel is not used as a thermal source, but only to produce a few spontaneous bubbles, sufficient to increase the fluid pressure by expansion as it rises in the wellbore. The additional thermal energy from the nuclear source will superheat the water for use in the power-generation apparatus at the surface. This equipment, operating on very-low radioactive fluid, will be protected by a secondary containment. The typical drill well is ideally suited for the insertion of spent fuel rods, which are smaller than downhole tools and instrumentation regularly installed in production wells

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

International Nuclear Information System (INIS)

Sabourin, G.

1998-01-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Spent fuel management in Spain

International Nuclear Information System (INIS)

Gago, J.A.; Gravalos, J.M.

1996-01-01

There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

Spent fuel storage requirements 1989--2020

International Nuclear Information System (INIS)

1989-10-01

Historical inventories of spent fuel are combined with Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the US to provide estimates of spent fuel storage requirements over the next 32 years, through the year 2020. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Historical data through December 1988 are derived from the 1989 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 14 refs., 3 figs., 28 tabs

Feasibility of deep ocean disposal of heat generating waste. V.1

International Nuclear Information System (INIS)

Hemming, C.R.

1988-06-01

This report summarises the research performed in the UK during the period 1977 to 1987 as part of the international programme investigating the feasibility of ocean disposal of heat generating radioactive waste. This study has involved: (i) the definition of the disposal operations needed to meet the minimum requirements for safely emplacing waste on or under the floor of the deep ocean; (ii) the identification and characterisation of areas of the deep ocean that might be suitable for containing heat generating waste; (iii) a study of the processes by which radionuclides might migrate through the multiple barriers that isolate the waste from man's environment; and (iv) a calculation of the radiological impact of the conceptual deep ocean repository. It is concluded that, from a technical and scientific viewpoint, disposal of heat generating waste in the deep ocean could provide a safe, economic and feasible alternative to deep disposal on land. (author)

Study on heat pipe assisted thermoelectric power generation system from exhaust gas

Science.gov (United States)

Chi, Ri-Guang; Park, Jong-Chan; Rhi, Seok-Ho; Lee, Kye-Bock

2017-11-01

Currently, most fuel consumed by vehicles is released to the environment as thermal energy through the exhaust pipe. Environmentally friendly vehicle technology needs new methods to increase the recycling efficiency of waste exhaust thermal energy. The present study investigated how to improve the maximum power output of a TEG (Thermoelectric generator) system assisted with a heat pipe. Conventionally, the driving energy efficiency of an internal combustion engine is approximately less than 35%. TEG with Seebeck elements is a new idea for recycling waste exhaust heat energy. The TEG system can efficiently utilize low temperature waste heat, such as industrial waste heat and solar energy. In addition, the heat pipe can transfer heat from the automobile's exhaust gas to a TEG. To improve the efficiency of the thermal power generation system with a heat pipe, effects of various parameters, such as inclination angle, charged amount of the heat pipe, condenser temperature, and size of the TEM (thermoelectric element), were investigated. Experimental studies, CFD simulation, and the theoretical approach to thermoelectric modules were carried out, and the TEG system with heat pipe (15-20% charged, 20°-30° inclined configuration) showed the best performance.

Simplified risk assessment for transporting ATR spent fuel within the INEL

International Nuclear Information System (INIS)

Franklin, E.M.; Courtney, J.C.

1994-01-01

Interest in characterizing the condition of stored spent fuels has generated the need to move spent fuels to hot cell facilities within the Idaho National Engineering Laboratory (INEL). A simplified probabilistic risk assessment (SPRA) and an evaluation of the radiological consequences in the event of an accident are discussed and applied to on-site Advanced Test Reactor (AYR) spent fuel shipments. Reported accident probabilities between 10 -4 and 10 -6 and low radiological consequences, affords this, and other spent fuel characterization efforts, an additional option to move spent fuels within the INEL

French nuclear power plants for heat generation

International Nuclear Information System (INIS)

Girard, Y.

1984-01-01

The considerable importance that France attributes to nuclear energy is well known even though as a result of the economic crisis and the energy savings it is possible to observe a certain downward trend in the rate at which new power plants are being started up. In July 1983, a symbolic turning-point was reached - at more than 10 thousand million kW.h nuclear power accounted, for the first time, for more than 50% of the total amount of electricity generated, or approx. 80% of the total electricity output of thermal origin. On the other hand, the direct contribution - excluding the use of electricity - of nuclear energy to the heat market in France remains virtually nil. The first part of this paper discusses the prospects and realities of the application, at low and intermediate temperatures, of nuclear heat in France, while the second part describes the French nuclear projects best suited to the heat market (excluding high temperatures). (author)

Determination of thermal characteristics of combustion products of fire-tube heat generator with flow turbulator

Directory of Open Access Journals (Sweden)

Lukjanov Alexander V.

2014-12-01

Full Text Available Boiler construction is one of the major industries of any state. The aim is to determine the effect of the turbulator on the intensity of heat transfer in the convective part of the fire-tube heat generator of domestic production. The improvement of convective heating surfaces is one of the ways to increase the energy efficiency of the fire-tube heat generator. Since model of the process of heat transfer of gas flow in the convective tubes is multifactorial and does not have clear analytical solution at present, the study of process above is carried out using the experimental method. The results of applying the flow turbulator as a broken tape in the fire-tube heat generator of KV-GM type are presented. On their basis it can be concluded about increasing of heat transfer in convective part of the unit. The use of efficient, reliable, easy to manufacture, relatively inexpensive turbulator in domestic fire-tube heat generators will allow to increase their energy conversion efficiency and reduce fuel consumption, which will have a positive economic effect.

