Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

The mechanism of selective corrugation removal by KOH anisotropic wet etching

International Nuclear Information System (INIS)

Shikida, M; Inagaki, N; Sasaki, H; Amakawa, H; Fukuzawa, K; Sato, K

2010-01-01

The mechanism of selective corrugation removal by anisotropic wet etching—which reduces a periodic corrugation, called 'scalloping', formed on the sidewalls of microstructures by the Bosch process in deep reactive-ion etching (D-RIE)—was investigated. In particular, the corrugation-removal mechanism was analyzed by using the etching rate distribution pattern, and two equations for predicting the corrugation-removal time by the etching were derived. A Si{1 0 0} wafer was first etched by D-RIE at a depth of 29.4 µm (60 cycles) to form the corrugation on the sidewall surface. The height and pitch of the corrugation were 196 and 494 nm, respectively. Selective removal of the corrugation by using 50% KOH (40 °C) was experimentally tried. The corrugation formed on Si{1 0 0} sidewall surfaces was gradually reduced in size as the etching progressed, and it was completely removed after 5 min of etching. Similarly, the corrugation formed on a Si{1 1 0} sidewall surface was also selectively removed by KOH etching (etching time: 3 min). The roughness value of the sidewall surface was reduced from 17.6 nm to a few nanometers by the etching. These results confirm that the corrugation-removal mechanism using anisotropic wet etching can be explained in terms of the distribution pattern of etching rate

Tropical Drosophila ananassae of wet-dry seasons show cross resistance to heat, drought and starvation

Directory of Open Access Journals (Sweden)

Chanderkala Lambhod

2017-11-01

Full Text Available Plastic responses to multiple environmental stressors in wet or dry seasonal populations of tropical Drosophila species have received less attention. We tested plastic effects of heat hardening, acclimation to drought or starvation, and changes in trehalose, proline and body lipids in Drosophila ananassae flies reared under wet or dry season-specific conditions. Wet season flies revealed significant increase in heat knockdown, starvation resistance and body lipids after heat hardening. However, accumulation of proline was observed only after desiccation acclimation of dry season flies while wet season flies elicited no proline but trehalose only. Therefore, drought-induced proline can be a marker metabolite for dry-season flies. Further, partial utilization of proline and trehalose under heat hardening reflects their possible thermoprotective effects. Heat hardening elicited cross-protection to starvation stress. Stressor-specific accumulation or utilization as well as rates of metabolic change for each energy metabolite were significantly higher in wet-season flies than dry-season flies. Energy metabolite changes due to inter-related stressors (heat versus desiccation or starvation resulted in possible maintenance of energetic homeostasis in wet- or dry-season flies. Thus, low or high humidity-induced plastic changes in energy metabolites can provide cross-protection to seasonally varying climatic stressors.

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Study on diverse passive decay heat removal approach

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

One of the most important principles for nuclear safety is the decay heat removal in accidents. Passive decay heat removal systems are extremely helpful to enhance the safety. In currently design of many advanced nuclear reactors, kinds of passive systems are proposed or developed, such as the passive residual heat removal system, passive injection system, passive containment cooling system. These systems provide entire passive heat removal paths from core to ultimate heat sink. Various kinds of passive systems for decay heat removal are summarized; their common features or differences on heat removal paths and design principle are analyzed. It is found that, these passive decay heat removal paths are similarly common on and connected by several basic heat transfer modes and steps. By the combinations or connections of basic modes and steps, new passive decay heat removal approach or diverse system can be proposed. (authors)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Energy and heat balance in wet DCT

Energy Technology Data Exchange (ETDEWEB)

Saxena, Viren; Moser, Alexander; Schaefer, Michael; Ritschel, Michael [BorgWarner Drivetrain Engineering GmbH, Ketsch (Germany)

2012-11-01

Wet clutch systems are well known for their thermal robustness and versatility in a wide range of automotive applications. Conventional automatics have used them for a long time as torque converter lock-up clutches, shift elements and launch clutches. With the development of DCTs, wet clutch technology has evolved in terms of launch and shift performance, controllability, robustness and efficiency. This paper discusses improvements in the wet clutch and their impact on today's vehicle applications in terms of heat and energy management. Thermal robustness is a crucial aspect for an automatic transmission. In addition to the clutch thermal performance, the influence of transmission oil cooler and oil sump warm-up behavior are discussed. Based on our latest development activities, test results and simulations, we shall discuss the latest friction material enhancement and its impact on DCTs in terms of efficiency and performance. Drag loss is a much-discussed topic during the development of wet clutch systems. This paper discusses in detail the cause and break-up of various energy losses in a wet DCT. Efficient energy management strategies for actuation systems, cooling, and lubrication, clutch apply, and pre-selection in modern power trains with engine start / stop are evaluated based on the latest test and simulation results. Finally, the paper summarizes the performance and efficiency optimized moist clutch system. (orig.)

Interaction between liquid droplets and heated surface

Energy Technology Data Exchange (ETDEWEB)

Nigmatulin, B I [Research and Engineering Centre, LWR Nuclear Plants Safety, Elektrogorsk (Russian Federation); Vasiliev, N I [Research and Engineering Centre, LWR Nuclear Plants Safety, Elektrogorsk (Russian Federation); Guguchkin, V V [Research and Engineering Centre, LWR Nuclear Plants Safety, Elektrogorsk (Russian Federation)

1993-06-01

In this paper, experimental methods and investigation results of interaction between droplets of different liquids and a heated surface are presented. Wetted area, contact time period and transition boundary from wetted to non-wetted interaction regimes are experimentally evaluated. A simple connection of the wetted area value and contact time period with the heat removal efficiency is shown. (orig.)

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

Science.gov (United States)

Krasikov, E.

2015-04-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible.

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

International Nuclear Information System (INIS)

Krasikov, E

2015-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation.There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment.The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. (paper)

Comparative study of heat transfer and wetting behaviour of conventional and bioquenchants for industrial heat treatment

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Peter; Prabhu, K. Narayan [Department of Metallurgical and Materials Engineering, National Institute of Technology Karnataka, Surathkal, P.O. Srinivasnagar 575 025 Mangalore, Karnataka State (India)

2008-02-15

An investigation was conducted to study the suitability of vegetable oils as bioquenchants for industrial heat treatment. The study involved the assessment of the severity of quenching and wetting behaviour of conventional and vegetable oil quench media. Quench severities of sunflower, coconut and palm oils were found to be greater than mineral oil. The quench severity of aqueous media is greater than oil media although their wettability is poor as indicated by their higher contact angles. A dimensionless contact angle parameter defined in this work is found to be a better parameter to compare the wetting behaviour with heat transfer. (author)

Enhanced NO{sub x} removal in wet scrubbers using metal chelates. Final report, Volume 2

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-12-01

Successful pilot plant tests of simultaneous removal of SO{sub 2} and NO{sub x} in a wet lime flue gas desulfurization system were concluded in December. The test, at up to 1.5 MW(e) capacity, were conducted by the Cincinnati Gas and Electric Company and Dravo Lime Company for the US Department of Energy at a pilot plant facility at the Miami Fort station of CG&E near Cincinnati, Ohio. The pilot plant scrubbed a slipstream of flue gas from Unit 7 a 530 MW coal-fired electric generating unit. Tests were conducted in three phases between April and December. The technology tested was wet scrubbing with Thiosorbic{reg_sign} magnesium-enhanced lime for SO{sub 2} removal and simultaneous NO scrubbing with ferrous EDTA, a metal chelate. Magnesium-enhanced lime-based wet scrubbing is used at 20 full-scale high-sulfur coal-fired electric generating units with a combined capacity of 8500 MW. Ferrous EDTA reacts with nitric oxide, NO, which comprises about 96% of NO{sub x} from coal-fired boilers. In this report, although not precise, NO and NO{sub x} are used interchangeably. A major objective of the tests was to combine NO{sub x} removal using ferrous EDTA, a developing technology, with SO{sub 2} removal using wet lime FGD, already in wide commercial use. If successful, this could allow wide application of this NO{sub x} removal technology. Volume 2 covers: description and results of NO{sub x} removal tests; and description and results of waste characterization studies.

Dry/wet performance of a plate-fin air-cooled heat exchanger with continuous corrugated fins

International Nuclear Information System (INIS)

Hauser, S.G.; Kreid, D.K.; Johnson, B.M.

1981-01-01

The performance and operating characteristics of a plate-fin heat exchanger in dry/wet or deluge operations was experimentally determined. Development of the deluge heat/mass transfer model continued. The experiments were conducted in a specially-designed wind tunnel at the PNL. Air that was first heated and humidified to specified conditions was circulated at a controlled rate through a 2 ft x 6 ft heat exchanger module. The heat exchanger used in the tests was a wavy surface, plate fin on tube configuration. Hot water was circulated through the tubes at high flow rates to maintain an essentially isothermal condition on the tube side. Deionized water sprayed on the top of the vertically oriented plate fins was collected at the bottom of the core and recirculated. Instrumentation was provided for measurement of flow rates and thermodynamic conditions in the air, in the core circulation water, and in the deluge water. Measurements of the air side pressure drop and heat rejection rate were made as a function of air flow rate, air inlet temperature and humidity, deluge water flow rate, and the core inclination from the vertical. An overall heat transfer coefficient and an effective deluge film convective coefficient was determined. The deluge model, for predicting heat transfer from a wet finned heat exchanger was further developed and refined, and a major extension of the model was formulated that permits simultaneous calculation of both the heat transfer and evaporation rates from the wetted surface. The experiments showed an increase in the heat rejection rate due to wetting, accompanied by a proportional increase in the air side pressure drop. For operation at the same air side pressure drop, the enhancement ratio Q/sub w//Q/sub d/ varied between 2 and 5 for the conditions tested. Thus, the potential enhancement of heat transfer due to wetting can be substantial

Design and modeling of an advanced marine machinery system including waste heat recovery and removal of sulphur oxides

DEFF Research Database (Denmark)

Frimann Nielsen, Rasmus; Haglind, Fredrik; Larsen, Ulrik

2013-01-01

-stroke diesel engine and a conventional waste heat recovery system. The results suggest that an organic Rankine cycle placed after the conventional waste heat recovery system is able to extract the sulphuric acid from the exhaust gas, while at the same time increase power generation from waste heat by 32...... consists of a two-stroke diesel engine, the wet sulphuric process for sulphur removal and an advanced waste heat recovery system including a conventional steam Rankine cycle and an organic Rankine cycle. The results are compared with those of a state-of-the-art machinery system featuring a two...

Position paper -- Waste storage tank heat removal

International Nuclear Information System (INIS)

Stine, M.D.

1995-01-01

The purpose of this paper is to develop and document a position on the heat removal system to be used on the waste storage tanks currently being designed for the Multi-Function Waste Tank Facility (MWTF), project W-236A. The current preliminary design for the waste storage primary tank heat removal system consists of the following subsystems: (1) a once-through dome space ventilation system; (2) a recirculation dome space ventilation system; and (3) an annulus ventilation system. Recently completed and ongoing studies have evaluated alternative heat removal systems in an attempt to reduce system costs and to optimize heat removal capabilities. In addition, a thermal/heat transfer analysis is being performed that will provide assurance that the heat removal systems selected will be capable of removing the total primary tank design heat load of 1.25 MBtu/hr at an allowable operating temperature of 190 F. Although 200 F is the design temperature limit, 190 F has been selected as the maximum allowable operating temperature limit based on instrumentation sensitivity, instrumentation location sensitivity, and other factors. Seven options are discussed and recommendations are made

Study on diverse passive decay heat removal approach and principle

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

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

Passive heat removal from containment

International Nuclear Information System (INIS)

Gou, P.F.; Townsend, H.E.

1990-01-01

This patent describes a heat removal system for removing heat from a containment of a nuclear reactor. It comprises: a sealed suppression chamber in the containment; means for venting steam from the nuclear reactor into the suppression chamber upon occurrence of an event requiring dissipation of heat from the nuclear reactor. The suppression chamber containing a quantity of water; the suppression chamber having a gas-containing space above the water; a heat exchanger disposed within the gas-containing space of the suppression chamber; the heat exchanger including an enclosed structure for holding a heat-exchange fluid; means for metering a supply of heat-exchange fluid to the heat exchanger to maintain a predetermined level thereof in the enclosed structure. The heat-exchange fluid boiling in the heat exchanger in consequence of heat transfer thereto from steam present in the suppression chamber; means for separating a heat-exchange fluid vapor in the heat exchanger from the heat-exchange fluid; and means for discharging the vapor immediately following its separation from heat-exchange fluid directly from the heat exchanger to a location exterior of the containment, whereby heat is discharged from the suppression chamber, and the containment is maintained at a temperature and pressure below its design value

Review of thermo-physical properties, wetting and heat transfer characteristics of nanofluids and their applicability in industrial quench heat treatment.

Science.gov (United States)

Ramesh, Gopalan; Prabhu, Narayan Kotekar

2011-04-14

The success of quenching process during industrial heat treatment mainly depends on the heat transfer characteristics of the quenching medium. In the case of quenching, the scope for redesigning the system or operational parameters for enhancing the heat transfer is very much limited and the emphasis should be on designing quench media with enhanced heat transfer characteristics. Recent studies on nanofluids have shown that these fluids offer improved wetting and heat transfer characteristics. Further water-based nanofluids are environment friendly as compared to mineral oil quench media. These potential advantages have led to the development of nanofluid-based quench media for heat treatment practices. In this article, thermo-physical properties, wetting and boiling heat transfer characteristics of nanofluids are reviewed and discussed. The unique thermal and heat transfer characteristics of nanofluids would be extremely useful for exploiting them as quench media for industrial heat treatment.

Review of thermo-physical properties, wetting and heat transfer characteristics of nanofluids and their applicability in industrial quench heat treatment

Directory of Open Access Journals (Sweden)

Ramesh Gopalan

2011-01-01

Full Text Available Abstract The success of quenching process during industrial heat treatment mainly depends on the heat transfer characteristics of the quenching medium. In the case of quenching, the scope for redesigning the system or operational parameters for enhancing the heat transfer is very much limited and the emphasis should be on designing quench media with enhanced heat transfer characteristics. Recent studies on nanofluids have shown that these fluids offer improved wetting and heat transfer characteristics. Further water-based nanofluids are environment friendly as compared to mineral oil quench media. These potential advantages have led to the development of nanofluid-based quench media for heat treatment practices. In this article, thermo-physical properties, wetting and boiling heat transfer characteristics of nanofluids are reviewed and discussed. The unique thermal and heat transfer characteristics of nanofluids would be extremely useful for exploiting them as quench media for industrial heat treatment.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Design of SMART waste heat removal dry cooling tower using solar energy

International Nuclear Information System (INIS)

Choi, Yong Jae; Jeong, Yong Hoon

2014-01-01

The 85% of cooling system are once-through cooling system and closed cycle wet cooling system. However, many countries are trying to reduce the power plant water requirement due to the water shortage and water pollution. Dry cooling system is investigated for water saving advantage. There are two dry cooling system which are direct and indirect cooling system. In direct type, turbine exhaust is directly cooled by air-cooled condenser. In indirect system, turbine steam is cooled by recirculating intermediate cooling water loop, then the loop is cooled by air-cooled heat exchanger in cooling tower. In this paper, the purpose is to remove SMART waste heat, 200MW by using newly designed tower. The possibility of enhancing cooling performance by solar energy is analyzed. The simple cooling tower and solar energy cooling tower are presented and two design should meet the purpose of removing SMART waste heat, 200MW. In first design, when tower diameter is 70m, the height of tower should be 360m high. In second design, the chimney height decrease from 360m to 180m as collector radius increase from 100m to 500m due to collector temperature enhancement by solar energy, To analyze solar cooling tower further, consideration of solar energy performance at night should be analyzed

Design of SMART waste heat removal dry cooling tower using solar energy

Energy Technology Data Exchange (ETDEWEB)

Choi, Yong Jae; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

The 85% of cooling system are once-through cooling system and closed cycle wet cooling system. However, many countries are trying to reduce the power plant water requirement due to the water shortage and water pollution. Dry cooling system is investigated for water saving advantage. There are two dry cooling system which are direct and indirect cooling system. In direct type, turbine exhaust is directly cooled by air-cooled condenser. In indirect system, turbine steam is cooled by recirculating intermediate cooling water loop, then the loop is cooled by air-cooled heat exchanger in cooling tower. In this paper, the purpose is to remove SMART waste heat, 200MW by using newly designed tower. The possibility of enhancing cooling performance by solar energy is analyzed. The simple cooling tower and solar energy cooling tower are presented and two design should meet the purpose of removing SMART waste heat, 200MW. In first design, when tower diameter is 70m, the height of tower should be 360m high. In second design, the chimney height decrease from 360m to 180m as collector radius increase from 100m to 500m due to collector temperature enhancement by solar energy, To analyze solar cooling tower further, consideration of solar energy performance at night should be analyzed.

Ecosystem and restoration consequences of invasive woody species removal in Hawaiian lowland wet forest

Science.gov (United States)

R. Ostertag; S. Cordell; J. Michaud; T.C. Cole; J.R. Schulten; K.M. Publico; J.H. Enoka

2009-01-01

A removal experiment was used to examine the restoration potential of a lowland wet forest in Hawaii, a remnant forest type that has been heavily invaded by non-native species and in which there is very little native species regeneration. All non-native woody and herbaceous biomass (approximately 45% of basal area) was removed in four 100-m² removal plots;...