Development of high-strength aluminum alloys for basket in transport and storage cask for high burn-up spent fuel

International Nuclear Information System (INIS)

Maeguchi, T.; Sakaguchi, Y.; Kamiwaki, Y.; Ishii, M.; Yamamoto, T.

2004-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has developed high-strength borated aluminum alloys (high-strength B-Al alloys), suitable for application to baskets in transport and storage casks for high burn-up spent fuels. Aluminum is a suitable base material for the baskets due to its low density and high thermal conductivity. The aluminum basket would reduce weight of the cask, and effectively release heat generated by spent fuels. MHI had already developed borated aluminum alloys (high-toughness B-Al alloy), and registered them as ASME Code Case ''N-673''. However, there has been a strong demand for basket materials with higher strength in the case of MSF (Mitsubishi Spent Fuel) casks for high-burn up spent fuels, since the basket is required to stand up to higher stress at higher temperature. The high-strength basket material enables the design of a compact cask under a limitation of total size and weight. MHI has developed novel high-strength B-Al alloys which meet these requirements, based on a new manufacturing process. The outline of mechanical and metallurgical characteristics of the high-strength B-Al alloys is described in this paper

Entropy Generation Analysis of Natural Convection in Square Enclosures with Two Isoflux Heat Sources

Directory of Open Access Journals (Sweden)

S. Z. Nejad

2017-04-01

Full Text Available This study investigates entropy generation resulting from natural convective heat transfer in square enclosures with local heating of the bottom and symmetrical cooling of the sidewalls. This analysis tends to optimize heat transfer of two pieces of semiconductor in a square electronic package. In this simulation, heaters are modeled as isoflux heat sources and sidewalls of the enclosure are isothermal heat sinks. The top wall and the non-heated portions of the bottom wall are adiabatic. Flow and temperature fields are obtained by numerical simulation of conservation equations of mass, momentum and energy in laminar, steady and two dimensional flows. With constant heat energy into the cavity, effect of Rayleigh number, heater length, heater strength ratios and heater position is evaluated on flow and temperature fields and local entropy generation. The results show that a minimum entropy generation rate is obtained under the same condition in which a minimum peak heater temperature is obtained.

A thermoelectric power generating heat exchanger: Part I – Experimental realization

DEFF Research Database (Denmark)

Bjørk, Rasmus; Sarhadi, Ali; Pryds, Nini

2016-01-01

An experimental realization of a heat exchanger with commercial thermoelectric generators (TEGs) is presented. The power producing capabilities as a function of flow rate and temperature span are characterized for two different commercial heat transfer fluids and for three different thermal...

ULF Generation by Modulated Ionospheric Heating

Science.gov (United States)

Chang, C.; Labenski, J.; Wallace, T.; Papadopoulos, K.

2013-12-01

Modulated ionospheric heating experiments designed to generate ULF waves using the HAARP heater have been conducted since 2007. Artificial ULF waves in the Pc1 frequency range were observed from space and by ground induction magnetometers located in the vicinity of the heater as well as at long distances. Two distinct generation mechanisms of artificial ULF waves were identified. The first was electroject modulation under geomagnetically disturbed conditions. The second was pressure modulation in the E and F regions of the ionosphere under quiet conditions. Ground detections of ULF waves near the heater included both Shear Alfven waves and Magnetosonic waves generated by electrojet and/or pressure modulations. Distant ULF detections involved Magnetosonic wave propagation in the Alfvenic duct with pressure modulation as the most likely source. Summary of our observations and theoretical interpretations will be presented at the meeting. We would like to acknowledge the support provided by the staff at the HAARP facility during our ULF experiments.

Forecast of power generation and heat production from renewable energy sources

Directory of Open Access Journals (Sweden)

Pydych Tadeusz

2017-01-01

Full Text Available The share of renewable energy sources (RES in the end use of energy in the UE will increase from the present level of about 25% to 50 % in 2030 according to the assumptions of the European Commission. In Poland the RES Act was passed in 2015. The act defines mechanisms and instruments for supporting the production of electricity and heat from renewable energy sources. Statistics (2003–2014 of electricity generation and heat production from RES in Poland were used in the research. Because of amendments to regulations connected with promoting RES and the emissions trading system (ETS as well as the uncertainty associated with further directions of the energy and environmental policy, generation of electricity and heat based on the use of RES must be modelled while taking risk into account. A number of dynamic processes incorporating random events may be modelled by stochastic equations using Ito calculus. By applying Euler’s method to solve stochastic differential equations (SDE, it is possible to simulate the development of the use of renewable energy carriers in electricity generation and heat production in the future.

Nanophotonic-Engineered Photothermal Harnessing for Waste Heat Management and Pyroelectric Generation.

Science.gov (United States)

Wang, Xiao-Qiao; Tan, Chuan Fu; Chan, Kwok Hoe; Xu, Kaichen; Hong, Minghui; Kim, Sang-Woo; Ho, Ghim Wei

2017-10-24

At present, there are various limitations to harvesting ambient waste heat which include the lack of economically viable material and innovative design features that can efficiently recover low grade heat for useful energy conversion. In this work, a thermal nanophotonic-pyroelectric (TNPh-pyro) scheme consisting of a metamaterial multilayer and pyroelectric material, which performs synergistic waste heat rejection and photothermal heat-to-electricity conversion, is presented. Unlike any other pyroelectric configuration, this conceptual design deviates from the conventional by deliberately employing back-reflecting NIR to enable waste heat reutilization/recuperation to enhance pyroelectric generation, avoiding excessive solar heat uptake and also retaining high visual transparency of the device. Passive solar reflective cooling up to 4.1 °C is demonstrated. Meanwhile, the photothermal pyroelectric performance capitalizing on the back-reflecting effect shows an open circuit voltage (V oc ) and short circuit current (I sc ) enhancement of 152 and 146%, respectively. In addition, the designed photoactive component (TiO 2 /Cu) within the metamaterial multilayer provides the TNPh-pyro system with an effective air pollutant photodegradation functionality. Finally, proof-of-concept for concurrent photothermal management and enhanced solar pyroelectric generation under a real outdoor environment is demonstrated.