Removal of fine particles in wet flue gas desulfurization system by heterogeneous condensation

Energy Technology Data Exchange (ETDEWEB)

Yang, L.J.; Bao, J.J.; Yan, J.P.; Liu, J.H.; Song, S.J.; Fan, F.X. [Southeast University, Nanjing (China). School of Energy & Environment

2010-01-01

A novel process to remove fine particles with high efficiency by heterogeneous condensation in a wet flue gas desulfurization (WFGD) system is presented. A supersaturated vapor phase, necessary for condensational growth of fine particles, was achieved in the SO{sub 2} absorption zone and at the top of the wet FGD scrubber by adding steam in the gas inlet and above the scrubbing liquid inlet of the scrubber, respectively. The condensational grown droplets were then removed by the scrubbing liquid and a high-efficiency demister. The results show that the effectiveness of the WFGD system for removal of fine particles is related to the SO{sub 2} absorbent employed. When using CaCO{sub 3} and NH{sub 3} {center_dot} H{sub 2}O to remove SO{sub 2} from flue gas, the fine particle removal efficiencies are lower than those for Na2CO{sub 3} and water, and the morphology and elemental composition of fine particles are changed. This effect can be attributed to the formation of aerosol particles in the limestone and ammonia-based FGD processes. The performance of the WFGD system for removal of fine particles can be significantly improved for both steam addition cases, for which the removal efficiency increases with increasing amount of added steam. A high liquid to gas ratio is beneficial for efficient removal of fine particles by heterogeneous condensation of water vapor.

Effects of sulphuric acid, mechanical scarification and wet heat ...

African Journals Online (AJOL)

Effects of different treatment methods on the germination of seeds of Parkia biglobosa (mimosaceae) were carried out. Prior treatment of seeds with sulphuric acid, wet heat and mechanical scarification were found to induce germination of the dormant seeds. These methods could be applied to raise seedlings of the plant for ...

After-heat removing system in FBR type reactor

International Nuclear Information System (INIS)

Ohashi, Yukio.

1990-01-01

The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

After-heat removal system

International Nuclear Information System (INIS)

Yamamoto, Michiyoshi; Mitani, Shinji.

1982-01-01

Purpose: To prevent contamination of suppression pool water and intrusion of corrosion products into a nuclear reactor. Constitution: Upon stop of an after-heat removing system, reactor water contained in pipelines is drained out to a radioactive wastes processing facility at the time the cooling operation mode has been completed. At the same time, water is injected from a pure water supply system to the after-heat removing system to discharge corrosion product and activated materials while cleaning the inside of the pipelines. Then, pure water is held in the pipelines and it is discharged again and replaced with pure water before entering the cooling mode operation. Thereafter, the cooling mode operation upon reactor shutdown is performed. (Yoshino, Y.)

Technology Demonstration of Wet Abrasive Blasting for Removal of Lead- and Asbestos-Containing Paint

National Research Council Canada - National Science Library

Race, Timothy

2003-01-01

...). This technology demonstration showed that wet blasting using an engineered abrasive can safely and effectively remove lead- and asbestos-containing paint from exterior concrete masonry unit walls...

Methods for sulfate removal in liquid-phase catalytic hydrothermal gasification of biomass

Science.gov (United States)

Elliott, Douglas C; Oyler, James

2013-12-17

Processing of wet biomass feedstock by liquid-phase catalytic hydrothermal gasification must address catalyst fouling and poisoning. One solution can involve heating the wet biomass with a heating unit to a pre-treatment temperature sufficient for organic constituents in the feedstock to decompose, for precipitates of inorganic wastes to form, for preheating the wet feedstock in preparation for subsequent removal of soluble sulfate contaminants, or combinations thereof. Processing further includes reacting the soluble sulfate contaminants with cations present in the feedstock material to yield a sulfate-containing precipitate and separating the inorganic precipitates and/or the sulfate-containing precipitates out of the wet feedstock. Having removed much of the inorganic wastes and the sulfate contaminants that can cause poisoning and fouling, the wet biomass feedstock can be exposed to the heterogenous catalyst for gasification.

Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen

2005-08-29

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

Nutrient Removal of Grey Water from Wet Market Using Sequencing Batch Reactor

International Nuclear Information System (INIS)

Omar Danial; Mohd Razman Salim; Salmiati

2016-01-01

Fresh water scarcity has become an important issue in this world today. Water reuse is known as one of the strategies to overcome this problem. Grey water is one of the sources of reused water. Several researches were carried out on water reuse, but limited attention was focused on reusing grey water from wet market, which contains high nutrient and organic matters. This study was carried out on nutrient removal from grey water using sequencing batch reactor (SBR). The grey water sample was taken from a wet market (Pasar Peladang, Skudai). About 1L of grey water was fed into the reactor with a total volume of 4L. Anoxic-aerobic phase were divided with a ratio of 30 % - 70 % of total time respectively. Mixing was maintained at 30 rpm during the start of each cycle until settling phase to achieve uniform condition. Influent and effluent were set for 30 minutes. The SBR was operated with 3 cycles/ day, temperature 30 degree Celsius, cycle time 8 hours and hydraulic retention time (HRT) 1.2 days. Aeration at 35 L/ min was induced for ammonia conversion and assisting nitrification.. The results show that the bacteria growing in alternating anoxic/ aerobic systems could remove organic substrates and nutrient. The COD, Total Nitrogen and Total Phosphorus removal efficiencies were maximum at the levels of 94 %, 88 % and 70 % respectively. Anaerobic-Aerobic-Anoxic phase was proposed to increase the removal percentage. (author)

Air-side performance of a micro-channel heat exchanger in wet surface conditions

Directory of Open Access Journals (Sweden)

Srisomba Raviwat

2017-01-01

Full Text Available The effects of operating conditions on the air-side heat transfer, and pressure drop of a micro-channel heat exchanger under wet surface conditions were studied experimentally. The test section was an aluminum micro-channel heat exchanger, consisting of a multi-louvered fin and multi-port mini-channels. Experiments were conducted to study the effects of inlet relative humidity, air frontal velocity, air inlet temperature, and refrigerant temperature on air-side performance. The experimental data were analyzed using the mean enthalpy difference method. The test run was performed at relative air humidities ranging between 45% and 80%; air inlet temperature ranges of 27, 30, and 33°C; refrigerant-saturated temperatures ranging from 18 to 22°C; and Reynolds numbers between 128 and 166. The results show that the inlet relative humidity, air inlet temperature, and the refrigerant temperature had significant effects on heat transfer performance and air-side pressure drop. The heat transfer coefficient and pressure drop for the micro-channel heat exchanger under wet surface conditions are proposed in terms of the Colburn j factor and Fanning f factor.

Pilot-scale field study for ammonia removal from lagoon biogas using an acid wet scrubber.

Science.gov (United States)

Lin, Hongjian; Wu, Xiao; Miller, Curtis; Zhu, Jun; Hadlocon, Lara Jane; Manuzon, Roderick; Zhao, Lingying

2014-01-01

The anaerobic activities in swine slurry storage and treatment generate biogas containing gaseous ammonia component which is a chemical agent that can cause adverse environmental impacts when released to the atmosphere. The aim of this pilot plant study was to remove ammonia from biogas generated in a covered lagoon, using a sulfuric acid wet scrubber. The data showed that, on average, the biogas contained 43.7 ppm of ammonia and its concentration was found to be exponentially related to the air temperature inside the lagoon. When the air temperature rose to 35°C and the biogas ammonia concentration reached 90 ppm, the mass transfer of ammonia/ammonium from the deeper liquid body to the interface between the air and liquid became a limiting factor. The biogas velocity was critical in affecting ammonia removal efficiency of the wet scrubber. A biogas flow velocity of 8 to 12 mm s(-1) was recommended to achieve a removal efficiency of greater than 60%. Stepwise regression revealed that the biogas velocity and air temperature, not the inlet ammonia concentration in biogas, affected the ammonia removal efficiency. Overall, when 73 g L(-1) (or 0.75 M) sulfuric acid solution was used as the scrubber solution, removal efficiencies varied from 0% to 100% with an average of 55% over a 40-d measurement period. Mass balance calculation based on ammonium-nitrogen concentration in final scrubber liquid showed that about 21.3 g of ammonia was collected from a total volume of 1169 m(3) of biogas, while the scrubber solution should still maintain its ammonia absorbing ability until its concentration reaches up to 1 M. These results showed promising use of sulfuric acid wet scrubber for ammonia removal in the digester biogas.

Analysis of decay heat removal following loss of RHR

International Nuclear Information System (INIS)

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

1991-01-01

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

Critical heat fluxes in tubular fuel elements of nuclear power reactors

International Nuclear Information System (INIS)

Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

1974-01-01

The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

Prediction of Heat Removal Capacity of Horizontal Condensation Heat Exchanger submerged in Pool

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Park, Goon-Cherl [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

As representative passive safety systems, there are the passive containment cooling system (PCCS) of ESBWR, the emergency condenser system (ECS) of the SWR-1000, the passive auxiliary feed-water system (PAFS) of the APR+ and etc. During the nuclear power plant accidents, these passive safety systems can cool the nuclear system effectively via the heat transfer through the steam condensation, and then mitigate the accidents. For the optimum design and the safety analysis of the passive safety system, it is essential to predict the heat removal capacity of the heat exchanger well. The heat removal capacity of the horizontal condensation heat exchanger submerged in a pool is determined by a combination of a horizontal in-tube condensation heat transfer and a boiling heat transfer on the horizontal tube. Since most correlations proposed in the previous nuclear engineering field were developed for the vertical tube, there is a certain limit to apply these correlations to the horizontal tube. Therefore, this study developed the heat transfer model for the horizontal Ushaped condensation heat exchanger submerged in a pool to predict well the horizontal in-tube condensation heat transfer, the boiling heat transfer on the horizontal tube and the overall heat removal capacity of the heat exchanger using the best-estimate system analysis code, MARS.

After-heat removal system of fast reactor

International Nuclear Information System (INIS)

Otsuka, Masaya; Shibata, Yoji; Ikeda, Takashi; Iwashige, Kengo; Yoneda, Yoshiyuki.

1988-01-01

Purpose: To remove after-heat by natural convection without disposing a movable portion even in a large-scaled reactor. Constitution: The exit of a reactor wall air-cooling duct disposed to the outside of a safety vessel is connected to the secondary inlet of an air cooler that conducts heat exchange with sodium in a high temperature plenum. That is, after-heat is removed only through the natural convection by a structure in which the reactor wall air-cooling duct and the secondary side of the air cooler are connected in series. Air exhausted from the exit of the air-cooling duct by the air cooler is further heated with sodium in the high temperature plenum. The flow rate of air flowing through the air-cooling duct is increased as compared with the case where the air cooler is not present. Accordingly, the flow rate of air at low temperature flowing through the inlet of the air duct is increased to increase the heat conduction amount. In this way, after-heat can be removed only by means of natural convection without providing movable portions even in a large-scaled reactor with the thermal power in excess of 2,000 MW. (Horiuchi, T.)

Advances in technologies for decay heat removal

International Nuclear Information System (INIS)

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

1999-01-01

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

Heat removing under hypersonic conditions

Directory of Open Access Journals (Sweden)

Semenov Mikhail E.

2016-01-01

Full Text Available In this paper we consider the heat transfer properties of the axially symmetric body with parabolic shape at hypersonic speeds (with a Mach number M > 5. We use the numerical methods based on the implicit difference scheme (Fedorenko method with direct method based on LU-decomposition and iterative method based on the Gauss-Seigel method. Our numerical results show that the heat removing process should be performed in accordance with the nonlinear law of heat distribution over the surface taking into account the hypersonic conditions of motion.

ABB wet flue gas desulfurization

Energy Technology Data Exchange (ETDEWEB)

Niijhawan, P.

1994-12-31

The wet limestone process for flue gas desulfurization (FGD) is outlined. The following topics are discussed: wet flue gas desulfurization, wet FGD characteristics, wet scrubbers, ABB wet FGD experience, wet FGD forced oxidation, advanced limestone FGD systems, key design elements, open spray tower design, spray tower vs. packed tower, important performance parameters, SO{sub 2} removal efficiency, influence by L/G, limestone utilization, wet FGD commercial database, particulate removal efficiencies, materials of construction, nozzle layout, spray nozzles, recycle pumps, mist elimination, horizontal flow demister, mist eliminator washing, reagent preparation system, spray tower FGDS power consumption, flue gas reheat options, byproduct conditioning system, and wet limestone system.

Solution of heat removal from nuclear reactors by natural convection

Directory of Open Access Journals (Sweden)

Zitek Pavel

2014-03-01

Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Removal of Atmospheric Ethanol by Wet Deposition: A Global Flux Estimate

Science.gov (United States)

Felix, J. D. D.; Willey, J. D.; Avery, B.; Thomas, R.; Mullaugh, K.; Kieber, R. J.; Mead, R. N.; Helms, J. R.; Campos, L.; Shimizu, M. S.; Guibbina, F.

2017-12-01

Global ethanol fuel consumption has increased exponentially over the last two decades and the US plans to double annual renewable fuel production in the next five years as required by the renewable fuel standard. Regardless of the technology or feedstock used to produce the renewable fuel, the primary end product will be ethanol. Increasing ethanol fuel consumption will have an impact on the oxidizing capacity of the atmosphere and increase atmospheric concentrations of the secondary pollutant peroxyacetyl nitrate as well a variety of VOCs with relatively high ozone reactivities (e.g. ethanol, formaldehyde, acetaldehyde). Despite these documented effects of ethanol emissions on atmospheric chemistry, current global atmospheric ethanol budget models have large uncertainties in the magnitude of ethanol sources and sinks. The presented work investigates the global wet deposition sink by providing the first estimate of the global wet deposition flux of ethanol (2.4 ± 1.6 Tg/yr) based on empirical wet deposition data (219 samples collected at 12 locations). This suggests the wet deposition sink removes between 6 and 17% of atmospheric ethanol annually. Concentrations of ethanol in marine wet deposition (25 ± 6 nM) were an order of magnitude less than in the majority of terrestrial deposition (345 ± 280 nM). Terrestrial deposition collected in locations impacted by high local sources of biofuel usage and locations downwind from ethanol distilleries were an order of magnitude higher in ethanol concentration (3090 ± 448 nM) compared to deposition collected in terrestrial locations not impacted by these sources. These results indicate that wet deposition of ethanol is heavily influenced by local sources and ethanol emission impacts on air quality may be more significant in highly populated areas. As established and developing countries continue to rapidly increase ethanol fuel consumption and subsequent emissions, understanding the magnitude of all ethanol sources and

Decay heat removal for the liquid metal fast breeder reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

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

1975-10-15

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

Highly heat removing radiation shielding material

International Nuclear Information System (INIS)

Asano, Norio; Hozumi, Masahiro.

1990-01-01

Organic materials, inorganic materials or metals having excellent radiation shielding performance are impregnated into expanded metal materials, such as Al, Cu or Mg, having high heat conductivity. Further, the porosity of the expanded metals and combination of the expanded metals and the materials to be impregnated are changed depending on the purpose. Further, a plurality of shielding materials are impregnated into the expanded metal of the same kind, to constitute shielding materials. In such shielding materials, impregnated materials provide shielding performance against radiation rays such as neutrons and gamma rays, the expanded metals provide heat removing performance respectively and they act as shielding materials having heat removing performance as a whole. Accordingly, problems of non-informity and discontinuity in the prior art can be dissolved be provide materials having flexibility in view of fabrication work. (T.M.)

An innovative pool with a passive heat removal system

International Nuclear Information System (INIS)

Vitale Di Maio, Damiano; Naviglio, Antonio; Giannetti, Fabio; Manni, Fabio

2012-01-01

Heat removal systems are of primary importance in several industrial processes. As heat sink, a water pool or atmospheric air may be selected. The first solution takes advantage of high heat transfer coefficient with water but it requires active systems to maintain a constant water level; the second solution takes benefit from the unlimited heat removal capacity by air, but it requires a larger heat exchanger to compensate the lower heat transfer coefficient. In NPPs (nuclear power plants) during a nuclear reactor shutdown, as well as in some chemical plants to control runaway reactions, it is possible to use an innovative heat sink that joins the advantages of the two previous solutions. This solution is based on a special heat exchanger submerged in a water pool designed so that when heat removal is requested, active systems are not required to maintain the water level; due to the special design, when the pool is empty, atmospheric air becomes the only heat sink. The special heat exchanger design allows to have a heat exchanger without being oversized and to have a system able to operate for unlimited period without external interventions. This innovative system provides an economic advantage as well as enhanced safety features.

Study on thermal-hydraulic phenomena identification of passive heat removal facilities

International Nuclear Information System (INIS)

Park, J. Y.