A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

International Nuclear Information System (INIS)

Kim, Chang Hyun

1997-02-01

A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

District heating and combined heat and power generation from biomass

International Nuclear Information System (INIS)

Veski, Rein

1999-01-01

An Altener programme seminar District Heating and Combined Heat and Power Generation from Biomass. Minitraining seminar and study tours and also Business forum, Exhibition and Short company presentations were held in Tallinn on March 21-23, 1999. The Seminar was organised by the VTT Energy, the Estonian Bioenergy Association and the Estonian Heat and Power Association in co-operation with the AFB-net. The Agricultural and Forestry Biomass Network (AFB-net) is part of the ALTENER programme. The Network aims at promoting and stimulating the implementation and commercial utilisation of energy from biomass and waste, through the initiation of business opportunities. This includes national and international co-operation and the exchange of the personnel. The Seminar was attended by consulting companies, scientists, municipal authorities and representatives of co-ordinating bodies engaged in renewable energy management as well as DH and CHP plant managers, equipment manufacturers and local energy planners from Finland, Estonia, Latvia, Lithuania, Sweden, Denmark, Belgium, Slovenia and Slovak Republic. At the Seminar minitraining issues were dealt with: the current situation and future trends in biomass DH in the Baltic Sea countries, and biomass DH and CHP in Eastern and Central Europe, planning and construction of biomass-based DH plants, biomass fuel procurement and handling technology, combustion technology, DH networks, financing of biomass projects and evaluating of projects, and case projects in Eastern and Central European countries. The following were presented: boilers with a capacity of 100 kW or more, stoker burners, wood and straw handling equipment, wood fuel harvesters, choppers, pelletisers, district heating pipelines and networks. (author)

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1988-09-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97/degree/C and whether the cladding of the stored spent fuel ever exceeds 350/degree/C. Limiting the borehole to temperatures of 97/degree/C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350/degree/C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97/degree/C for the full 1000-yr analysis period

Development of Thermoelectric Power Generators for high temperature Waste Heat Recovery

DEFF Research Database (Denmark)

Van Nong, Ngo; Pryds, Nini

By converting heat directly into electricity, thermoclectric generators (TEGs) provide a very promising solution for emerging energy saving and environmental issues. These devices could be incorporated in a variety of applications, in particular those making use of waste heat recovery. To expand...

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

International Nuclear Information System (INIS)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

2004-01-01

We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)

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

Entropy Generation of Desalination Powered by Variable Temperature Waste Heat

Directory of Open Access Journals (Sweden)

David M. Warsinger

2015-10-01

Full Text Available Powering desalination by waste heat is often proposed to mitigate energy consumption and environmental impact; however, thorough technology comparisons are lacking in the literature. This work numerically models the efficiency of six representative desalination technologies powered by waste heat at 50, 70, 90, and 120 °C, where applicable. Entropy generation and Second Law efficiency analysis are applied for the systems and their components. The technologies considered are thermal desalination by multistage flash (MSF, multiple effect distillation (MED, multistage vacuum membrane distillation (MSVMD, humidification-dehumidification (HDH, and organic Rankine cycles (ORCs paired with mechanical technologies of reverse osmosis (RO and mechanical vapor compression (MVC. The most efficient technology was RO, followed by MED. Performances among MSF, MSVMD, and MVC were similar but the relative performance varied with waste heat temperature or system size. Entropy generation in thermal technologies increases at lower waste heat temperatures largely in the feed or brine portions of the various heat exchangers used. This occurs largely because lower temperatures reduce recovery, increasing the relative flow rates of feed and brine. However, HDH (without extractions had the reverse trend, only being competitive at lower temperatures. For the mechanical technologies, the energy efficiency only varies with temperature because of the significant losses from the ORC.

Heat transfer study on convective–radiative semi-spherical fins with temperature-dependent properties and heat generation using efficient computational methods

International Nuclear Information System (INIS)

Atouei, S.A.; Hosseinzadeh, Kh.; Hatami, M.; Ghasemi, Seiyed E.; Sahebi, S.A.R.; Ganji, D.D.

2015-01-01

In this study, heat transfer and temperature distribution equations for semi-spherical convective–radiative porous fins are presented. Temperature-dependent heat generation, convection and radiation effects are considered and after deriving the governing equation, Least Square Method (LSM), Collocation Method (CM) and fourth order Runge-Kutta method (NUM) are applied for predicting the temperature distribution in the described fins. Results reveal that LSM has excellent agreement with numerical method, so can be suitable analytical method for solving the problem. Also, the effect of some physical parameters which are appeared in the mathematical formulation on fin surface temperature is investigated to show the effect of radiation and heat generation in a solid fin temperature. - Highlights: • Thermal analysis of a semi-spherical fin is investigated. • Collocation and Least Square Methods are applied on the problem. • Convection, radiation and heat generation is considered. • Physical results are compared to numerical outcomes.