2011-01-01

Recently, passive heat removal facilities have been integral features of new generation or future reactor designs worldwide. This is because the passive heat removal facilities depending on a natural force such as buoyancy can give much higher operational reliability compared to active heat removal facilities depending on pumped fluid flow and as a result they can decrease core damage frequency of a nuclear power plant drastically ever achievable before. Keeping pace with this global trend, SMART and APR+ reactors also have introduced passive heat removal features such as a passive residual heat removal system (PRHRS) and a passive auxiliary feed water system (PAFS) in their designs. Since many thermal-hydraulic (T-H) phenomena including steam condensation are involved during operation of the passive heat removal facilities, they ought to be properly simulated by T-H codes such as MARS-KS and RELAP5 in order to guarantee reliable safety analysis by these codes. Unfortunately, however, these T-H codes are not well validated with respect to phenomena related to passive heat removal mechanism because previous focus on these codes validation was mainly on the LB LOCA and resulting phenomena. To resolve this gap, Korea Institute of Nuclear Safety has initiated a research program on the development of safety analysis technology for passive heat removal facilities. The main target of this program is PRHRS and PAFS in SMART and APR+ reactors and through this program, validation of capability of existing T-H codes and improvement of codes regarding passive facilities analysis are to be sought. In part of this research, T-H phenomena important to passive heat removal facilities (PRHRS and PAFS) are investigated in the present study

Assessment of ASME code examinations on regenerative, letdown and residual heat removal heat exchangers

International Nuclear Information System (INIS)

Gosselin, Stephen R.; Cumblidge, Stephen E.; Anderson, Michael T.; Simonen, Fredric A.; Tinsley, G A.; Lydell, B.; Doctor, Steven R.

2005-01-01

Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principle reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers regenerative heat exchangers and residual (decay) heat removal heat exchangers Design, operating, common preventative maintenance practices, and potential degradation mechanisms are reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp- database and covers PWR plants worldwide for the period 1970-2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers is performed. The results are then reviewed to discuss the examinations relative to plant safety and

CAREM-25: Residual heat removal system

International Nuclear Information System (INIS)

Arvia, Roberto P.; Coppari, Norberto R.; Gomez de Soler, Susana M.; Ramilo, Lucia B.

2000-01-01

The objective of this work was the definition and consolidation of the residual heat removal system for the CAREM 25 reactor. The function of this system is cool down the primary circuit, removing the core decay heat from hot stand-by to cold shutdown and during refueling. In addition, this system heats the primary water from the cold shutdown condition to hot stand-by condition during the reactor start up previous to criticality. The system has been designed according to the requirements of the standards: ANSI/ANS 51.1 'Nuclear safety criteria for the design of stationary PWR plants'; ANSI/ANS 58.11 'Design criteria for safe shutdown following selected design basis events in light water reactors' and ANSI/ANS 58.9 'Single failure criteria for light water reactor safety-related fluid systems'. The suggested design fulfills the required functions and design criteria standards. (author)

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

International Nuclear Information System (INIS)

Nishi, Y.; Kinoshita, I.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Passive heat removal in CANDU

International Nuclear Information System (INIS)

Hart, R.S.

1997-01-01

CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

International Nuclear Information System (INIS)

1991-03-01

An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Design and analysis of a new passive residual heat removal system

Energy Technology Data Exchange (ETDEWEB)

Lv, Xing [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Peng, Minjun, E-mail: heupmj@163.com [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Yuan, Xiao [Guangxi Fangchenggang Nuclear Power Co., Ltd (China); Xia, Genglei [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China)

2016-07-15

Highlights: • An air cooling passive residual heat removal System (PRHRs) is designed. • Using RELAP5/MOD3.4 code to analyze the operation characteristics of the PRHRs. • Noncondensable gas is used to simulate the hydrodynamic behavior in the air cooling tower. • The natural circulations could respectively establish in the primary circuit and the PRHRs circuit. • The PRHRs could remove the residual heat effectively. - Abstract: The inherent safety functions will mitigate the consequences of the accidents, and it can be accomplished through the passive safety systems which employed in the typical pressurized water reactor (PWR). In this paper, a new passive residual heat removal system (PRHRS) is designed for a typical nuclear power plant. PRHRS consists of a steam generator (SG), a cooling tank with two groups of cooling pipes, an air-cooling heat exchanger (AHX), an air-cooling tower, corresponding pipes and valves. The cooling tank which works as an intermediate buffer device is used to transfer the core decay heat to the AHX, and then the core decay heat will be removed to the atmosphere finally. The RELAP5/MOD3.4 code is used to analyze the operation characteristics of PRHRS and the primary loop system. It shows PRHRS could remove the decay heat from the primary loop effectively, and the natural circulations can be established in the primary circuit and the PRHRS circuit respectively. Furthermore, the sensitivity study has also been done to research the effect of various factors on the heat removal capacity.

Passive Decay Heat Removal System for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

The kinetics of removal of heat-induced excess nuclear protein

International Nuclear Information System (INIS)

Roti, J.L.R.; Uygur, N.; Higashikubo, R.

1984-01-01

To investigate the role of protein content, temperature and heating time in the removal of heat-induced excess protein associated with the isolated nucleus, the kinetics of protein removal was monitored for 6 to 8 hours following exposure to 7 hyperthermic protocols. Four of these (47 0 C-7.5 min., 46 0 C-15 min., 45 0 C-30 min., and 44 0 C-60 min.) resulted in a nuclear protein content approximately twice that of nuclei from unheated cells (2.05 +- .14) following heat exposure. Three protocols (45 0 C-15 min., 44 0 C-30 min. and 43 0 C-60 min.) resulted in a nuclear protein content approximately 1.6 times normal (1.63 +- .12). If nuclear protein content were the only determinant in the recovery rate, then the same half time for nuclear protein removal would be expected within each group of protocols. Rate constants for nuclear protein removal were obtained by regression analysis. The half-time for nuclear protein removal increased with decreasing temperature and increasing heating time for the same nuclear protein content. This result suggests that the heating time and temperature are more of a determinant in the removal kinetics than protein content alone. Extended kinetics of recovery (to 36 hours) showed incomplete recovery and a secondary increase in protein associated with the isolated nucleus. These results were due to cell-cycle rearrangement (G/sub 2/ block) and unbalanced growth

Evaluation of specific lipid production and nutrients removal from wastewater by Rhodosporidium toruloides and biodiesel production from wet biomass via microwave irradiation

International Nuclear Information System (INIS)

Ling, Jiayin; Nip, Saiwa; Alves de Toledo, Renata; Tian, Yuan; Shim, Hojae

2016-01-01

This study investigated the potential use of yeast Rhodosporidium toruloides for lipid production and removal of organics and nutrients while treating a mixture of distillery and domestic wastewater at low cost without sterilization. The highest specific biomass (24.1 × 10"−"9 g L"−"1 cell"−"1 h"−"1), lipid yield (9.9 × 10"−"9 g L"−"1 cell"−"1 h"−"1), and lipid content (8.54 × 10"−"3 g lipid g"−"1 biomass h"−"1) were achieved on the 2nd day of cultivation. The organics and nutrients removal also reached the highest removal rates within 2 days, with the specific removal rates for COD (chemical oxygen demand), TN (total nitrogen), and TP (total phosphorus) at 39.5, 2.57, and 0.29 × 10"−"9 mg cell"−"1 h"−"1, respectively. The indigenous microorganisms slightly contributed to both lipid production and removal of organics and nutrients. The direct transesterification process via microwave irradiation from wet biomass was further optimized under the conditions of solvent to wet biomass rate 16 mL/g, methanol and chloroform mix rate 1:1, addition of catalyst H_2SO_4 6%, and reaction time 10 min at the reaction temperature of 60 °C. The FAMEs (fatty acid methyl esters) composition and the possibility of biodiesel production from wet oleaginous yeast biomass by the direct transesterification were also assessed. - Highlights: • Yeast Rhodosporidium toruloides used for lipid production in real mixed wastewater. • No significant lipid production and nutrients removal by indigenous organisms. • Highest specific lipid production and nutrient removal rates achieved with 0.05 × 10"8 cells mL"−"1. • Biodiesel production from wet biomass via microwave assisted direct transesterification is feasible.

The heat engine cycle, the heat removal cycle, and ergonomics of the control room displays

International Nuclear Information System (INIS)

Beltracchi, L.

1986-01-01

This paper discusses and illustrates the ergonomics of an integrated display, which will allow operators to monitor the heat engine cycle during normal operation of the plant, and the heat removal cycle during emergency operation of the plant. A computer-based iconic display is discussed as an overview to monitor these cycles. Specific emphasis is placed upon the process variables and process functions within each cycle, and the action of control systems and engineered safeguard systems within each cycle. This paper contains examples of display formats for the heat engine cycle and the heat removal cycle in a pressurized water reactor

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

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

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

Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

International Nuclear Information System (INIS)

Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

2015-01-01

Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

After-heat removing device in nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Mizuno, K [Nippon Atomic Industry Group Co. Ltd., Tokyo

1977-01-14

Purpose: To prevent water hammer in a BWR type reactor or the like by moving water in pipe lines having stagnant portions in an after-heat removing device. Constitution: To a reactor container, is provided a recycling pump which constitutes a closed loop type recycling system in a nuclear power plant together with a pressure vessel and pipe lines. A pump and a heat exchanger are provided outside of the reactor container and they are connected to up- and down-streams of the recycling pump to form an after-heat removing device in the plant. Upon shutdown of the nuclear power plant, since water in the stagnant portion flows to the intake port of the recycling pump and water from the reactor is spontaneously supplemented thereafter to the stagnant portion, neither pressurized water nor heated steam is generated and thus water hammer is prevented.

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

Heat removal capability of core-catcher with inclined cooling channels

International Nuclear Information System (INIS)

Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

2009-01-01

A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

Postaccident heat removal. II. Heat transfer from an internally heated liquid to a melting solid

International Nuclear Information System (INIS)

Faw, R.E.; Baker, L. Jr.

1976-01-01

Microwave heating has been used in studies of heat transfer from a horizontal layer of internally heated liquid to a melting solid. Experiments were designed to simulate heat transfer and meltthrough processes of importance in the analysis of postaccident heat removal capabilities of nuclear reactors. Glycerin, heated by 2.45-GHz microwave radiation, was used to simulate molten fuel. Paraffin wax was used to simulate a melting barrier confining the fuel. Experimentally measured heat fluxes and melting rates were consistent with a model based on downward heat transfer by conduction through a stagnant liquid layer and upward heat transfer augmented by natural convection. Melting and displacement of the barrier material occurred by upward-moving droplets randomly distributed across the melting surface. Results indicated that the melting and displacement process had no effect on the heat transfer process

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

International Nuclear Information System (INIS)

Kim, Yong Hoon; Suh, Kune Y.

2000-01-01

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

Passive heat removal system with injector-condenser

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Heat removal in INTOR via a toroidal limiter

International Nuclear Information System (INIS)

Mioduszewski, P.

1981-01-01

In the present paper the potential of removing about 100 MW of thermal plasma power via a toroidal limiter in INTOR is studied. The heat flux distributions on various limiter configurations are calculated and the thermal response of a graphite tile limiter is estimated on the base of a one-dimensional heat conduction approach. The evaporation rates which have to be expected for the given energy flux densities and radiation cooled graphite tiles are evaluated. According to the present understanding it should be possible to remove 100 MW power from the INTOR plasma via a radiation cooled toroidal limiter. (author)

Decision Document for Heat Removal from High-Level Waste Tanks

International Nuclear Information System (INIS)

WILLIS, W.L.

2000-01-01

This document establishes the combination of design and operational configurations that will be used to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. The chosen method--to use the primary and annulus ventilation systems to remove heat from the high-level waste tanks--is documented herein

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Corletti, M.M.; Schulz, T.L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Safety studies on heat transport and afterheat removal for GCR accident conditions

International Nuclear Information System (INIS)

Hishida, Makoto

1996-01-01

The IAEA coordinated an international research program on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions (CRP-3)'. America, China, France, Germany, Japan, Netherlands and Russia participate the program. Final goal of the program is to show clearly to the world one of the most important salient features of the HTGR, that is the HTGR reactor can be cooled down by passive measures without causing any damage to the nuclear reactor system even in accidental conditions, and to make clear the boundaries (or restrictions) for the passive cooling regime. The first 5 year term of the coordinate program started in 1993 and established a goal to improve common knowledge for decay heat removal and to improve our tools, like computer codes and analytical models for the prediction of the performance of decay heat removal system. We are now performing benchmark problems for these purposes. The present efforts are concentrated on the benchmark for the passive heat removal performance outside the reactor vessel, partly because we have two different type of the HTGR in the world, the pebble bed type and the block type reactor. They have quite different heat dissipation behavior inside the reactor vessel. However, they have quite similar residual heat removal process outside the reactor vessel. For the first step of the international cooperation, we selected the common problem. After finishing the present benchmark we are planning to proceed to tackle the inside heat removal problem. (J.P.N.)

Residual heat removal pump retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; McKenna, J.M.

1990-01-01

Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

Heat transfer and flow characteristics of a cooling thimble in a molten salt reactor residual heat removal system

Directory of Open Access Journals (Sweden)

Zonghao Yang

2017-12-01

Full Text Available In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

Heat balance model for a human body in the form of wet bulb globe temperature indices.

Science.gov (United States)

Sakoi, Tomonori; Mochida, Tohru; Kurazumi, Yoshihito; Kuwabara, Kohei; Horiba, Yosuke; Sawada, Shin-Ichi

2018-01-01

The purpose of this study is to expand the empirically derived wet bulb globe temperature (WBGT) index to a rational thermal index based on the heat balance for a human body. We derive the heat balance model in the same form as the WBGT for a human engaged in moderate intensity work with a metabolic heat production of 174W/m 2 while wearing typical vapor-permeable clothing under shady and sunny conditions. Two important relationships are revealed based on this derivation: (1) the natural wet bulb and black globe temperature coefficients in the WBGT coincide with the heat balance equation for a human body with a fixed skin wettedness of approximately 0.45 at a fixed skin temperature; and (2) the WBGT can be interpreted as the environmental potential to increase skin temperature rather than the heat storage rate of a human body. We propose an adjustment factor calculation method that supports the application of WBGT for humans dressed in various clothing types and working under various air velocity conditions. Concurrently, we note difficulties in adjusting the WBGT by using a single factor for humans wearing vapor-impermeable protective clothing. The WBGT for shady conditions does not need adjustment depending on the positive radiant field (i.e., when a radiant heat source exists), whereas that for the sunny condition requires adjustments because it underestimates heat stress, which may result in insufficient human protection measures. Copyright © 2017 Elsevier Ltd. All rights reserved.

Shutdown risk analysis for a BWR plant (residual heat removal systems)

International Nuclear Information System (INIS)

Rebollo Garcia, C.; Merino Teillet, A.; Cerezo, L.

1994-01-01

This report analyses the different risk situations which may arise during refuelling outage at Cofrentes NPP. The most critical situations are determined in terms of the small amount of coolant available and the lowest number of heat removal and water make-up systems available. The available times before the boiling point of the coolant is reached and the subsequent moment when the fuel elements are left uncovered in the event of the failure of the normal heat removal functions are determined. The analysis identifies the alternative systems which can be used besides those required by the technical specification and their capacity for residual heat removal and coolant make-up functions. (Author)

Tritium removal by CO2 laser heating

International Nuclear Information System (INIS)

Skinner, C.H.; Kugel, H.; Mueller, D.

1997-01-01

Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

Tritium removal by CO2 laser heating

International Nuclear Information System (INIS)

Skinner, C.H.; Kugel, H.; Mueller, D.

1997-10-01

Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:YAG laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

Toluene removal by oxidation reaction in spray wet scrubber: experimental, modeling and optimization

Directory of Open Access Journals (Sweden)

Roumporn Nikom

2006-11-01

Full Text Available Toluene, an important volatile organic compound (VOC, is used in many kinds of industries, such as painting, printing, coating, and petrochemical industries. The emission of toluene causes serious air pollution, odor problem, flammability problem and affects human health. This paper proposes the removal of toluene from waste air using a spray wet scrubber combining the absorption and oxidation reaction. Aqueous sodium hypochlorite (NaOCl solution was used as the scrubbing liquid in the system. NaOCl, the strongest oxidative agent, presents an effective toluene removal. As the scrubbed toluene is reacted, recirculation of the scrubbing liquid could be operated with a constant removal efficiency throughout the operting time. The investigated variables affecting the removal efficiency were air flow rate, inlet toluene concentration, NaOCl concentration, scrubbing liquid flow rate and size of spray nozzle. Influence of the scrubbing parameters was experimentally studied to develop a mathematical model of the toluene removal efficiency. The removal model reveals that the increase of scrubbing liquid flow rate, toluene concentration, and NaOCl concentration together with the decrease of air flow rate and size of spray nozzle can increase the toluene removal efficiency. Optimization problem with an objective function and constraints was set to provide the maximum toluene removal efficiency and solved by Matlab optimization toolbox. The optimization constraints were formed from the mathematical model and process limitation. The solution of the optimization was an air flow rate of 100 m3/h, toluene concentration of 1500 ppm, NaOCl concentration of 0.02 mol/l, NaOCl solution feed rate of 0.8 m3/h, and spray nozzle size of 0.5 mm. Solution of the optimization gave the highest toluene removal efficiency of 91.7%.

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.)

Results from evaporation tests to support the MWTF heat removal system design

International Nuclear Information System (INIS)

Crea, B.A.

1994-01-01

An experimental tests program was conducted to measure the evaporative heat removal from the surface of a tank of simulated waste. The results contained in this report constitute definition design data for the latest heat removal function of the MWTF primary ventilation system

Passive decay heat removal from the core region

International Nuclear Information System (INIS)

Hichen, E.F.; Jaegers, H.

2002-01-01

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

AEA studies on passive decay heat removal in advanced reactors

International Nuclear Information System (INIS)

Lillington, J.N.