Entropy generation of nanofluid flow in a microchannel heat sink

Science.gov (United States)

Manay, Eyuphan; Akyürek, Eda Feyza; Sahin, Bayram

2018-06-01

Present study aims to investigate the effects of the presence of nano sized TiO2 particles in the base fluid on entropy generation rate in a microchannel heat sink. Pure water was chosen as base fluid, and TiO2 particles were suspended into the pure water in five different particle volume fractions of 0.25%, 0.5%, 1.0%, 1.5% and 2.0%. Under laminar, steady state flow and constant heat flux boundary conditions, thermal, frictional, total entropy generation rates and entropy generation number ratios of nanofluids were experimentally analyzed in microchannel flow for different channel heights of 200 μm, 300 μm, 400 μm and 500 μm. It was observed that frictional and total entropy generation rates increased as thermal entropy generation rate were decreasing with an increase in particle volume fraction. In microchannel flows, thermal entropy generation could be neglected due to its too low rate smaller than 1.10e-07 in total entropy generation. Higher channel heights caused higher thermal entropy generation rates, and increasing channel height yielded an increase from 30% to 52% in thermal entropy generation. When channel height decreased, an increase of 66%-98% in frictional entropy generation was obtained. Adding TiO2 nanoparticles into the base fluid caused thermal entropy generation to decrease about 1.8%-32.4%, frictional entropy generation to increase about 3.3%-21.6%.

Distillery spent wash: Treatment technologies and potential applications

International Nuclear Information System (INIS)

Mohana, Sarayu; Acharya, Bhavik K.; Madamwar, Datta

2009-01-01

Distillery spent wash is the unwanted residual liquid waste generated during alcohol production and pollution caused by it is one of the most critical environmental issue. Despite standards imposed on effluent quality, untreated or partially treated effluent very often finds access to watercourses. The distillery wastewater with its characteristic unpleasant odor poses a serious threat to the water quality in several regions around the globe. The ever-increasing generation of distillery spent wash on the one hand and stringent legislative regulations of its disposal on the other has stimulated the need for developing new technologies to process this effluent efficiently and economically. A number of clean up technologies have been put into practice and novel bioremediation approaches for treatment of distillery spent wash are being worked out. Potential microbial (anaerobic and aerobic) as well as physicochemical processes as feasible remediation technologies to combat environmental pollution are being explored. An emerging field in distillery waste management is exploiting its nutritive potential for production of various high value compounds. This review presents an overview of the pollution problems caused by distillery spent wash, the technologies employed globally for its treatment and its alternative use in various biotechnological sectors

Thermal Analysis Evaluation of Spent Fuel Storage Rack for Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangjin; Oh, Jinho; Kwak, Jinsung; Lee, Jongmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Spent fuel storage rack is to store spent fuel assemblies. The spent fuel storage rack is submerged in the designated pool for cooling. Due to the condition change of the pool water, the effect of thermal load on spent fuel storage rack must be analyzed and evaluated. In this paper, thermal stress analysis is performed and evaluated on a spent fuel storage rack. For thermal stress evaluation of the spent fuel storage rack, load combinations and allowable criteria in ASME Sec. III NB-3220 are applied. In cases of A-1 and B-1, the same temperature applied on the whole model, thermal stress doesn't occur because there is no constraint about the thermal expansion. The support frame is located on the pool bottom in free standing type and the racks are located in the support frame with enough space. Thermal expansion was considered and reflected in the design of spent fuel storage rack in advance. Thermal stress analysis is performed and evaluated on a spent fuel storage rack with consideration of pool water temperature variation. The thermal analysis including a linear heat transfer and the thermal stress analysis is performed for the racks and support frame and resulted stresses are within allowable criteria.

Random generation of bubble sizes on the heated wall during subcooled boiling

International Nuclear Information System (INIS)

Koncar, B.; Mavko, B.

2003-01-01

In subcooled flow boiling, a locally averaged bubble diameter significantly varies in the transverse direction to the flow. From the experimental data of Bartel, a bent crosssectional profile of local bubble diameter with the maximum value shifted away from the heated wall may be observed. In the present paper, the increasing part of the profile (near the heated wall) is explained by a random generation of bubble sizes on the heated wall. The hypothesis was supported by a statistical analysis of different CFD simulations, varying by the size of the generated bubble (normal distribution) and the number of generated bubbles per unit surface. Local averaging of calculated void fraction distributions over different bubble classes was performed. The increasing curve of the locally averaged bubble diameter in the near-wall region was successfully predicted. (author)

Experiments and simulations on heat exchangers in thermoelectric generator for automotive application

International Nuclear Information System (INIS)

Liu, X.; Deng, Y.D.; Zhang, K.; Xu, M.; Xu, Y.; Su, C.Q.

2014-01-01

In this work, an energy-harvesting system which extracts heat from an automotive exhaust pipe and turns the heat into electricity by using thermoelectric power generators (TEGs) was built. Experiments show that the temperature difference in automotive system is not constant, especially the heat exchanger, which cannot provide the thermoelectric modules (TMs) large amount of heat. The thermal performance of different heat exchangers in exhaust-based TEGs is studied in this work, and the thermal characteristics of heat exchangers with different internal structures and thickness are discussed, to obtain higher interface temperature and thermal uniformity. Following computational fluid dynamics simulations, infrared experiments and output power testing system are carried out on a high-performance production engine with a dynamometer. Results show that a plate-shaped heat exchanger with chaos-shaped internal structure and thickness of 5 mm achieves a relatively ideal thermal performance, which is practically useful to enhance the thermal performance of the TEG, and larger total output power can be thus obtained. - Graphical abstract: The thermal and electrical characteristics of different heat exchangers of automotive exhaust-based thermoelectric generator are discussed, to obtain higher interface temperature and thermal uniformity. - Highlights: • Different internal structures and thickness of heat exchangers were proposed. • Power output testing system of the two heat exchangers was characterized. • Chaos-shaped heat exchanger (5 mm thickness) shows better performance

Effects of the generator and evaporator temperature differences on a double absorption heat transformer—Different control strategies on utilizing heat sources

International Nuclear Information System (INIS)

Wang, Hanzhi; Li, Huashan; Bu, Xianbiao; Wang, Lingbao

2017-01-01

Highlights: • Effects of the GETD on the DAHT system performance are analyzed. • Three different configurations are compared in detail. • Suggestions on the heat source control strategies are given. - Abstract: The combination of the absorption heat transformer with renewable energy systems, like solar thermal systems, is raising more and more concern. In those combined systems the strategies on utilizing heat sources can affect system thermodynamic performance significantly. Therefore, this study presents a detailed analysis on the effect of the heat source temperature and different heat source flow patterns on the performance of a double absorption heat transformer (DAHT). A detailed comparative study is carried out to clarify the impact of the generator and evaporator temperature differences (GETD) on the coefficient of performance (COP), exergy efficient (ECOP), exergy destruction rates in the individual components and heat transfer areas needed for each component. The results show that the generator, condenser and absorber-evaporator are responsible for most of the exergy destruction rate in the DAHT system; the parallel-flow configuration (the generator temperature is equal to the evaporator temperature) performs better under the high gross temperature lift conditions; in the case of the counter-flow configuration (the generator temperature is relatively higher), better performance can be obtained in both the COP and ECOP under the proper heat source temperature (85 and 95 °C); the fair-flow configuration (higher temperature in the evaporator) is not recommended in this paper due to no advantages found in either thermodynamic performance or system size.

Safety Analysis of Spent Nuclear Fuel and Radwaste Facilities

International Nuclear Information System (INIS)

Poskas, P.; Ragaisis, V.

2001-01-01

The overview of the activities in the Laboratory of Heat Transfer in Nuclear Reactors related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Activities related with decommissioning of Ignalina NPP are also reviewed. (author)

Spent fuel storage requirements, 1991--2040

International Nuclear Information System (INIS)

1991-12-01

Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 50 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1990 are derived from the 1991 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges

Power generation and heating performances of integrated system of ammonia–water Kalina–Rankine cycle

International Nuclear Information System (INIS)

Zhang, Zhi; Guo, Zhanwei; Chen, Yaping; Wu, Jiafeng; Hua, Junye

2015-01-01

Highlights: • Integrated system of ammonia–water Kalina–Rankine cycle (AWKRC) is investigated. • Ammonia–water Rankine cycle is operated for cogenerating room heating-water in winter. • Kalina cycle with higher efficiency is operated for power generation in other seasons. • Power recovery efficiency accounts thermal efficiency and waste heat absorbing ratio. • Heating water with 70 °C and capacity of 55% total reclaimed heat load is cogenerated. - Abstract: An integrated system of ammonia–water Kalina–Rankine cycle (AWKRC) for power generation and heating is introduced. The Kalina cycle has large temperature difference during evaporation and small one during condensation therefore with high thermal efficiency for power generation, while the ammonia–water Rankine cycle has large temperature difference during condensation as well as evaporation, thus it can be adopted to generate heating-water as a by-product in winter. The integrated system is based on the Kalina cycle and converted to the Rankine cycle with a set of valves. The performances of the AWKRC system in different seasons with corresponding cycle loops were studied and analyzed. When the temperatures of waste heat and cooling water are 300 °C and 25 °C respectively, the thermal efficiency and power recovery efficiency of Kalina cycle are 20.9% and 17.4% respectively in the non-heating seasons, while these efficiencies of the ammonia–water Rankine cycle are 17.1% and 13.1% respectively with additional 55.3% heating recovery ratio or with comprehensive efficiency 23.7% higher than that of the Kalina cycle in heating season

Proceedings of the third international steam generator and heat exchanger conference

International Nuclear Information System (INIS)

1998-01-01

The Third International Steam Generator and Heat Exchanger conference had the objective to present the state of knowledge of steam generator performance and life management, and also heat exchanger technology. As this conference followed on from the previous conferences held in Toronto in 1990 and 1994, the emphasis was on recent developments, particularly those of the last 4 years. The conference provided an opportunity to operators, designers and researchers in the field of steam generation associated with electricity generation by nuclear energy to present their findings and exchange ideas. The conference endeavoured to do this over the widest possible range of subject areas, including: general operating experience, life management and fitness for service strategies, maintenance and inspection, thermalhydraulics, vibration, fretting and fatigue, materials, chemistry and corrosion and the regulatory issues

Proceedings of the third international steam generator and heat exchanger conference

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-07-01

The Third International Steam Generator and Heat Exchanger conference had the objective to present the state of knowledge of steam generator performance and life management, and also heat exchanger technology. As this conference followed on from the previous conferences held in Toronto in 1990 and 1994, the emphasis was on recent developments, particularly those of the last 4 years. The conference provided an opportunity to operators, designers and researchers in the field of steam generation associated with electricity generation by nuclear energy to present their findings and exchange ideas. The conference endeavoured to do this over the widest possible range of subject areas,including: general operating experience, life management and fitness for service strategies, maintenance and inspection, thermalhydraulics, vibration, fretting and fatigue, materials, chemistry and corrosion and the regulatory issues.

Spent fuel behaviour during dry storage - a review

International Nuclear Information System (INIS)

Shivakumar, V.; Anantharaman, K.