1994-01-01

The main objectives of the UK study were: to identify, describe and compare different types of systems proposed in current designs; to identify key scenarios in which passive decay heat removal systems play an important preventative or mitigative role; to assess the adequacy of the relevant experimental database; to assess the applicability and suitability of current generation models/codes for predicting passive decay heat removal; to assess the potential effectiveness of different systems in respect of certain key licensing questions

Native trees show conservative water use relative to invasive trees: results from a removal experiment in a Hawaiian wet forest.

Science.gov (United States)

Cavaleri, Molly A; Ostertag, Rebecca; Cordell, Susan; Sack, Lawren

2014-01-01

While the supply of freshwater is expected to decline in many regions in the coming decades, invasive plant species, often 'high water spenders', are greatly expanding their ranges worldwide. In this study, we quantified the ecohydrological differences between native and invasive trees and also the effects of woody invasive removal on plot-level water use in a heavily invaded mono-dominant lowland wet tropical forest on the Island of Hawaii. We measured transpiration rates of co-occurring native and invasive tree species with and without woody invasive removal treatments. Twenty native Metrosideros polymorpha and 10 trees each of three invasive species, Cecropia obtusifolia, Macaranga mappa and Melastoma septemnervium, were instrumented with heat-dissipation sap-flux probes in four 100 m(2) plots (two invaded, two removal) for 10 months. In the invaded plots, where both natives and invasives were present, Metrosideros had the lowest sap-flow rates per unit sapwood, but the highest sap-flow rates per whole tree, owing to its larger mean diameter than the invasive trees. Stand-level water use within the removal plots was half that of the invaded plots, even though the removal of invasives caused a small but significant increase in compensatory water use by the remaining native trees. By investigating the effects of invasive species on ecohydrology and comparing native vs. invasive physiological traits, we not only gain understanding about the functioning of invasive species, but we also highlight potential water-conservation strategies for heavily invaded mono-dominant tropical forests worldwide. Native-dominated forests free of invasive species can be conservative in overall water use, providing a strong rationale for the control of invasive species and preservation of native-dominated stands.

Heat removal in gas-cooled fuel rod clusters

International Nuclear Information System (INIS)

Rehme, K.

1975-01-01

For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

Heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

International Nuclear Information System (INIS)

Islam, M.S.; Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio; Monde, Masanori.

1997-03-01

Heat transfer augmentation in narrow rectangular channels in a target system is a very important method to remove high heat flux up to 12 MW/m 2 generated at target plates of a high-intensity proton accelerator of 1.5 GeV and 1 mA with a proton beam power of 1.5 MW. In this report, heat transfer coefficients and friction factors in narrow rectangular channels with one-sided rib-roughened surface were evaluated for fully developed flows in the range of the Reynolds number from 6,000 to 1,00,000; the rib pitch-to-height ratios (p/k) were 10,20 and 30; the rib height-to-equivalent diameter ratios (k/De) were 0.025, 0.03 and 0.1 by means of previous existing experimental correlations. The rib-roughened surface augmented heat transfer coefficients approximately 4 times higher than the smooth surface at Re=10,000, p/k=10 and k/De=0.1; friction factors increase around 22 times higher. In this case, higher heat flux up to 12 MW/m 2 could be removed from the rib-roughened surface without flow boiling which induces flow instability; but pressure drop reaches about 1.8 MPa. Correlations obtained by air-flow experiments have showed lower heat transfer performance with the water-flow conditions. The experimental apparatus was proposed for further investigation on heat transfer augmentation in very narrow channels under water-flow conditions. This report presents the evaluation results and an outline of the test apparatus. (author)

Post-accident fuel relocation and heat removal in the LMFBR

International Nuclear Information System (INIS)

Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

1976-08-01

Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

Removal of contaminated concrete surfaces by microwave heating: Phase 1 results

International Nuclear Information System (INIS)

White, T.L.; Grubb, R.G.; Pugh, L.P.; Foster, D. Jr.; Box, W.D.

1992-01-01

Oak Ridge National Laboratory (ORNL) is developing a microwave heating process to remove radiologically contaminated surface layers from concrete. The microwave energy is directed at the concrete surface and heats the concrete and free water present in the concrete matrix. Continued heating produces steam-pressure-induced mechanical stresses that cause the concrete surface to burst. The concrete particles from this steam explosion are small enough to be removed by a vacuum system, yet less than 1% of the debris is small enough to pose an airborne contamination hazard. The first phase of this program has demonstrated reliable removal of noncontaminated concrete surfaces at frequencies of 2.45 GHz and 10.6 GHz. Continuous concrete removal rates of 1.07 cm 3 /s with 5.2 kW of 2.45.-GHz power and 2.11 cm 3 /s with 3.6 kW of 10.6-GHz power have been demonstrated. Figures-of-merit for microwave removal of concrete have been calculated to be 0.21 cm 3 /s/kW at 2.45 GHz and 0.59 cm 3 /s/kW at 10.6 GHz. The amount of concrete removed in a single pass can be controlled by choosing the frequency and power of the microwave system

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)

Tests for removal of decay heat by natural convection

International Nuclear Information System (INIS)

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

1993-01-01

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

Multiple pollutant removal using the condensing heat exchanger

Energy Technology Data Exchange (ETDEWEB)

Jankura, B. J. [McDermott Technology Inc., Alliance, OH (United States); Kudlac, G. A. [McDermott Technology Inc., Alliance, OH (United States); Bailey, R. T. [McDermott Technology Inc., Alliance, OH (United States)

1998-06-01

The Integrated Flue Gas Treatment (IFGT) system is a new concept whereby a Teflon ® covered condensing heat exchanger is adapted to remove certain flue gas constituents, both particulate and gaseous, while recovering low level heat. The pollutant removal performance and durability of this device is the subject of a USDOE sponsored program to develop this technology. The program was conducted under contract to the United States Department of Energy's Fossil Energy Technology Center (DOE-FETC) and was supported by the Ohio Coal Development Office (OCDO) within the Ohio Department of Development, the Electric Power Research Institute's Environmental Control Technology Center (EPRI-ECTC) and Babcock and Wilcox - a McDermott Company (B&W). This report covers the results of the first phase of this program. This Phase I project has been a two year effort. Phase I includes two experimental tasks. One task dealt principally with the pollutant removal capabilities of the IFGT at a scale of about 1.2MW_t. The other task studied the durability of the Teflon ® covering to withstand the rigors of abrasive wear by fly ash emitted as a result of coal combustion. The pollutant removal characteristics of the IFGT system were measured over a wide range of operating conditions. The coals tested included high, medium and low-sulfur coals. The flue gas pollutants studied included ammonia, hydrogen chloride, hydrogen fluoride, particulate, sulfur dioxide, gas phase and particle phase mercury and gas phase and particle phase trace elements. The particulate removal efficiency and size distribution was investigated. These test results demonstrated that the IFGT system is an effective device for both acid gas absorption and fine particulate collection. Although soda ash was shown to be the most effective reagent for acid gas absorption, comparative cost analyses suggested that magnesium enhanced lime was the most promising avenue for future study. The durability of the

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Cryogen spray cooling: Effects of droplet size and spray density on heat removal.

Science.gov (United States)

Pikkula, B M; Torres, J H; Tunnell, J W; Anvari, B

2001-01-01

Cryogen spray cooling (CSC) is an effective method to reduce or eliminate non-specific injury to the epidermis during laser treatment of various dermatological disorders. In previous CSC investigations, fuel injectors have been used to deliver the cryogen onto the skin surface. The objective of this study was to examine cryogen atomization and heat removal characteristics of various cryogen delivery devices. Various cryogen delivery device types including fuel injectors, atomizers, and a device currently used in clinical settings were investigated. Cryogen mass was measured at the delivery device output orifice. Cryogen droplet size profiling for various cryogen delivery devices was estimated by optically imaging the droplets in flight. Heat removal for various cryogen delivery devices was estimated over a range of spraying distances by temperature measurements in an skin phantom used in conjunction with an inverse heat conduction model. A substantial range of mass outputs were measured for the cryogen delivery devices while heat removal varied by less than a factor of two. Droplet profiling demonstrated differences in droplet size and spray density. Results of this study show that variation in heat removal by different cryogen delivery devices is modest despite the relatively large difference in cryogen mass output and droplet size. A non-linear relationship between heat removal by various devices and droplet size and spray density was observed. Copyright 2001 Wiley-Liss, Inc.

Summary report of RAMONA investigations into passive decay heat removal

International Nuclear Information System (INIS)

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

1995-07-01

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

Numerical analysis of cavitating flow characteristics in impeller of residual heat removal pump

NARCIS (Netherlands)

Hong, Feng; Yuan, Jianping; Zhou, Banglun

2016-01-01

In order to investigate internal cavitating flow characteristics of the impeller in residual heat removal pumps, the three-dimensional cavitating flow in a residual heat removal model pump is numerically calculated by using the homogeneous mixture cavitation model based on the Rayleigh-Plesset

Removal kinetics for gaseous NO and SO2by an aqueous NaClO2solution mist in a wet electrostatic precipitator

KAUST Repository

Park, Hyun-Woo; Park, Dong-Wha

2016-01-01

Removal kinetics for NO and SO2 by NaClO2 solution mist were investigated in a wet electrostatic precipitator. By varying the molar concentrations of NO, SO2, and NaClO2, the removal rates of NO and SO2 confirmed to range from 34.8 to 72.9 mmol/m3

Heat index and adjusted temperature as surrogates for wet bulb globe temperature to screen for occupational heat stress.

Science.gov (United States)

Bernard, Thomas E; Iheanacho, Ivory

2015-01-01

Ambient temperature and relative humidity are readily ava-ilable and thus tempting metrics for heat stress assessment. Two methods of using air temperature and relative humidity to create an index are Heat Index and Adjusted Temperature. The purposes of this article are: (1) to examine how well Heat Index and Adjusted Temperature estimated the wet bulb globe temperature (WBGT) index, and (2) to suggest how Heat Index and Adjusted Temperature can be used to screen for heat stress level. Psychrometric relationships were used to estimate values of actual WBGT for conditions of air temperature, relative humidity, and radiant heat at an air speed of 0.5 m/s. A relationship between Heat Index [°F] and WBGT [°C] was described by WBGT = -0.0034 HI(2) + 0.96 HI - 34. At lower Heat Index values, the equation estimated WBGTs that were ± 2 °C-WBGT around the actual value, and to about ± 0.5 °C-WBGT for Heat Index values > 100 °F. A relationship between Adjusted Temperature [°F] and WBGT [°C] was described by WBGT = 0.45 Tadj - 16. The actual WBGT was between 1 °C-WBGT below the estimated value and 1.4 °C-WBGT above. That is, there was a slight bias toward overestimating WBGT from Adjusted Temperature. Heat stress screening tables were constructed for metabolic rates of 180, 300, and 450 W. The screening decisions were divided into four categories: (1) exposure limit at rest. The authors do not recommend using Heat Index or Adjusted Temperature instead of WBGT, but they may be used to screen for circumstances when a more detailed analysis using WBGT is appropriate. A particular weakness is accounting for radiant heat; and neither air speed nor clothing was considered.

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

Impact of the amount of working fluid in loop heat pipe to remove waste heat from electronic component

Directory of Open Access Journals (Sweden)

Smitka Martin

2014-03-01

Full Text Available One of the options on how to remove waste heat from electronic components is using loop heat pipe. The loop heat pipe (LHP is a two-phase device with high effective thermal conductivity that utilizes change phase to transport heat. It was invented in Russia in the early 1980’s. The main parts of LHP are an evaporator, a condenser, a compensation chamber and a vapor and liquid lines. Only the evaporator and part of the compensation chamber are equipped with a wick structure. Inside loop heat pipe is working fluid. As a working fluid can be used distilled water, acetone, ammonia, methanol etc. Amount of filling is important for the operation and performance of LHP. This work deals with the design of loop heat pipe and impact of filling ratio of working fluid to remove waste heat from insulated gate bipolar transistor (IGBT.

Study on constraints for heat removal duties of the main fractionator in delayed coking units

International Nuclear Information System (INIS)

Lei, Yang; Zhang, Bingjian; Qi, Xin; Chen, Qinglin; Hui, Chi-Wai

2014-01-01

A novel method is presented in this paper to quantitatively define the heat removal of the main fractionator in delayed coking units on the basis of a fractionating precision diagram (Houghland diagram) and column grand composite curve (CGCC). By referring to the CGCC method, several envelopes are illustrated at draw trays including the top pumparound draw, diesel draw, intermediate pumparound draw and gas oil draw, the energy and material balances are then calculated. Assuming practical near-minimum thermodynamic condition (PNMTC), the minimum liquid reflux flow is zero in the envelope for pumparound trays without product draw and the minimum liquid reflux flow is defined by Houghland diagram for pumparound trays with product draw. The PNMTC-CGCC is constructed by calculating the enthalpy-flow deficit to quantitatively define the heat removal constraints in each envelope. Meanwhile, the corresponding practical heat removal curve is constructed. A case study shows that the high temperature heat removal ratio within the main fractionator increased by 8%. The proposed method offers heat removal inequality constraints for the model to optimize the heat integration between the main fractionator and the heat exchanger network. - Highlights: • A novel method defines the heat removal constraints of the main fractionator. • Fractionating precision diagram and column grand composite curve are combined. • The results are the inequality constraints in a simultaneous optimization model

Investigation of characteristics of passive heat removal system based on the assembled heat transfer tube

Energy Technology Data Exchange (ETDEWEB)

Wu, Xiang Cheng; Yan, Changqi; Meng, Zhao Ming; Chen, Kailun; Song, Shao Chuang; Yang, Zong Hao; Yu, Jie [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

2016-12-15

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450 .deg. C to 700 .deg. C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

Investigation of Characteristics of Passive Heat Removal System Based on the Assembled Heat Transfer Tube

Directory of Open Access Journals (Sweden)

Xiangcheng Wu

2016-12-01

Full Text Available To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450°C to 700°C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

International Nuclear Information System (INIS)

WILLIS, W.L.

2000-01-01

This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein

Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

Energy Technology Data Exchange (ETDEWEB)

WILLIS, W.L.

2000-06-15

This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

Analysis of decay heat removal by natural convection in PFBR

International Nuclear Information System (INIS)

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

1993-01-01

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

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

Energy Technology Data Exchange (ETDEWEB)

Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-06-15

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

International Nuclear Information System (INIS)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

Energy Technology Data Exchange (ETDEWEB)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance.

Feasibility of passive heat removal systems

Energy Technology Data Exchange (ETDEWEB)

Ashurko, Yu M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1996-12-01

This paper presents a review of decay heat removal systems (DHRSs) used in liquid metal-cooled fast reactors (LMFRs). Advantages and the disadvantages of these DHRSs, extent of their passivity and prospects for their use in advanced fast reactor projects are analyzed. Methods of extending the limitations on the employment of individual systems, allowing enhancement in their effectiveness as safety systems and assuring their total passivity are described. (author). 10 refs, 10 figs.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-02-01

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

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

International Nuclear Information System (INIS)

1993-02-01

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

Air-side performance evaluation of three types of heat exchangers in dry, wet and periodic frosting conditions

Energy Technology Data Exchange (ETDEWEB)

Zhang, Ping [Zhejiang Vocational College of Commerce, Hangzhou, Binwen Road 470 (China); Department of Mechanical Science and Engineering University of Illinois at Urbana-Champaign 1206 West Green Street, Urbana, IL 61801 (United States); Hrnjak, P.S. [Department of Mechanical Science and Engineering University of Illinois at Urbana-Champaign 1206 West Green Street, Urbana, IL 61801 (United States)

2009-08-15

The performances of three types of heat exchangers that use the louver fin geometry: (1) parallel flow parallel fin with extruded flat tubes heat exchanger (PF{sup 2}), (2) parallel flow serpentine fin with extruded flat tubes heat exchanger (PFSF) and (3) round tube wave plate fin heat exchanger (RTPF) have been experimentally studied under dry, wet and frost conditions and results are presented. The parameters quantified include air-side pressure drop, water retention on the surface of the heat exchanger, capacity and overall heat transfer coefficient for air face velocity 0.9, 2 and 3 m/s, air humidity 70% and 80% and different orientations. The performances of three types of heat exchanger are compared and the results obtained are presented. The condensate drainage behavior of the air-side surface of these three heat exchanger types was studied using both the dip testing method and wind tunnel experiment. (author)

Device for removing alkali metal residues from heat exchanger

International Nuclear Information System (INIS)

Matal, O.

1987-01-01

The main parts of the facility consists of a condensing vessel and a vacuum pump unit interconnected via a vacuum pipe. The heat exchanger is heated to a temperature at which the alkali metal residues evaporate. Metal vapors are collected in the condensing vessel where they condense. The removal of the alkali metal residues from the heat exchanger pipes allows thorough inspection of the pipe inside during scheduled nuclear power plant shutdowns. The facility can be used especially with reverse steam generators. (E.S.). 1 fig

After heat removing system of a nuclear reactor

International Nuclear Information System (INIS)

Hayashi, Takao; Yamada, Masao; Ohashi, Kazutaka.

1994-01-01

In a variable conductance heat pipe of an after heat removing system, an evaporation portion and a condensator are connected by a steam diffusing path for an operation fluid and a liquid condensate recycling path. Further, incondensible gases are sealed at the inside together with the operation fluid, and a gas reservoir for the incondensible gases is disposed at the downstream of a condensation portion. If heat input is applied to the evaporation portion of the heat pipe, the incondensible gases are separated to form a boundary between both of them. When the amount of heat applied is small, the incondensible gases partially seal the condensation portion to form a local condensation insensitive portion, so that a heat conductance can be suppressed low. On the other hand, as the amount of heat inputted is increased, the incondensible gases are compressed, the heat conduction area of the condensation portion is increased and a heat conductance is increased to conduct self-control so as to increase heat transfer performance of the heat pipe. Then, the liquid condensate is recycled to the evaporation portion by spontaneous dripping of the condensate itself without wick, thereby enabling to conduct automatic switching so as to increase the heat dissipation amount to maximum. (N.H.)

Evaporation and condensation devices for passive heat removal systems in nuclear power engineering

International Nuclear Information System (INIS)

Gershuni, A.N.; Pis'mennyj, E.N.; Nishchik, A.P.

2016-01-01

The paper justifies advantages of evaporation and condensation heat transfer devices as means of passive heat removal and thermal shielding in nuclear power engineering. The main thermophysical factors that limit heat transfer capacity of evaporation and condensation systems have been examined in the research. The results of experimental studies of heat engineering properties of elongated (8-m) vertically oriented evaporation and condensation devices (two-phase thermosyphons), which showed a high enough heat transfer capacity, as well as stability and reliability both in steady state and in start-up modes, are provided. The paper presents the examples of schematic designs of evaporation and condensation systems for passive heat removal and thermal shielding in application to nuclear power equipment

Excessive heat removal due to feedwater system malfunction

International Nuclear Information System (INIS)

Beader, D.; Peterlin, G.

1986-01-01

Excessive heat removal transient of the Krsko Nuclear Power Plant, caused by steam generators feedwater system malfunctions was simulated by RELAP5/MOD1 computer code. The results are increase of power and reactor scram caused by high-high steam generator level. (author)

Hydrodynamical tests with an original PWR heat removal pump

International Nuclear Information System (INIS)

Wietstock, P.

1984-01-01

GKSS-Forschungszentrum performes hydrodynamical tests with an original PWR heat removal pump to analyse the influences of fluid parameters on the capacity and cavitation behavior of the pump in order to get further improvements of the quantification of the reached safety-level. It can be concluded, that in case of the tested heat removal pump the additional loads during transition from cavitation free operation into fully cavitation for the investigated operation point with 980 m 3 /h will be smaller than the alteration of loads during passing through the total characteristic. The results from cavitation tests for other operation points indicate, that this very important consequence especially for accident operation will be valid for the total specified pump flow area. (orig.)

Evaluation on nitrogen oxides and nanoparticle removal and nitrogen monoxide generation using a wet-type nonthermal plasma reactor

Science.gov (United States)

Takehana, Kotaro; Kuroki, Tomoyuki; Okubo, Masaaki

2018-05-01

Nitrogen oxides (NOx) emitted from power plants and combustion sources cause air pollution problems. Selective catalytic reduction technology is remarkably useful for NOx removal. However, there are several drawbacks such as preparation of reducing agents, usage of harmful heavy metals, and higher cost. On the other hand, trace NO is a vasodilator agent and employed in inhalation therapies for treating pulmonary hypertension in humans. Considering these factors, in the present study, a wet-type nonthermal plasma reactor, which can control NOx and nanoparticle emissions and generate NO, is investigated. The fundamental characteristics of the reactor are investigated. First, the experiment of nanoparticle removal is carried out. Collection efficiencies of over 99% are achieved for nanoparticles at 50 and 100 ml min‑1 of liquid flow rates. Second, experiments of NOx removal under air atmosphere and NOx generation under nitrogen atmosphere are carried out. NOx-removal efficiencies of over 95% under the air plasma are achieved in 50–200 ml min‑1 liquid flow rates. Moreover, under nitrogen plasma, NOx is generated, of which the major portion is NO. For example, NO concentration is 25 ppm, while NOx concentration is 31 ppm at 50 ml min‑1 liquid flow rate. Finally, experiments of NO generation under the nitrogen atmosphere with or without flowing water are carried out. When water flows on the inner surface of the reactor, approximately 14 ppm of NO is generated. Therefore, NO generation requires flowing water. It is considered that the reaction of N and OH, which is similar to the extended Zeldovich mechanism, could occur to induce NO formation. From these results, it is verified that the wet-type plasma reactor is useful for NOx removal and NO generation under nitrogen atmosphere with flowing water.

Wet torrefaction of microalga Chlorella vulgaris ESP-31 with microwave-assisted heating

International Nuclear Information System (INIS)

Bach, Quang-Vu; Chen, Wei-Hsin; Lin, Shih-Cheng; Sheen, Herng-Kuang; Chang, Jo-Shu

2017-01-01

Highlights: • A microwave-assisted heating system is used for wet torrefaction (WT) of microalga. • Microalga Chlorella vulgaris ESP-31 is adopted as the feedstock. • The ash content in the microalga is reduced after WT. • The calorific value of the microalga can be intensified up to 21% after WT. • At least 61.5% of energy in the biomass is retained after WT. - Abstract: Microalgae are a prime source of third generation biofuels. Many thermochemical processes can be applied to convert them into fuels and other valuable products. However, some types of microalgae are characterized by very high moisture and ash contents, thereby causing several problems in further conversion processes. This study presents wet torrefaction (WT) as a promising pretreatment method to overcome the aforementioned drawbacks coupled with microalgal biomass. For this purpose, a microwave-assisted heating system was used for WT of microalga Chlorella vulgaris ESP-31 at different reaction temperatures (160, 170, and 180 °C) and durations (5, 10, and 30 min). The results show several improvements in the fuel properties of the microalga after WT such as increased calorific value and hydrophobicity as well as reduced ash content. A correlation in terms of elemental analysis can be adopted to predict the higher heating value of the torrefied microalga. The structure analysis by Fourier transform infrared (FT-IR) spectroscopy reveals that the carbohydrate content in the torrefied microalgae is lowered, whereas their protein and lipid contents are increased if the WT extent is not severe. However, the protein and lipid contents are reduced significantly at more severe WT conditions. The thermogravimetric analysis shows that the torrefied microalgae have lower ignition temperatures but higher burnout temperatures than the raw microalga, revealing significant impact of WT on the combustion reactivity of the microalga. Overall, the calorific value of the microalga can be intensified up to

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

Directory of Open Access Journals (Sweden)

Giacomino Bandini

2008-01-01

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

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

International Nuclear Information System (INIS)

Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

2000-04-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Development of an integrated MOX-scrap recycling flow-sheet by dry and wet routes using microwave heating techniques

Energy Technology Data Exchange (ETDEWEB)

Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur (India)

1999-01-01

A simple, short and efficient scrap, recycling flow-sheet, which is exclusively based on microwave heating techniques and, includes both dry and wet routes, for (U,Pu)O{sub 2} fuel scrap recycling has been developed and evaluated. (author) 6 refs., 1 tab.

Thermal performance of a spirally coiled finned tube heat exchanger under wet-surface conditions

International Nuclear Information System (INIS)

Wongwises, Somchai; Naphon, Paisarn

2006-01-01

This paper is a continuation of the author's previous work on spiral coil heat exchangers. In the present study, the heat transfer characteristics and the performance of a spirally coiled finned tube heat exchanger under wet-surface conditions are theoretically and experimentally investigated. The test section is a spiral-coil heat exchanger which consists of a steel shell and a spirally coiled tube unit. The spiral-coil unit consists of six layers of concentric spirally coiled finned tubes. Each tube is fabricated by bending a 9.6 mm diameter straight copper tube into a spiral-coil of four turns. The innermost and outermost diameters of each spiral-coil are 145.0 and 350.4 mm, respectively. Aluminium crimped spiral fins with thickness of 0.6 mm and outer diameter of 28.4 mm are placed around the tube. The edge of fin at the inner diameter is corrugated. Air and water are used as working fluids in shell side and tube side, respectively. The experiments are done under dehumidifying conditions. A mathematical model based on the conservation of mass and energy is developed to simulate the flow and heat transfer characteristics of working fluids flowing through the heat exchanger. The results obtained from the present model show reasonable agreement with the experimental data

Heating Changes Bio-Schwertmannite Microstructure and Arsenic(III Removal Efficiency

Directory of Open Access Journals (Sweden)

Xingxing Qiao

2017-01-01

Full Text Available Schwertmannite (Sch is an efficient adsorbent for arsenic(III removal from arsenic(III-contaminated groundwater. In this study, bio-schertmannite was synthesized in the presence of dissolved ferrous ions and Acidithiobacillus ferrooxidans LX5 in a culture media. Bio-synthesized Sch characteristics, such as total organic carbon (TOC, morphology, chemical functional groups, mineral phase, specific surface area, and pore volume were systematically studied after it was dried at 105 °C and then heated at 250–550 °C. Differences in arsenic(III removal efficiency between 105 °C dried-sch and 250–550 °C heated-sch also were investigated. The results showed that total organic carbon content in Sch and Sch weight gradually decreased when temperature increased from 105 °C to 350 °C. Sch partly transformed to another nanocrystalline or amorphous phase above 350 °C. The specific surface area of 250 °C heated-sch was 110.06 m2/g compared to 5.14 m2/g for the 105 °C dried-sch. Total pore volume of 105 °C dried-sch was 0.025 cm3/g with 32.0% mesopore and 68.0% macropore. However, total pore volume of 250 °C heated-mineral was 0.106 cm3/g with 23.6% micropore, 33.0% mesopore, and 43.4% macropore. The arsenic(III removal efficiency from an initial 1 mg/L arsenic(III solution (pH 7.5 was 25.1% when 0.25 g/L of 105 °C dried-sch was used as adsorbent. However, this efficiency increased to 93.0% when using 250 °C heated-sch as adsorbent. Finally, the highest efficiency for arsenic(III removal was obtained with sch-250 °C due to high amounts of sorption sites in agreement with the high specific surface area (SSA obtained for this sample.

Study on concrete cask for practical use. Heat removal test under normal condition

International Nuclear Information System (INIS)

Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

2005-01-01

In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

Wet gas sampling

Energy Technology Data Exchange (ETDEWEB)

Welker, T.F.

1997-07-01

The quality of gas has changed drastically in the past few years. Most gas is wet with hydrocarbons, water, and heavier contaminants that tend to condense if not handled properly. If a gas stream is contaminated with condensables, the sampling of that stream must be done in a manner that will ensure all of the components in the stream are introduced into the sample container as the composite. The sampling and handling of wet gas is extremely difficult under ideal conditions. There are no ideal conditions in the real world. The problems related to offshore operations and other wet gas systems, as well as the transportation of the sample, are additional problems that must be overcome if the analysis is to mean anything to the producer and gatherer. The sampling of wet gas systems is decidedly more difficult than sampling conventional dry gas systems. Wet gas systems were generally going to result in the measurement of one heating value at the inlet of the pipe and a drastic reduction in the heating value of the gas at the outlet end of the system. This is caused by the fallout or accumulation of the heavier products that, at the inlet, may be in the vapor state in the pipeline; hence, the high gravity and high BTU. But, in fact, because of pressure and temperature variances, these liquids condense and form a liquid that is actually running down the pipe as a stream or is accumulated in drips to be blown from the system. (author)

Passive decay heat removal by natural circulation

International Nuclear Information System (INIS)

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

1990-01-01

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

Control of the ASTRA decay heat removal system

International Nuclear Information System (INIS)

Nedelik, A.

1982-11-01

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

Design of CAREM-25 Residual Heat Removal System: Nuclear Safety Aspects

International Nuclear Information System (INIS)

Zanocco, Pablo; Gimenez, Marcelo; Schlamp, Miguel; Barrera, M.

2000-01-01

In this paper Carem-25 residual heat removal system (RHRS) design is analyzed from the nuclear safety point of view.The proposed RHRS is a condenser that transfers the heat to a pool located in the upper level of the containment.The RHRS design basis accident is a reactor loss of heat sink.The following requirements were settled to be verified: a) To remove 2 MW, for a primary circuit pressure of 12.25 MPa and a pool temperature of 100 0 C. b) No condenser tubes flooding, for a primary circuit pressure of 14 MPa and a pool temperature of 100 0 C. c) To reach hot shutdown in 48-hrs, that is to remove of 0.6 MW for a primary circuit pressure of 2.3 MPa and a pool temperature of 120 0 C.Heat transfer regimes inside and outside the condenser and flow patterns were analyzed.Steady state conditions for the above design conditions were modeled.The design requirements were verified taking into account heat transfer coefficients uncertainties and their propagation to the equipment elevation in the containment over the RPV, in order to minimize its elevation and its possible flooding.The resulting condenser tubes were 2 S CH 160 TP 347 SS, with a total area of 4 m 2 and a required minimum height of 6 m from the RPV water level to the condenser outlet headers

Dust emission from wet, low-emission coke quenching process

Science.gov (United States)

Komosiński, Bogusław; Bobik, Bartłomiej; Konieczny, Tomasz; Cieślik, Ewelina

2018-01-01

Coke plants, which produce various types of coke (metallurgical, foundry or heating), at temperatures between 600 and 1200°C, with limited access to oxygen, are major emitters of particulates and gaseous pollutants to air, water and soils. Primarily, the process of wet quenching should be mentioned, as one of the most cumbersome. Atmospheric pollutants include particulates, tar substances, organic pollutants including B(a)P and many others. Pollutants are also formed from the decomposition of water used to quench coke (CO, phenol, HCN, H2S, NH3, cresol) and decomposition of hot coke in the first phase of quenching (CO, H2S, SO2) [1]. The development of the coke oven technology has resulted in the changes made to different types of technological installations, such as the use of baffles in quench towers, the removal of nitrogen oxides by selective NOx reduction, and the introduction of fabric filters for particulates removal. The BAT conclusions for coke plants [2] provide a methodology for the measurement of particulate emission from a wet, low-emission technology using Mohrhauer probes. The conclusions define the emission level for wet quenching process as 25 g/Mgcoke. The conducted research was aimed at verification of the presented method. For two of three quench towers (A and C) the requirements included in the BAT conclusions are not met and emissions amount to 87.34 and 61.35 g/Mgcoke respectively. The lowest particulates emission was recorded on the quench tower B and amounted to 22.5 g/Mgcoke, therefore not exceeding the requirements.

Residual Heat Removal System qualitative probabilistic safety analysis before and after auto closure interlock removal

International Nuclear Information System (INIS)

Mikulicic, V.; Simic, Z.

1992-01-01

The analysis evaluates the consequences of the removal of the auto closure interlock (ACI) on the Residual Heat Removal System (RHRS) suction/isolation valves at the nuclear power plant. The deletion of the RHRS ACI is in part based on a probabilistic safety analysis (PSA) which justifies the removal based on a criterion of increased availability and reliability. Three different areas to be examined in PSA: the likelihood of an interfacing system LOCA; RHRS availability and reliability; and low temperature overpressurization control. The paper emphasizes particularly the RHRS unavailability and reliability evaluation utilizing the current control circuitry configuration and then with the proposed modification to the control circuitry. (author)

After-heat removing system in FBR type reactor

International Nuclear Information System (INIS)

Goto, Tadashi; Inoue, Kotaro; Yamakawa, Masanori; Ikeda, Takashi.

1988-01-01

Purpose: To promote more positive forcive circulation of primary circuit fluids thereby increase the heat removing amount. Constitution: The primary side of an electromagnetic flow coupler type heat exchanger is opened to the primary fluid of a reactor, while the secondary side is connected with the secondary circuit comprising an air cooler and an electromagnetic pump. Since the secondary circuit stands-by during normal operation, the electromagnetic flow coupler does not operate and does not generate force for flowing primary circuit fluid. If flow due to the external force to the primary circuit fluid should occur in the electromagnetic flow coupler type heat exchanger, an electromagnetic force tending to flow the secondary circuit fluid is exerted oppositely. However the coupler undergoes reaction inertia of the fluid or flowing resistance, to exert in the direction of suppressing the flow, thereby prevent the heat loss. (Yoshihara, H.)

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

International Nuclear Information System (INIS)

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

1976-01-01

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

The results of the measurements of mass- and heat-transfer in the wet cooling tower

International Nuclear Information System (INIS)

Fabjan, Lj.; Gaspersic, B.

1979-01-01

These are the results of our investigations carried out on a packing inside a wet cooling tower for the purpose of studying the mass and heat transfer at the counterflow of water and humid air. The measurements on the experimental tower of the corresponding mathematical model reflect the average coefficient of mass and heat transfer for the unity of the active volume. Further the measurements of pressure drop at the air flow were carried out and thus the coefficient of aerodynamic losses were obtained. The results of measurements are given in the corresponding equations with the dimensionless numbers and diagrams. They will be of great use for the planning of new cooling towers. (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)

Pain Relief with Wet Cupping Therapy in Rats is Mediated by Heat Shock Protein 70 and ß-Endorphin.

Science.gov (United States)

Subadi, Imam; Nugraha, Boya; Laswati, Hening; Josomuljono, Harjanto

2017-07-01

Wet cupping therapy is a complementary therapy in pain management. The mechanism of this therapy, however, needs further elucidation. Cells injured by wet cupping therapy seem to stimulate the expression of heat shock protein 70 (HSP70). Its benefit in pain reduction could be mediated by the expression of ß-endorphin. This study aimed at determining the correlation between HSP70 and ß-endorphin after wet cupping therapy. Sixteen male Wistar rats were divided into control (CG; n=8) and treatment (TG; n=8) groups. The rats in both groups were injected with complete Freund's adjuvant (CFA) at the footpad. In the TG, wet cupping therapy was done at the left and right paralumbar regions 48 hours after the CFA injection. Twenty-four hours after therapy, the hot plate test was done to assess pain threshold. Thereafter, immunohistochemistry from the skin subjected to wet cupping therapy was conducted for HSP70 and ß-endorphin. The expression of HSP70 was significantly higher in the keratinocytes of the TG (20.25±3.53; Pcupping therapy was significantly higher in the TG (22.81±6.34 s; P=0.003) than in the CG (11.78±3.56 s). The benefit of wet cupping therapy in terms of pain reduction in rats could be mediated by the expression of HSP70 and ß-endorphin.