1997-09-01

One of the strategies employed for management of spent fuel prior to their final disposal/reprocessing is their dry storage in casks, after they have been sufficiently cooled in spent fuel pools. In this interim storage, one of the main consideration is that the fuel should retain its integrity to ensure (a) radiological health hazard remains minimal and (b) the fuel is retrievable for down steam fuel management processes such as geological disposal or reprocessing. For dry storage of spent fuel in air, oxidation of the exposed UO 2 is the most severe of phenomena affecting the integrity of fuel. This is kept within acceptable limits for desired storage time by limiting the fuel temperature in the storage cask. The limit on the fuel temperature is met by having suitable limits on maximum burn-up of fuel, minimum cooling period in storage pool and optimum arrangement of fuel bundles in the storage cask from heat removal considerations. The oxidation of UO 2 by moist air has more deleterious effects on the integrity of fuel than that by dry air. The removal of moisture from the storage cask is therefore a very important aspect in dry storage practice. The kinetics of the oxidation phenomena at temperatures expected during dry storage in air is very slow and therefore the majority of the existing data is based on extrapolation of data obtained at higher fuel temperatures. This and the complex effects of factors like fission products in fuel, radiolysis of storage medium etc. has necessitated in having a conservative limiting criteria. The data generated by various experimental programmes and results from the on going programmes have shown that dry storage is a safe and economical practice. (author)

Heat savings and heat generation technologies: Modelling of residential investment behaviour with local health costs

International Nuclear Information System (INIS)

Zvingilaite, Erika; Klinge Jacobsen, Henrik

2015-01-01

The trade-off between investing in energy savings and investing in individual heating technologies with high investment and low variable costs in single family houses is modelled for a number of building and consumer categories in Denmark. For each group the private economic cost of providing heating comfort is minimised. The private solution may deviate from the socio-economical optimal solution and we suggest changes to policy to incentivise the individuals to make choices more in line with the socio-economic optimal mix of energy savings and technologies. The households can combine their primary heating source with secondary heating e.g. a woodstove. This choice results in increased indoor air pollution with fine particles causing health effects. We integrate health cost due to use of woodstoves into household optimisation of heating expenditures. The results show that due to a combination of low costs of primary fuel and low environmental performance of woodstoves today, included health costs lead to decreased use of secondary heating. Overall the interdependence of heat generation technology- and heat saving-choice is significant. The total optimal level of heat savings for private consumers decrease by 66% when all have the option to shift to the technology with lowest variable costs. - Highlights: • Heat saving investment and heat technology choice are interdependent. • Health damage costs should be included in private heating choice optimisation. • Flexibility in heating technology choice reduce the optimal level of saving investments. • Models of private and socioeconomic optimal heating produce different technology mix. • Rebound effects are moderate but varies greatly among consumer categories

Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

International Nuclear Information System (INIS)

Johnson, G.L.

1991-11-01

Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. On such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97 degrees C and whether the cladding of the stored spent fuel ever exceeds 350 degrees C. Limiting the borehole to temperatures of 97 degrees C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350 degrees C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97 degrees C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350 degrees C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft x 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40 degrees C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation

CFD analysis of fin tube heat exchanger with a pair of delta winglet vortex generators

International Nuclear Information System (INIS)

Hwang, Seong Won; Kim, Dong Hwan; Min, June Kee; Jeong, Ji Hwan

2012-01-01

Among tubular heat exchangers, fin tube types are the most widely used in refrigeration and air-conditioning equipment. Efforts to enhance the performance of these heat exchangers included variations in the fin shape from a plain fin to a slit and louver type. In the context of heat transfer augmentation, the performance of vortex generators has also been investigated. Delta winglet vortex generators have recently attracted research interest, partly due to experimental data showing that their addition to fin-tube heat exchangers considerably reduces pressure loss at heat transfer capacity of nearly the same level. The efficiency of the delta winglet vortex generators widely varies depending on their size and shape, as well as the locations where they are implemented. In this paper, the flow field around delta winglet vortex generators in a common flow up arrangement was analyzed in terms of flow characteristics and heat transfer using computational fluid dynamics methods. Flow mixing due to vortices and delayed separation due to acceleration influence the overall fin performance. The fin with delta winglet vortex generators exhibited a pressure loss lower than that of a plain fin, and the heat transfer performance was enhanced at high air velocity or Reynolds number

CFD analysis of fin tube heat exchanger with a pair of delta winglet vortex generators

Energy Technology Data Exchange (ETDEWEB)

Hwang, Seong Won; Kim, Dong Hwan; Min, June Kee; Jeong, Ji Hwan [Pusan National Univ., Busan (Korea, Republic of)

2012-09-15

Among tubular heat exchangers, fin tube types are the most widely used in refrigeration and air-conditioning equipment. Efforts to enhance the performance of these heat exchangers included variations in the fin shape from a plain fin to a slit and louver type. In the context of heat transfer augmentation, the performance of vortex generators has also been investigated. Delta winglet vortex generators have recently attracted research interest, partly due to experimental data showing that their addition to fin-tube heat exchangers considerably reduces pressure loss at heat transfer capacity of nearly the same level. The efficiency of the delta winglet vortex generators widely varies depending on their size and shape, as well as the locations where they are implemented. In this paper, the flow field around delta winglet vortex generators in a common flow up arrangement was analyzed in terms of flow characteristics and heat transfer using computational fluid dynamics methods. Flow mixing due to vortices and delayed separation due to acceleration influence the overall fin performance. The fin with delta winglet vortex generators exhibited a pressure loss lower than that of a plain fin, and the heat transfer performance was enhanced at high air velocity or Reynolds number.

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

Milliwatt-generator heat source. Progress report, January-June 1983

International Nuclear Information System (INIS)

Mershad, E.A.

1983-01-01

Progress is reported in the following: heat source shipments, reimbursable orders, hardware shipments, raw material qualification/procurement, DOE audit and milliwatt generator process review, surveillance capsule evaluations, pressure burst testing, and hardware fabrication and quality

Spent fuel storage criticality safety

Energy Technology Data Exchange (ETDEWEB)

Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

1995-10-01

The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs.

Spent fuel storage criticality safety

International Nuclear Information System (INIS)

Amin, E.M.; Elmessiry, A.M.

1995-01-01

The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

Salt disposal of heat-generating nuclear waste

International Nuclear Information System (INIS)

Leigh, Christi D.; Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

Salt disposal of heat-generating nuclear waste.

Energy Technology Data Exchange (ETDEWEB)

Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

2011-01-01

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

Steam generator concept of a small HTR for reheating and for removal of the residual heat

Energy Technology Data Exchange (ETDEWEB)

Singh, J; Barnert, H; Hohn, H; Mondry, M [Institut fuer Reaktorenentwicklung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

1988-07-01

The steam generator of a small HTR is arranged above the core in an in line design of the primary loop, thereby helium flows upwards. Water flows downwards in the steam generator to realize cross flow. To achieve stable evaporation conditions during part load operation it is desired to realize upward evaporation in the steam generator. Moreover if the steam generator is also used as a heat sink for removal of residual heat, this desire of upwards evaporation becomes more imperative. It is possible to realize the design of steam generator with upwards evaporation by arranging a hot gas duct in its central region, so that hot helium can flow upwards through it. Therefore helium enters the steam generator from the top and flows downwards and water upwards. In the presented design, a heat exchanger is arranged in the central region of the steam generator instead of a hot gas duct. Hot helium of 750 deg. C flows upwards in this heat exchanger and thereby cools down to the temperature of about 700 deg. C before it enters the bundle of the steam generator at the top. Through an intermediate loop this heat is transferred outside the primary loop, where in an extra heat exchanger live steam is reheated to improve the thermal efficiency of the plant. This intermediate loop works on the basis of forced convection and transfer about 25 MW for reheating. During the shutdown operation of the reactor, this heat exchanger in the central region of the steam generator serves as a heat sink for removal of the residual heat through natural convection in the primary loop. At the same time it is further possible, that intermediate loop also works on the basis of natural convection, because during shutdown operation only a very small amount of heat has to be removed and moreover the outside heat exchanger can be arranged much higher above the central heat exchanger to get favourable conditions for the natural convection. Some of the highlights of the central heat exchanger are: coaxial

Nonlinear radiative heat flux and heat source/sink on entropy generation minimization rate

Science.gov (United States)

Hayat, T.; Khan, M. Waleed Ahmed; Khan, M. Ijaz; Alsaedi, A.

2018-06-01

Entropy generation minimization in nonlinear radiative mixed convective flow towards a variable thicked surface is addressed. Entropy generation for momentum and temperature is carried out. The source for this flow analysis is stretching velocity of sheet. Transformations are used to reduce system of partial differential equations into ordinary ones. Total entropy generation rate is determined. Series solutions for the zeroth and mth order deformation systems are computed. Domain of convergence for obtained solutions is identified. Velocity, temperature and concentration fields are plotted and interpreted. Entropy equation is studied through nonlinear mixed convection and radiative heat flux. Velocity and temperature gradients are discussed through graphs. Meaningful results are concluded in the final remarks.

Spent fuel management in Japan

International Nuclear Information System (INIS)

Shirahashi, K.; Maeda, M.; Nakai, T.

1996-01-01

Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

Effect of thermal radiation and suction on convective heat transfer of nanofluid along a wedge in the presence of heat generation/absorption

Energy Technology Data Exchange (ETDEWEB)

Kasmani, Ruhaila Md; Bhuvaneswari, M. [Centre for Foundation Studies in Science, University of Malaya, 50603 Kuala Lumpur (Malaysia); Sivasankaran, S.; Siri, Zailan [Institute of Mathematical Science, University of Malaya, 50603 Kuala Lumpur (Malaysia)

2015-10-22

An analysis is presented to find the effects of thermal radiation and heat generation/absorption on convection heat transfer of nanofluid past a wedge in the presence of wall suction. The governing partial differential equations are transformed into a system of ordinary differential equations using similarity transformation. The resulting system is solved numerically using a fourth-order Runge–Kutta method with shooting technique. Numerical computations are carried out for different values of dimensionless parameters to predict the effects of wedge angle, thermophoresis, Brownian motion, heat generation/absorption, thermal radiation and suction. It is found that the temperature increases significantly when the value of the heat generation/absorption parameter increases. But the opposite observation is found for the effect of thermal radiation.

Steam generators and waste heat boilers for process and plant engineers

CERN Document Server

Ganapathy, V

2014-01-01

Incorporates Worked-Out Real-World ProblemsSteam Generators and Waste Heat Boilers: For Process and Plant Engineers focuses on the thermal design and performance aspects of steam generators, HRSGs and fire tube, water tube waste heat boilers including air heaters, and condensing economizers. Over 120 real-life problems are fully worked out which will help plant engineers in evaluating new boilers or making modifications to existing boiler components without assistance from boiler suppliers. The book examines recent trends and developments in boiler design and technology and presents novel idea

Appropriate heat load ratio of generator for different types of air cooled lithium bromide–water double effect absorption chiller

International Nuclear Information System (INIS)

Li, Zeyu; Liu, Jinping

2015-01-01

Highlights: • Effect of heat load ratio of generator on the performance was analyzed. • The performance is sensitive to heat load ratio of generator. • The appropriate heat load ratio of generator for four systems was obtained. • The change of appropriate heat load ratio of generator for four systems was studied. - Abstract: The lower coefficient of performance and higher risk of crystallization in the higher surrounding temperature is the primary disadvantage of air cooled lithium bromide–water double effect absorption chiller. Since the coefficient of performance and risk of crystallization strongly depend on the heat load ratio of generator, the appropriate heat load ratio of generator can improve the performance as the surrounding temperature is higher. The paper mainly deals with the appropriate heat load ratio of generator of air cooled lithium bromide–water double effect absorption chiller. Four type systems named series, pre-parallel, rear parallel and reverse parallel flow configuration were considered. The corresponding parametric model was developed to analyze the comprehensive effect of heat load ratio of generator on the coefficient of performance and risk of crystallization. It was found that the coefficient of performance goes up linearly with the decrease of heat load ratio of generator. Simultaneously, the risk of crystallization also rises slowly at first but increases fast finally. Consequently, the appropriate heat load ratio of generator for the series and pre-parallel flow type systems is suggested to be 0.02 greater than the minimum heat load ratio of generator and that for the rear parallel and reverse parallel flow chillers should be 0.01 higher than the minimum heat load ratio of generator. Besides, the changes of minimum heat load ratio of generator for different type systems with the working condition were analyzed and compared. It was found that the minimum heat load ratio of generator goes up with the increase of