Improved of effective wetting area and film thickness on a concentric helical bank of a generator for an absorption heat transformer

International Nuclear Information System (INIS)

Lazcano-Véliz, Y.; Hernández, J.A.; Juárez-Romero, D.; Huicochea-Rodríguez, A.; Álvarez-Gallegos, A.; Siqueiros, J.

2016-01-01

Highlights: • The falling film of lithium bromide - water was improved in the generator of an AHT. • The design of the distributor and the concentric tube helical bank was modified. • Wetting efficiency for different operation conditions was obtained. • Improved operation flow in the heat exchange system was determined. - Abstract: This work was performed in the generator of an absorption heat transformer (AHT) applied for water purification, which has two concentric helical coils. The AHT used LiBr-H_2O to 55%wt for the heat transfer through a heat exchange by falling film. The objective of this study was to define the operating condition of the generator. Different falling film regimes were analyzed: drop mode, liquid column, and jet mode. The effective area of heat transfer of the two helical coils, wetting efficiency, and experimental film thickness were determined for the four operating flows (0.003, 0.008, 0.014, 0.019 kg/s) through digital image processing. The theoretical film thickness was measured and compared with the one calculated by the Nusselt equation. The flow of 0.008 kg/s maintained a drop mode distribution favoring a homogeneous fall along the helical test bank. A wetting efficiency of 99.52% was obtained, so it is proposed as operating flow in the generator. The theoretical film thickness for this flow was 0.289 cm and the one obtained experimentally through digital image processing was 0.293 cm. It was concluded that the distribution in the drop mode was more favorable for a better efficiency in the values of the falling film exchangers.

Separately removable tubes in heavy duty heat exchanger assemblies

International Nuclear Information System (INIS)

Neudeck, G.T.

1980-01-01

The invention is directed to removable heat exchanger tube assemblies in heavy duty equipment radiators in which the tubes are each separately removable if they become defective in service. An inwardly facing annular ledge or abutment is molded into the inside diameter of each upper and lower sealing member to receive the respective ends of the tubes and prevent vertical movement of the tubes in service. A flange or shoulder is also provided on the lower portions of each tube and engages the inside of the lower sealing member to further restrain downward movement of the tubes in service. Each tube may be removed by pushing the tube upwardly to overcome the upper ledge abutment and thereby lift the tube free of the lower seal. Each tube may then be removed sidewise from the radiator. Variations of the removable sealing arrangement can be made and are described herein

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

Energy Technology Data Exchange (ETDEWEB)

Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

2014-12-15

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

International Nuclear Information System (INIS)

Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

2014-01-01

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Characteristics of wetting temperature during spray cooling

International Nuclear Information System (INIS)

Mitsutake, Yuichi; Monde, Masanori; Hidaka, Shinichirou

2006-01-01

An experimental study has been done to elucidate the effects of mass flux and subcooling of liquid and thermal properties of solid on the wetting temperature during cooling of a hot block with spray. A water spray was impinged at one of the end surfaces of a cylindrical block initially heated at 400 or 500degC. The experimental condition was mass fluxes G=1-9 kg/m 2 s and degrees of subcooling ΔT sub =20, 50, 80 K. Three blocks of copper, brass and carbon steel were prepared. During spray cooling internal block temperature distribution and sputtering sound pressure level were recorded and the surface temperature and heat flux were evaluated with 2D inverse heat conducting analysis. Cooling process on cooling curves is divided into four regimes categorized by change in a flow situation and the sound level. The wetting temperature defined as the wall temperature at a minimum heat flux point was measured over an extensive experimental range. The wetting wall temperature was correlated well with the parameter of GΔT sub . The wetting wall temperature increases as GΔT sub increases and reaches a constant value depending on the material of the surface at higher region of GΔT sub . (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Method for removal of decay heat of radioactive substances

International Nuclear Information System (INIS)

Hesky, H.; Wunderer, A.

1981-01-01

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

Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

Science.gov (United States)

Bersano, Andrea

With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

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

International Nuclear Information System (INIS)

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

1982-06-01

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

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Schulz, T.L.; Corletti, M.M.

1994-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

Possibility of a pressurized water reactor concept with highly inherent heat removal following capability

International Nuclear Information System (INIS)

Araya, Fumimasa; Murao, Yoshio

1995-01-01

If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such 'highly inherent heat removal following capability', transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code. The possibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10x10 3 kg·cm 3 which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR. (author)

Simplified analysis of passive residual heat removal systems for small size PWR's

International Nuclear Information System (INIS)

Botelho, D.A.

1992-02-01

The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

Determination of optimal wet ethanol composition as a fuel in spark ignition engine

International Nuclear Information System (INIS)

Fagundez, J.L.S.; Sari, R.L.; Mayer, F.D.; Martins, M.E.S.; Salau, N.P.G.

2017-01-01

Highlights: • Batch distillation to produce HEF and fuel blends of wet ethanol. • Conversion efficiency of a SI engine operating with HEF and wet ethanol. • NEF as a new metric to calculate the energy efficiency of HEF and wet ethanol. • Optimal wet ethanol composition as a fuel in SI engine based on NEF. - Abstract: Studies are unanimous that the greatest fraction of the energy necessary to produce hydrous ethanol fuel (HEF), i.e. above 95%v/v of ethanol in water, is spent on water removal (distillation). Previous works have assessed the energy efficiency of HEF; but few, if any, have done the same for wet ethanol fuel (sub-azeotropic hydrous ethanol). Hence, a new metric called net energy factor (NEF) is proposed to calculate the energy efficiency of wet ethanol and HEF. NEF calculates the ratio of Lower Heating Value (LHV) derived from ethanol fuel, total energy out, to energy used to obtain ethanol fuel as distillate, total energy in. Distillation tests were performed batchwise to obtain as distillate HEF and four different fuel blends of wet ethanol with a range from 60%v/v to 90%v/v of ethanol and the amount of energy spent to distillate each ethanol fuel calculated. The efficiency parameters of a SI engine operating with the produced ethanol fuels was tested to calculate their respective conversion efficiency. The results of net energy factors show a clear advantage of wet ethanol fuels over HEF; the optimal efficiency was wet ethanol fuel with 70%v/v of ethanol.

Sodium removal of fuel elements by vacuum distillation

International Nuclear Information System (INIS)

Buescher, E.; Haubold, W.; Jansing, W.; Kirchner, G.

1978-01-01

Cleaning of sodium-wetted core components can be performed by using either lead, moist nitrogen, or alcohol. The advantages of these methods for cleaning fuel elements without causing damage are well known. The disadvantage is that large amounts of radioactive liquids are formed during handling in the latter two cases. In this paper a new method to clean components is described. The main idea is to remove all liquid metal from the core components within a comparatively short period of time. Fuel elements removed from the reactor must be cooled because of high decay heat release. To date, vacuum distillation of fuel elements has not yet been applied

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

International Nuclear Information System (INIS)

Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

2013-01-01

The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

Phycoremediation of Heavy Metals in Wet Market Wastewater

Science.gov (United States)

Apandi, Najeeha; Saphira Radin Mohamed, Radin Maya; Al-Gheethi, Adel; Latiffi, Atikah; Nor Hidayah Arifin, Siti; Gani, Paran

2018-04-01

The efficiency of phycoremediation using microalgae for removing nutrients and heavy metals from wastewaters has been proved. However, the differences in the composition of wastewaters as well as microalgae species play an important role in the efficient of this process. Therefore, the present study aimed to investigate the effectiveness of Scenedesmus sp. to removal of heavy metals from wet market wastewater. Scenedesmus sp. was inoculated with 106 cells/mL into each wet market wastewater concentration included 10, 25, 50, 75 and 100% and incubated for 18 days. The highest growth rate was recorded in 50% WM with a maximum dry weight of 2006 mg L-1 which subsequently removed 93.06% of Cd, 91.5% of Cr, 92.47% of Fe, 92.40% of Zn. These findings reflected the high potential of Scenedesmus sp. in the treatment of wet market wastewater and production microalgae biomass.

Effects of infrared radiation heating on peeling performance and quality attributes of clingstone peaches

Science.gov (United States)

Salinity and wastewater disposal problems associated with the conventional wet-lye method for peeling clingstone peaches result in considerable negative environment impacts. This study investigated the efficacy of using infrared (IR) heating as an alternative method for peach peel removal without us...

Solid state de-wetting observed for vapor deposited copper films on carbon substrates

International Nuclear Information System (INIS)

Schrank, C.; Eisenmenger-Sittner, C.; Neubauer, E.; Bangert, H.; Bergauer, A.

2004-01-01

Copper-Carbon composites are a good example for novel materials consisting of components with extremely different physical and chemical properties. They have a high potential for an application as heat sinks for electronic components, but the joining of the two materials is a difficult task. To obtain reasonable mechanical and thermal contact between copper and carbon the following route was chosen. First glassy-carbon substrates were subjected to an RF-Nitrogen plasma treatment. Then 300 nm thick copper coatings were sputter-deposited on the plasma treated surface within the same vacuum chamber. Finally, the samples were removed from the deposition chamber and either investigated immediately or thermally annealed at 850 deg. C under high vacuum conditions (10 -4 Pa). While non-annealed copper-coatings were continuous and showed excellent adhesion values of approximately 700 N/cm 2 , the heat treated samples lose their continuity by a de-wetting process. At the beginning holes are formed, then a labyrinth-like morphology develops and finally the coating consists of isolated droplets. All these processes occur well below the melting temperature of copper and were observed by AFM and SEM. The mechanism of this solid-state de-wetting process is investigated in relation to the recent literature on de-wetting and its consequences on the manufacturing of copper-carbon composites are discussed

Photovoltaic cell electrical heating system for removing snow on panel including verification.

Science.gov (United States)

Weiss, Agnes; Weiss, Helmut

2017-11-16

Small photovoltaic plants in private ownership are typically rated at 5 kW (peak). The panels are mounted on roofs at a decline angle of 20° to 45°. In winter time, a dense layer of snow at a width of e.g., 10 cm keeps off solar radiation from the photovoltaic cells for weeks under continental climate conditions. Practically, no energy is produced over the time of snow coverage. Only until outside air temperature has risen high enough for a rather long-time interval to allow partial melting of snow; the snow layer rushes down in an avalanche. Following this proposal, snow removal can be arranged electrically at an extremely positive energy balance in a fast way. A photovoltaic cell is a large junction area diode inside with a threshold voltage of about 0.6 to 0.7 V (depending on temperature). This forward voltage drop created by an externally driven current through the modules can be efficiently used to provide well-distributed heat dissipation at the cell and further on at the glass surface of the whole panel. The adhesion of snow on glass is widely reduced through this heating in case a thin water film can be produced by this external short time heating. Laboratory experiments provided a temperature increase through rated panel current of more than 10 °C within about 10 min. This heating can initiate the avalanche for snow removal on intention as described before provided the clamping effect on snow at the edge of the panel frame is overcome by an additional heating foil. Basics of internal cell heat production, heating thermal effects in time course, thermographic measurements on temperature distribution, power circuit opportunities including battery storage elements and snow-removal under practical conditions are described.

Removal of corrosion products of construction materials in heat carrier

International Nuclear Information System (INIS)

1975-01-01

A review of reported data has been made on the removal of structural material corrosion products into the heat-carrying agent of power reactors. The corrosion rate, and at the same time, removal of corrosion products into the heat-carrying agent (water) decreases with time. Thus, for example, the corrosion rate of carbon steel in boiling water at 250 deg C and O 2 concentration of 0.1 mg/1 after 3000 hr is 0.083 g/m 2 . day; after 9000 hr the corrosion rate has been reduced 2.5 times. Under static conditions the transfer rate of corrosion products into water has been smaller than in the stream and also depends on time. The corrosion rate of carbon steel under nuclear plant operating conditions is almost an order higher over that of steel Kh18N10T

Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

International Nuclear Information System (INIS)

Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

2015-01-01

As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

Engineering and economic evaluation of wet/dry cooling towers for water conservation

International Nuclear Information System (INIS)

Hu, M.C.

1976-11-01

The results are presented of a design and cost study for wet/dry tower systems used in conjunction with 1000 MWe nuclear power plants to reject waste heat while conserving water. Design and cost information for wet/dry tower systems are presented, and these cooling system alternatives are compared with wet and dry tower systems to determine whether the wet/dry tower concept is an economically viable alternative. The wet/dry cooling tower concept investigated is one which combines physically separated wet towers and dry towers into an operational unit. In designing the wet/dry tower, a dry cooling tower is sized to carry the plant heat load at low ambient temperatures, and a separate wet tower is added to augment the heat rejection of the dry tower at higher ambient temperatures. These wet/dry towers are designed to operate with a conventional low back pressure turbine commercially available today. The component wet and dry towers are state-of-the-art designs. From this study it was concluded that: wet/dry cooling systems can be designed to provide a significant economic advantage over dry cooling yet closely matching the dry tower's ability to conserve water, a wet/dry system which saves as much as 99 percent of the make-up water required by a wet tower can maintain that economic advantage, and therefore, for power plant sites where water is in short supply, wet/dry cooling is the economic choice over dry cooling

Natural convection as the way of heat removal from fast reactor core at cooldown regimes

International Nuclear Information System (INIS)

Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

2000-01-01

The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

Valve arrangement for a nuclear plant residual heat removal system

International Nuclear Information System (INIS)

Fidler, G.L.; Hill, R.A.; Carrera, J.P.

1978-01-01

Disclosed is an improved valve arrangement for a two-train Residual Heat Removal System (RHRS) of a nuclear reactor plant which ensures operational integrity of the system under single failure circumstances including loss of one of two electrical power sources

RCS pressure under reduced inventory conditions following a loss of residual heat removal

International Nuclear Information System (INIS)

Palmrose, D.E.; Hughes, E.D.; Johnsen, G.W.

1992-01-01

The thermal-hydraulic response of a closed-reactor coolant system to loss of residual heat removal (RHR) cooling is investigated. The processes examined include: core coolant boiling and steam generator reflux condensation, pressure increase on the primary side, heat transfer mechanisms on the steam generator primary and secondary sides, and effects of noncondensible gas on heat transfer processes

Nuclear power plant with improved arrangements for the removal of post fission and emergency heating

International Nuclear Information System (INIS)

Buescher, E.; Vinzens, K.

1977-01-01

This is concerned with additional equipment for emergency heat removal in a sodium cooled reactor, which operates on failure of the post fission heat removal system. The space for pressure relieving spaces and concrete masses as heat sinks within the reactor cell is no longer required. In this nuclear power plant, a heat exchanger chain transmits heat and power: There is a first sodium circuit between pressure vessel and the first heat exchanger, a second one between the first and second heat excahngers, and a third (Steam) circuit with turbine, condenser and return pump. A fourth circuit connects the secondary side of the condenser with a cooling tower. There is a threee component heat excahgner in the primary circuit after the first heat exchanger, which is built around the first heat exchanger, and is sealed into an unloading space. This space is situated next to the reactor cell and is above the operating level of the sodium in the pressure vessel. It is connected to the cell by an upper duct, normally closed by a bursting disc, and by a lower duct. In the three comopnent heat exchanger, a liquid lead-bismuth eutectic mixture transmits the heat from sodium pipes to water pipes. In normal operation it is used for steam superheating or feedwater preheating. The three component heat exchanger bridges the first and second heat exchangers as an emergency heat exchanger. If in such a case the post fission heat removal has failed, the sodium evaporating in the pressure vessel flows into the unloading space and condenses on the ribs of the emergency heat exchanger. The post fission heat is fed by the water secondary medium directly into the tertiary circuit. The sodium condensate flows back from the unloading space via the lower duct into the reactor cell and maintains the emergency level there. (RW) 891 RW [de

A decay heat removal methodology for reuseable orbital transfer vehicles

Science.gov (United States)

McDaniel, Patrick J.; Perkins, David R.

1992-07-01

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

Effect of ultrasound treatment on the wet heating Maillard reaction between mung bean [Vigna radiate (L.)] protein isolates and glucose and on structural and physico-chemical properties of conjugates.

Science.gov (United States)

Wang, Zhongjiang; Han, Feifei; Sui, Xiaonan; Qi, Baokun; Yang, Yong; Zhang, Hui; Wang, Rui; Li, Yang; Jiang, Lianzhou

2016-03-30

The objective of this study was to determine the effect of ultrasound treatment on the wet heating Maillard reaction between mung bean protein isolates (MBPIs) and glucose, and on structural and physico-chemical properties of the conjugates. The degree of glycosylation of MBPI-glucose conjugates treated by ultrasound treatment and wet heating (MBPI-GUH) was higher than that of MBPI-glucose conjugates only treated by wet heating (MBPI-GH). Solubility, emulsification activity, emulsification stability and surface hydrophobicity of MBPI-GUH were higher than that of MBPI-GH. Grafted MBPIs had a lower content of α-helix and unordered coil, but a higher content of β-sheet and β-turn structure than MBPIs. No significant structural changes were observed in β-turn and random coil structure of MBPI-GUH, while α-helix content increased with ultrasonic time, and decreased at 300 W ultrasonic power with the increase of β-sheet. MBPI-GUH had a less compact tertiary structure compared to MBPI-GH and MBPI. Grafting MBPIs with glucose formed conjugates of higher molecular weight, while no significant changes were observed in electrophoresis profiles of MBPI-GUH. Ultrasound-assisted wet heating Maillard reaction between MBPIs and glucose could be a promising way to improve functional properties of MBPIs. © 2015 Society of Chemical Industry.