Heat generation and temperature-rise in ordinary concrete due to capture of thermal neutrons

International Nuclear Information System (INIS)

Abdo, E.A.; Amin, E.

1997-01-01

The aim of this work is the evaluation of the heat generation and temperature-rise in local ordinary concrete as a biological shield due to capture of total thermal and reactor thermal neutrons. The total thermal neutron fluxes were measured and calculated. The channel number 2 of the ETRR-1 reactor was used in the measurements as a neutron source. Computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 was used in the calculations. The heat generation and temperature-rise in local ordinary concrete were evaluated and calculated. The results were displayed in curves to show the distribution of thermal neutron fluxes and heat generation as well as temperature-rise with the shield thickness. The results showed that, the heat generation as well as the temperature-rise have their maximum values in the first layers of the shield thickness. 4 figs., 12 refs

Estimation of the Waste Mass from a Pyro-Process of Spent Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo (and others)

2008-04-15

Pyro-Process is now developing to retrieve reusable uranium and TRU, and to reduce the volume of high level waste from a nuclear power plant. In this situation, it is strongly required for the estimation of expected masses and their physical properties of the wastes. In this report, the amount of wastes and their physical properties are presupposed through some assumptions in regard to 10MTHM of Oxide Fuel with 4.5wt% U-235, 45,000 MWD/MTU, and 5yrs cooling. The produced wastes can be divided into three categories such as metal, CWF(Ceramic Waste Form), and VWF(Vitrified Waste Form). The 42 nuclrides in a spent nuclear fuel are distributed into the waste categories on the their physical and thermodynamic properties when they exist in metal, oxide, or chloride forms. The treated atomic groups are Uranium, TRU, Noble metal, Rare earth, Alkali metal, Halogens, and others. The mass of each waste is estimated by the distribution results. The off-gas waste is included into a CWF. The heat generations by the wastes in this Pyro-Process are calculated using a ORIGEN-ARP program. It is possible to estimate the amounts of wastes and their heat generation rates in this Pyro-Process analysis. These information are very helpful to design a waste container and its quantity also can be determined. The number of container and its heat generation rate will be key factor for the construction of interim storage facilities including a underground disposal site.

Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

International Nuclear Information System (INIS)

1979-08-01

This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

Reactor-specific spent fuel discharge projections: 1985 to 2020

International Nuclear Information System (INIS)

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

Reracking Possibilities of the NPP Krsko Spent Fuel Pool

International Nuclear Information System (INIS)

Bace, M.; Pevec, D.; Smuc, T.

1998-01-01

Using the SCALE-4 code package reracking possibilities of the NPP Krsko spent fuel pool were analyzed. Two cases were considered: the first case assuming the 40 years lifetime of the plant, and the second case assuming the 50 years lifetime of the plant. It was shown that it is possible to design the additional racks in free space of the spent fuel pool with the sufficient total capacity to store all the spent fuel generated during the 40 years lifetime of the plant. In the case of 50 years plant lifetime, completely new racks (capacity of 1890 spent fuel assemblies), containing 4mm boral in storage cell walls, were proposed for the NPP Krsko spent fuel pool. The effective multiplication factor of the spent fuel pool fully loaded with new racks containing spent fuel assemblies of initial enrichment 4.3 w/o, burned to 40 GWd/tU and cooled 2 years is lower than the value required by standard. It showed the possibility of the safe disposal of all spent fuel accumulated during more than 50 years lifetime of the plant. (author)

Heat generation and hemolysis at the shaft seal in centrifugal blood pumps.

Science.gov (United States)

Araki, K; Taenaka, Y; Wakisaka, Y; Masuzawa, T; Tatsumi, E; Nakatani, T; Baba, Y; Yagura, A; Eya, K; Toda, K

1995-01-01

The heat and hemolysis around a shaft seal were investigated. Materials were original pumps (Nikkiso HMS-15:N-original, and 3M Delphin:D-original), vane-removed pumps (Nvane(-), Dvane(-)), and a small chamber with a shaft coiled by nichrome wire (mock pump). The original pumps were driven at 500 mmHg and 5 L/min, and vane-removed pumps were driven at the same rotation number. An electrical powers of 0, 0.5, 2, and 10 W was supplied to the mock pumps. In vitro hemolytic testing showed that hemolytic indices were 0.027 g/100 L in N-original, 0.013 in Nvane(-), 0.061 in D-original, and 0.012 in Dvane(-). Measurement of heat with a thermally insulated water chamber showed total heat within the pump of 8.62 and 10.85 W, and heat at the shaft seal of 0.87 and 0.62 W in the Nikkiso and Delphin pumps, respectively. Hemolysis and heat generation of mock pumps remained low. The results indicate that the heat generated around the shaft seal was minimal. Hemolysis at the shaft-seal was considerable but not major. Local heat did not affect hemolysis. It was concluded that the shaft-seal affected hemolysis, not by local heat but friction itself.

Effects of boundary conditions on thermomechanical calculations: Spent fuel test - climax

International Nuclear Information System (INIS)