Efficient on-chip hotspot removal combined solution of thermoelectric cooler and mini-channel heat sink

International Nuclear Information System (INIS)

Hao, Xiaohong; Peng, Bei; Xie, Gongnan; Chen, Yi

2016-01-01

Highlights: • A combined solution of thermoelectric cooler (TEC) and mini-channel heat sink to remove the hotspot of the chip has been proposed. • The TEC's mathematical model is established to assess its work performance. • A comparative study on the proposed efficient On-Chip Hotspot Removal Combined Solution. - Abstract: Hotspot will significantly degrade the reliability and performance of the electronic equipment. The efficient removal of hotspot can make the temperature distribution uniform, and ensure the reliable operation of the electronic equipment. This study proposes a combined solution of thermoelectric cooler (TEC) and mini-channel heat sink to remove the hotspot of the chip in the electronic equipment. Firstly, The TEC's mathematical model is established to assess its work performance under different boundary conditions. Then, the hotspot removal capability of the TEC is discussed for different cooling conditions, which has shown that the combined equipment has better hotspot removal capability compared with others. Finally, A TEC is employed to investigate the hotspot removal capacity of the combined solution, and the results have indicated that it can effectively remove hotspot in the diameter of 0.5 mm, the power density of 600W/cm 2 when its working current is 3A and heat transfer thermal resistance is 0 K/W.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Pain Relief with Wet Cupping Therapy in Rats is Mediated by Heat Shock Protein 70 and ß-Endorphin

Directory of Open Access Journals (Sweden)

Imam Subadi

2017-07-01

Full Text Available Background: Wet cupping therapy is a complementary therapy in pain management. The mechanism of this therapy, however, needs further elucidation. Cells injured by wet cupping therapy seem to stimulate the expression of heat shock protein 70 (HSP70. Its benefit in pain reduction could be mediated by the expression of ß-endorphin. This study aimed at determining the correlation between HSP70 and ß-endorphin after wet cupping therapy. Methods: Sixteen male Wistar rats were divided into control (CG; n=8 and treatment (TG; n=8 groups. The rats in both groups were injected with complete Freund’s adjuvant (CFA at the footpad. In the TG, wet cupping therapy was done at the left and right paralumbar regions 48 hours after the CFA injection. Twenty-four hours after therapy, the hot plate test was done to assess pain threshold. Thereafter, immunohistochemistry from the skin subjected to wet cupping therapy was conducted for HSP70 and ß-endorphin. Results: The expression of HSP70 was significantly higher in the keratinocytes of the TG (20.25±3.53; P<0.001 than in the keratinocytes of the CG (10.50±2.44; P<0.001. The expression of ß-endorphin was significantly higher in the keratinocytes of the TG (22.37±3.52; P<0.001 than in the keratinocytes of the CG (5.12±1.72; P<0.001. The results also revealed a high correlation between HSP70 and ß-endorphin (β=0.864; P<0.001. Pain threshold after wet cupping therapy was significantly higher in the TG (22.81±6.34 s; P=0.003 than in the CG (11.78±3.56 s. Conclusions: The benefit of wet cupping therapy in terms of pain reduction in rats could be mediated by the expression of HSP70 and ß-endorphin.

7 CFR 51.897 - Wet.

Science.gov (United States)

2010-01-01

... the grapes are wet from moisture from crushed, leaking, or decayed berries or from rain. Grapes which are moist from dew or other moisture condensation such as that resulting from removing grapes from a...

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

The heat removal capability of actively cooled plasma-facing components for the ITER divertor

Science.gov (United States)

Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

2011-12-01

Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

The heat removal capability of actively cooled plasma-facing components for the ITER divertor

International Nuclear Information System (INIS)

Missirlian, M; Richou, M; Loarer, T; Riccardi, B; Gavila, P; Constans, S

2011-01-01

Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m - 2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m - 2 for the CFC-armoured tiles and 15 MW m - 2 for the W-armoured tiles, respectively.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Evaluation of wet bulb globe temperature index for estimation of heat strain in hot/humid conditions in the Persian Gulf

OpenAIRE

Habibolah Dehghan; Seyed Bagher Mortazavi; Mohammad J Jafari; Mohammad R Maracy

2012-01-01

Background: Heat exposure among construction workers in the Persian Gulf region is a serious hazard for health. The aim of this study was to evaluate the performance of wet bulb globe temperature (WBGT) Index for estimation of heat strain in hot/humid conditions by the use of Physiological Strain Index (PSI) as the gold standard. Material and Methods : This cross-sectional study was carried out on 71 workers of two Petrochemical Companies in South of Iran in 2010 summer. The WBGT index, heart...

Nuclear turbine efficiency improvement by wet steam study

International Nuclear Information System (INIS)

Nishikawa, Tsuyoshi; Morson, A.; Markytan, R.

2000-01-01

Most of the turbine used at the nuclear power plant are operated at environment of wet steam, which composes of a big factor of its inner loss in comparison with those of the thermal power plant. If an analytical method predictable on behavior of the wet steam is established, it could be upgraded efficiency of the turbine and also reliability against corrosion formed by moisture. This study, therefore, aims at understanding of physical property of the wet steam flow scarcely known at present, development of an optimum turbine cascade design tool reflected by the property, development of a turbine cascade design reducible of steam loss due to wet steam by using the tool, and development on a method of removing moisture in the turbine to its outer portion. For the tool, a new three dimensional flow numerical analysis is necessary to be developed, to aim at accurately and numerically understanding of the behavior of wet steam. As this study is in advancing now, by using a turbine cascade optimized on the wet steam flow and a developed moisture removing apparatus, about 0.6 % of upgrading in turbine efficiency can be predicted in comparison with that of the advanced aero-cascade of the GE Corporation. (G.K.)

Evaluation of wet air oxidation variables for removal of organophosphorus pesticide malathion using Box-Behnken design.

Science.gov (United States)

Isgoren, Melike; Gengec, Erhan; Veli, Sevil

2017-02-01

This paper deals with finding optimum reaction conditions for wet air oxidation (WAO) of malathion aqueous solution, by Response Surface Methodology. Reaction conditions, which affect the removal efficiencies most during the non-catalytic WAO system, are: temperature (60-120 °C), applied pressure (20-40 bar), the pH value (3-7), and reaction time (0-120 min). Those were chosen as independent parameters of the model. The interactions between parameters were evaluated by Box-Behnken and the quadratic model fitted very well with the experimental data (29 runs). A higher value of R 2 and adjusted R 2 (>0.91) demonstrated that the model could explain the results successfully. As a result, optimum removal efficiency (97.8%) was obtained at pH 5, 20 bars of pressure, 116 °C, and 96 min. These results showed that Box-Behnken is a suitable design to optimize operating conditions and removal efficiency for non-catalytic WAO process. The EC 20 value of raw wastewater was measured as 35.40% for malathion (20 mg/L). After the treatment, no toxicity was observed at the optimum reaction conditions. The results show that the WAO is an efficient treatment system for malathion degradation and has the ability of converting malathion to the non-toxic forms.

Passive safety systems for decay heat removal of MRX

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Application of grey model on analyzing the passive natural circulation residual heat removal system of HTR-10

Institute of Scientific and Technical Information of China (English)

ZHOU Tao; PENG Changhong; WANG Zenghui; WANG Ruosu

2008-01-01

Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR (High Temperature gas-cooled Reactor),the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM(1,1) and GM(1, 3) model are built based on the inlet air temperature of chimney, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an important role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR.

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

International Nuclear Information System (INIS)

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

2009-01-01

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

Characterization of Wet-Heat Inactivation of Single Spores of Bacillus Species by Dual-Trap Raman Spectroscopy and Elastic Light Scattering▿

Science.gov (United States)

Zhang, Pengfei; Kong, Lingbo; Setlow, Peter; Li, Yong-qing

2010-01-01

Dual-trap laser tweezers Raman spectroscopy (LTRS) and elastic light scattering (ELS) were used to investigate dynamic processes during high-temperature treatment of individual spores of Bacillus cereus, Bacillus megaterium, and Bacillus subtilis in water. Major conclusions from these studies included the following. (i) After spores of all three species were added to water at 80 to 90°C, the level of the 1:1 complex of Ca2+ and dipicolinic acid (CaDPA; ∼25% of the dry weight of the spore core) in individual spores remained relatively constant during a highly variable lag time (Tlag), and then CaDPA was released within 1 to 2 min. (ii) The Tlag values prior to rapid CaDPA release and thus the times for wet-heat killing of individual spores of all three species were very heterogeneous. (iii) The heterogeneity in kinetics of wet-heat killing of individual spores was not due to differences in the microscopic physical environments during heat treatment. (iv) During the wet-heat treatment of spores of all three species, spore protein denaturation largely but not completely accompanied rapid CaDPA release, as some changes in protein structure preceded rapid CaDPA release. (v) Changes in the ELS from individual spores of all three species were strongly correlated with the release of CaDPA. The ELS intensities of B. cereus and B. megaterium spores decreased gradually and reached minima at T1 when ∼80% of spore CaDPA was released, then increased rapidly until T2 when full CaDPA release was complete, and then remained nearly constant. The ELS intensity of B. subtilis spores showed similar features, although the intensity changed minimally, if at all, prior to T1. (vi) Carotenoids in B. megaterium spores' inner membranes exhibited two changes during heat treatment. First, the carotenoid's two Raman bands at 1,155 and 1,516 cm−1 decreased rapidly to a low value and to zero, respectively, well before Tlag, and then the residual 1,155-cm−1 band disappeared, in parallel

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Experimental investigation on the evaporation of a wet porous layer inside a vertical channel with resolution of the heat equation by inverse method

International Nuclear Information System (INIS)

Terzi, A.; Foudhil, W.; Harmand, S.; Ben Jabrallah, S.

2016-01-01

Highlights: • Experimental study of the evaporation of a wet porous layer inside a vertical channel. • Resolution of the heat equation by inverse method. • The use of the porous layer is more efficient for high heating flux and low liquid inlet flow. • To improve the evaporation, the system must operate at low water inlet flow. - Abstract: In this paper, we realize an Experimental study of the evaporation of a wet porous layer inside a vertical channel. To develop this study, an experimental dispositive was realised. We measure the temperature along the plate and the evaporated flow rate using the test bed. From these measurements we note that the profiles of the temperature are divided into two areas: the heating and the evaporation zone. We also note that the use of the porous layer is more efficient for high heating flux and low liquid inlet flow. In addition, we studied different dimensionless numbers by solving the energy equation by inverse method. We note that the latent Nusselt number is more important than the sensible Nusselt Number, which proves that the flow dissipated by evaporation is greater than the one used by the film to increase its temperature.

Biodiesel production from wet microalgae feedstock using sequential wet extraction/transesterification and direct transesterification processes.

Science.gov (United States)

Chen, Ching-Lung; Huang, Chien-Chang; Ho, Kao-Chia; Hsiao, Ping-Xuan; Wu, Meng-Shan; Chang, Jo-Shu

2015-10-01

Although producing biodiesel from microalgae seems promising, there is still a lack of technology for the quick and cost-effective conversion of biodiesel from wet microalgae. This study was aimed to develop a novel microalgal biodiesel producing method, consisting of an open system of microwave disruption, partial dewatering (via combination of methanol treatment and low-speed centrifugation), oil extraction, and transesterification without the pre-removal of the co-solvent, using Chlamydomonas sp. JSC4 with 68.7 wt% water content as the feedstock. Direct transesterification with the disrupted wet microalgae was also conducted. The biomass content of the wet microalgae increased to 56.6 and 60.5 wt%, respectively, after microwave disruption and partial dewatering. About 96.2% oil recovery was achieved under the conditions of: extraction temperature, 45°C; hexane/methanol ratio, 3:1; extraction time, 80 min. Transesterification of the extracted oil reached 97.2% conversion within 15 min at 45°C and 6:1 solvent/methanol ratio with simultaneous Chlorophyll removal during the process. Nearly 100% biodiesel conversion was also obtained while conducting direct transesterification of the disrupted oil-bearing microalgal biomass. Copyright © 2015 Elsevier Ltd. All rights reserved.

Measurement of wetted area fraction in subcooled pool boiling of water using infrared thermography

International Nuclear Information System (INIS)

Kim, Hyungdae; Park, Youngjae; Buongiorno, Jacopo

2013-01-01

The wetted area fraction in subcooled pool boiling of water at atmospheric pressure is measured using the DEPIcT (DEtection of Phase by Infrared Thermography) technique. DEPIcT exploits the contrast in infrared (IR) light emissions between wet and dry areas on the surface of an IR-transparent heater to visualize the instantaneous distribution of the liquid and gas phases in contact with the heater surface. In this paper time-averaged wetted area fraction data in nucleate boiling are reported as functions of heat flux (from 30% up to 100% of the Critical Heat Flux) and subcooling (ΔT sub = 0, 5, 10, 30 and 50 °C). The results show that the wetted area fraction monotonically decreases with increasing heat flux and increases with increasing subcooling: both trends are expected. The range of time-averaged wetted area fractions is from 90%, at low heat flux and high subcooling, to 50% at high heat flux (right before CHF) and low subcooling. It is also shown that the dry areas are periodically rewetted by liquid sloshing on the surface at any subcooling and heat flux; however, the dry areas expand irreversibly at CHF

Performance of ALMR passive decay heat removal system

International Nuclear Information System (INIS)

Boardman, C.E.; Hunsbedt, A.

1991-01-01

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

Heat Removal from Bipolar Transistor by Loop Heat Pipe with Nickel and Copper Porous Structures

Science.gov (United States)

Smitka, Martin; Malcho, Milan

2014-01-01

Loop heat pipes (LHPs) are used in many branches of industry, mainly for cooling of electrical elements and systems. The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work was to develop porous structures from copper and nickel powder with different grain sizes. For experiment copper powder with grain size of 50 and 100 μm and nickel powder with grain size of 10 and 25 μm were used. Analysis of these porous structures and LHP design are described in the paper. And the measurements' influences of porous structures in LHP on heat removal from the insulated gate bipolar transistor (IGBT) have been made. PMID:24959622

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

International Nuclear Information System (INIS)

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

1996-03-01

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

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

International Nuclear Information System (INIS)

Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

2004-01-01

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

Simultaneous heat and mass transfer to air from a compact heat exchanger with water spray precooling and surface deluge cooling

International Nuclear Information System (INIS)

Zhang, Feini; Bock, Jessica; Jacobi, Anthony M.; Wu, Hailing

2014-01-01

Various methods are available to enhance heat exchanger performance with evaporative cooling. In this study, evaporative mist precooling, deluge cooling, and combined cooling schemes are examined experimentally and compared to model predictions. A flexible model of a compact, finned-tube heat exchanger with a wetted surface is developed by applying the governing conservation and rate equations and invoking the heat and mass transfer analogy. The model is applicable for dry, partially wet, or fully wet surface conditions and capable of predicting local heat/mass transfer, wetness condition, and pressure drop of the heat exchanger. Experimental data are obtained from wind tunnel experiments using a louver-fin flat-tube heat exchanger with single-phase tube-side flow. Total capacity, pressure drop, and water drainage behavior under various water usage rates and air face velocities are analyzed and compared to data for dry-surface conditions. A heat exchanger partitioning method for evaporative cooling is introduced to study partially wet surface conditions, as part of a consistent and general method for interpreting wet-surface performance data. The heat exchanger is partitioned into dry and wet portions by introducing a wet surface factor. For the wet part, the enthalpy potential method is used to determine the air-side sensible heat transfer coefficient. Thermal and hydraulic performance is compared to empirical correlations. Total capacity predictions from the model agree with the experimental results with an average deviation of 12.6%. The model is also exercised for four water augmentation schemes; results support operating under a combined mist precooling and deluge cooling scheme. -- Highlights: • A new spray-cooled heat exchanger model is presented and is validated with data. • Heat duty is shown to be asymptotic with spray flow rate. • Meaningful heat transfer coefficients for partially wet conditions are obtained. • Colburn j wet is lower than j dry

Concepts for passive heat removal and filtration systems under core meltdown conditions

International Nuclear Information System (INIS)

Wilhelm, J.G.; Neitzel, H.-J.

1993-01-01

The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

Analysis of Decay Heat Removal by Natural Convection in LMR with a Combined Steam Generator

International Nuclear Information System (INIS)

Kim, Eui Kwang; Eoh, Jae Hyuk; Han, Ji Woong; Lee, Tae Ho

2011-01-01

Liquid metal reactors (LMRs) conventionally employ an intermediate heat transport system (IHTS) to protect the nuclear core during a sodium-water reaction (SWR) event. However these SWR-related components increase plant construction costs. In order to eliminate the need for an IHTS, a combined steam generator, which is an integrated heat exchanger of a steam generator and intermediate heat exchanger (IHX), was proposed by the Korea Atomic Energy Research Institute (KAERI). The objective of this work is to analyze the natural circulation heat removal capability of the rector system using a combined steam generator. As a means of decay heat removal, a normal heat transport path is composed of a primary sodium system, intermediate lead-bismuth circuit combined with SG and steam/water system. This paper presents the results of the possible temperature and natural circulation flows in all circuits during a steady state for a given reactor power level varied as a function of time

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Removal of contaminated asphalt layers by using heat generating powder metallic systems

International Nuclear Information System (INIS)

Barinov, A.S.; Karlina, O.K.; Ojovan, M.I.

1996-01-01

Heat generating systems on the base of powder metallic fuel were used for the removal of contaminated asphalt layers. Decontamination of spots which had complex geometric form was performed. Asphalt layers with deep contamination were removed essentially all radionuclides being retained in asphalt residue. Only a small part (1 - 2 %) of radionuclides could pass to combustion slag. No radionuclides were detected in aerosol-gas phase during decontamination process

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Proteomic effects of wet cupping (Al-hijamah).

Science.gov (United States)

Almaiman, Amer A

2018-01-01

Wet cupping (Al-hijamah) is a therapeutic technique practiced worldwide as a part of the Unani system of medicine. It involves bloodletting from acupoints on a patient's skin to produce a therapeutic outcome. A thorough review of research articles on wet cupping with relevance to proteomics field that are indexed by Google Scholar, PubMed, and/or Science Direct databases was performed. Eight original research articles were summarized in this paper. Overall, wet cupping did not have a significant effect on C-reactive protein, Hsp-27, sister chromatid exchanges, and cell replication index. In contrast, wet cupping was found to produce higher oxygen saturation, eliminate lactate from subcutaneous tissues, remove blood containing higher levels of malondialdehyde and nitric oxide, and produce higher activity of myeloperoxidase. The proteomic effects of wet cupping therapy have not been adequately investigated. Thus, future studies on wet cupping that use systemic and sound protocols to avoid bias should be conducted.

Proteomic effects of wet cupping (Al-hijamah

Directory of Open Access Journals (Sweden)

Amer A. Almaiman

2018-01-01

Full Text Available Wet cupping (Al-hijamah is a therapeutic technique practiced worldwide as a part of the Unani system of medicine. It involves bloodletting from acupoints on a patient’s skin to produce a therapeutic outcome. A thorough review of research articles on wet cupping with relevance to proteomics field that are indexed by Google Scholar, PubMed, and/or Science Direct databases was performed. Eight original research articles were summarized in this paper. Overall, wet cupping did not have a significant effect on C-reactive protein, Hsp-27, sister chromatid exchanges, and cell replication index. In contrast, wet cupping was found to produce higher oxygen saturation, eliminate lactate from subcutaneous tissues, remove blood containing higher levels of malondialdehyde and nitric oxide, and produce higher activity of myeloperoxidase. The proteomic effects of wet cupping therapy have not been adequately investigated. Thus, future studies on wet cupping that use systemic and sound protocols to avoid bias should be conducted.

EFFECT OF HEAT-DISPERSING ON STICKIES AND THEIR REMOVAL IN POST-FLOTATION

OpenAIRE

Yang Gao,; Menghua Qin,; Hailong Yu,; Fengshan Zhang

2012-01-01

The effect of heat-dispersing on sticky substances in a deinking pulping line was studied under different conditions including varying temperature, disc clearance, and pulp consistency. Sticky substances were quantitatively investigated before and after the heat-dispersing, and categorized into macro-, mini-, and micro-stickies as well as dissolved and colloidal substances. Meanwhile, their extents of removal in post-flotation were evaluated. The results showed that raising temperature, reduc...

Heat removing device for nuclear reactor container facility

Energy Technology Data Exchange (ETDEWEB)

Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

1994-09-30

A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

International Nuclear Information System (INIS)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

1995-01-01

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

Residual heat removal system diagnostic advisor

International Nuclear Information System (INIS)

Tripp, L.

1991-01-01

This paper reports on the Residual Heat Removal System (RHRS) Diagnostic Advisor which is an expert system designed to alert the operators to abnormal conditions that exits in the RHRS and offer advice about the cause of the abnormal conditions. The Advisor uses a combination of rule-based and model-based diagnostic techniques to perform its functions. This diagnostic approach leads to a deeper understanding of the RHRS by the Advisor and consequently makes it more robust to unexpected conditions. The main window of the interactive graphic display is a schematic diagram of the RHRS piping system. When a conclusion about a failed component can be reached, the operator can bring up windows that describe the failure mode of the component and a brief explanation about how the Advisor arrived at its conclusion

Heat Removal from Bipolar Transistor by Loop Heat Pipe with Nickel and Copper Porous Structures

Directory of Open Access Journals (Sweden)

Patrik Nemec

2014-01-01

Full Text Available Loop heat pipes (LHPs are used in many branches of industry, mainly for cooling of electrical elements and systems. The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work was to develop porous structures from copper and nickel powder with different grain sizes. For experiment copper powder with grain size of 50 and 100 μm and nickel powder with grain size of 10 and 25 μm were used. Analysis of these porous structures and LHP design are described in the paper. And the measurements’ influences of porous structures in LHP on heat removal from the insulated gate bipolar transistor (IGBT have been made.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

International Nuclear Information System (INIS)

Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

2014-01-01

A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

Heat transport and afterheat removal for gas cooled reactors under accident conditions

International Nuclear Information System (INIS)

2001-01-01

The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

Experimental research on passive residual heat remove system for advanced PWR

International Nuclear Information System (INIS)

Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

2003-01-01

The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

Tritium Removal by Laser Heating and Its Application to Tokamaks

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M.; Nishi, M.; Shu, W.

2001-01-01

A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm 2 , and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

2015-09-15

Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

Experimental research on influencing factors of wet removal of NO from coal-fired flue gas by UV/H2O2 advanced oxidation process

Institute of Scientific and Technical Information of China (English)

无

2010-01-01

Wet removal of NO from coal-fired flue gas by UV/H2O2 Advanced Oxidation Process (AOP) were investigated in a self-designed UV-bubble reactor. Several main influencing factors (UV intensity, H2O2 initial concentration, initial pH value, solution temperature, NO initial concentration, liquid-gas ratio and O2 percentage content) on the NO removal efficiency were studied. The results showed that UV intensity, H2O2 initial concentration, NO initial concentration and liquid-gas ratio are the main influencing factors. In the best conditions, the highest NO removal efficiency by UV/H2O2 advanced oxidation process could reach 82.9%. Based on the experimental study, the influencing mechanism of the relevant influencing factors were discussed in depth.

Concept of CFD model of natural draft wet-cooling tower flow

Directory of Open Access Journals (Sweden)

Hyhlík T.

2014-03-01

Full Text Available The article deals with the development of CFD model of natural draft wet-cooling tower flow. The physical phenomena taking place within a natural draft wet cooling tower are described by the system of conservation law equations along with additional equations. The heat and mass transfer in the counterflow wet-cooling tower fill are described by model [1] which is based on the system of ordinary differential equations. Utilization of model [1] of the fill allows us to apply commonly measured fill characteristics as shown by [2].The boundary value problem resulting from the fill model is solved separately. The system of conservation law equations is interlinked with the system of ordinary differential equations describing the phenomena occurring in the counterflow wet-cooling tower fill via heat and mass sources and via boundary conditions. The concept of numerical solution is presented for the quasi one dimensional model of natural draft wet-cooling tower flow. The simulation results are shown.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Comparative Batch and Column Evaluation of Thermal and Wet ...

African Journals Online (AJOL)

The efficiency of regenerated spent commercial activated carbon for synthetic dye removal was studied using thermal and wet oxidative regeneration methods. Two types of experiments were carried out, batch adsorption experiments and continous flow (fixed bed) column experiment to study the mechanism of dye removal ...

Mixing systems for wet and dry plumes and cleaning equipment for the heat exchangers of the dry section. Two indispensible components of an effective and safe hybrid cooling tower

International Nuclear Information System (INIS)

Alt, W.

1990-01-01

At first glance, the hybrid cooling tower seems to be an ingenious combination of the well known components of an evaporative cooling tower and a dry cooling tower. The calculation of the air mass flows for both the wet and dry sections required to achieve an invisible plume does not represent an unsolvable problem to the engineer experienced in thermodynamics. The same also applies to the dimensioning of the heat exchangers and cooling fills. The hybrid cooling tower requires a well designed mixing system in order to ideally mix, the dry plume into the wet plume. If the cooling tower proves its efficiency during commissioning it is important that the ratio of the performance of the wet section to that of the dry section be maintained also in the long term. The performance of the fill in a wet cooling tower is consistently stable. Dirt deposits can form very quickly on the inner and outer surfaces of the heat exchangers of the dry section. In this case the thermal resistance increases rapidly. The respective performance of the wet and dry sections is then no longer balanced and the invisibility of the plume is no longer assured. This can be avoided by providing appropriate cleaning equipment

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Study of passive residual heat removal system of a modular small PWR reactor

International Nuclear Information System (INIS)

Araujo, Nathália N.; Su, Jian

2017-01-01

This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

Design Aspects of Wet Scrubber System

International Nuclear Information System (INIS)

Lee, Hyun-Chul; Bang, Young-suk; Jung, Woo-Young; Lee, Doo-Yong

2015-01-01

The water pool in the wet scrubber system has advantage to cope with decay heat based on the thermal hydraulic balance such as condensation and evaporation inside it. This study focuses on the design aspects of the wet scrubber system to estimate the required water pool mass during the mission time and size of the scrubbing tank including inner structures. The design of the wet scrubber system include the estimation of the required water mass during the mission time and sizing of the scrubber vessel to contain the water pool. The condensation due to the inlet steam and evaporation due to the steam and non-condensable gas superheat and decay heat from filtered fission products should be considered to estimate the water mass required to maintain its function during the mission time. On the other hand, the level swelling due to the noncondensable gas is another important design aspect on the sizing of the scrubber vessel and determination of the entry elevation of the filtration components such as the droplet separator or filter. The minimum water level based on the minimum collapsed water level should be higher than the exit of scrubber nozzle

Design Aspects of Wet Scrubber System

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyun-Chul; Bang, Young-suk; Jung, Woo-Young; Lee, Doo-Yong [FNC Technology Co. Ltd., Yongin (Korea, Republic of)

2015-10-15

The water pool in the wet scrubber system has advantage to cope with decay heat based on the thermal hydraulic balance such as condensation and evaporation inside it. This study focuses on the design aspects of the wet scrubber system to estimate the required water pool mass during the mission time and size of the scrubbing tank including inner structures. The design of the wet scrubber system include the estimation of the required water mass during the mission time and sizing of the scrubber vessel to contain the water pool. The condensation due to the inlet steam and evaporation due to the steam and non-condensable gas superheat and decay heat from filtered fission products should be considered to estimate the water mass required to maintain its function during the mission time. On the other hand, the level swelling due to the noncondensable gas is another important design aspect on the sizing of the scrubber vessel and determination of the entry elevation of the filtration components such as the droplet separator or filter. The minimum water level based on the minimum collapsed water level should be higher than the exit of scrubber nozzle.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Cyclic process for producing methane from carbon monoxide with heat removal

Science.gov (United States)

Frost, Albert C.; Yang, Chang-lee

1982-01-01

Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

An analysis of heat removal during cryogen spray cooling and effects of simultaneous airflow application.

Science.gov (United States)

Torres, J H; Tunnell, J W; Pikkula, B M; Anvari, B

2001-01-01

Cryogen spray cooling (CSC) is a method used to protect the epidermis from non-specific thermal injury that may occur as a result of various dermatological laser procedures. However, better understanding of cryogen deposition and skin thermal response to CSC is needed to optimize the technique. Temperature measurements and video imaging were carried out on an epoxy phantom as well as human skin during CSC with and without simultaneous application of airflow which was intended to accelerate cryogen evaporation from the substrate surface. An inverse thermal conduction model was used to estimate heat flux and total heat removed. Lifetime of the cryogen film deposited on the surface of skin and epoxy phantom lasted several hundred milliseconds beyond the spurt, but could be reduced to the spurt duration by application of airflow. Values over 100 J/cm(3) were estimated for volumetric heat removed from the epidermis using CSC. "Film cooling" instead of "evaporative cooling" appears to be the dominant mode of CSC on skin. Estimated values of heat removed from the epidermis suggest that a cryogen spurt as long as 200 milliseconds is required to counteract heat generated by high laser fluences (e.g., in treatment of port wine stains) in patients with high concentration of epidermal melanin. Additional cooling beyond spurt termination can be avoided by simultaneous application of airflow, although it is unclear at the moment if avoiding the additional cooling would be beneficial in the actual clinical situation. Copyright 2001 Wiley-Liss, Inc.

Catalytic wet air oxidation of bisphenol A solution in a batch-recycle trickle-bed reactor over titanate nanotube-based catalysts.

Science.gov (United States)

Kaplan, Renata; Erjavec, Boštjan; Senila, Marin; Pintar, Albin

2014-10-01

Catalytic wet air oxidation (CWAO) is classified as an advanced oxidation process, which proved to be highly efficient for the removal of emerging organic pollutant bisphenol A (BPA) from water. In this study, BPA was successfully removed in a batch-recycle trickle-bed reactor over bare titanate nanotube-based catalysts at very short space time of 0.6 min gCAT g(-1). The as-prepared titanate nanotubes, which underwent heat treatment at 600 °C, showed high activity for the removal of aqueous BPA. Liquid-phase recycling (5- or 10-fold recycle) enabled complete BPA conversion already at 200 °C, together with high conversion of total organic carbon (TOC), i.e., 73 and 98 %, respectively. The catalyst was chemically stable in the given range of operating conditions for 189 h on stream.

Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

International Nuclear Information System (INIS)

Lofaro, R.J.; Aggarwal, S.

1992-01-01

The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

Experimental and numerical studies on the treatment of wet astronaut trash by forced-convection drying

Science.gov (United States)

Arquiza, J. M. R. Apollo; Morrow, Robert; Remiker, Ross; Hunter, Jean B.

2017-09-01

During long-term space missions, astronauts generate wet trash, including food containers with uneaten portions, moist hygiene wipes and wet paper towels. This waste produces two problems: the loss of water and the generation of odors and health hazards by microbial growth. These problems are solved by a closed-loop, forced-convection, heat-pump drying system which stops microbial activity by both pasteurization and desiccation, and recovers water in a gravity-independent porous media condensing heat exchanger. A transient, pseudo-homogeneous continuum model for the drying of wet ersatz trash was formulated for this system. The model is based on the conservation equations for energy and moisture applied to the air and solid phases and includes the unique trash characteristic of having both dry and wet solids. Experimentally determined heat and mass transfer coefficients, together with the moisture sorption equilibrium relationship for the wet material are used in the model. The resulting system of differential equations is solved by the finite-volume method as implemented by the commercial software COMSOL. Model simulations agreed well with experimental data under certain conditions. The validated model will be used in the optimization of the entire closed-loop system consisting of fan, air heater, dryer vessel, heat-pump condenser, and heat-recovery modules.

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

International Nuclear Information System (INIS)

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

1993-04-01

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

Application of optimal estimation techniques to FFTF decay heat removal analysis

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

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

Energy and exergy analysis of counter flow wet cooling towers

Directory of Open Access Journals (Sweden)

Saravanan Mani

2008-01-01

Full Text Available Cooling tower is an open system direct contact heat exchanger, where it cools water by both convection and evaporation. In this paper, a mathematical model based on heat and mass transfer principle is developed to find the outlet condition of water and air. The model is solved using iterative method. Energy and exergy analysis infers that inlet air wet bulb temperature is found to be the most important parameter than inlet water temperature and also variation in dead state properties does not affect the performance of wet cooling tower. .

Studies related to emergency decay heat removal in EBR-II

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Removal of organic pollutants in tannery wastewater from wet-blue fur processing by integrated Anoxic/Oxic (A/O) and Fenton: Process optimization

DEFF Research Database (Denmark)

Wang, Yong; Li, Weiguang; Angelidaki, Irini

2014-01-01

Treatment of tannery wastewater has been a challenge in remediation of aquatic environment in developing countries. Removal of organic pollutants in tannery wastewater from wet-blue fur processing was studied using integrated processes of Anoxic/Oxic and Fenton. Analysis of COD composition based...... 80%. In the subsequent Fenton oxidation, effects of initial pH and H2O2 dose on COD removal were investigated, and response surface methodology was adopted to obtain the optimal conditions as initial pH of 4.0, H2O2 dose of 14.0mM, H2O2:Fe2+ molar ratio of 10.6, and reaction time of 3h to achieve...... the highest COD removal of 55.87%. GC-MS analysis was carried out to observe the change of organic composition during Fenton oxidation, and most of the residual organic pollutants resistant to Fenton treatment belonged to organosilanes and saturated alkanes. This study will provide useful information...

Feasibility study for a postaccident heat removal facility

International Nuclear Information System (INIS)

Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

1978-01-01

An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

Heat removal tests on dry storage facilities for nuclear spent fuels

International Nuclear Information System (INIS)

Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

1999-01-01

In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

Processes for exhaust purification of biomass combustion systems, dust removal, heat recovery, technologies and practical experience; Verfahren zur Abgasreinigung nach Biomasseverbrennung, Abgasentstaubung, Abgasreinigung Moeglichkeiten zur Waermerueckgewinnung; Technologien und Praxiserfahrungen

Energy Technology Data Exchange (ETDEWEB)

Jirkowsky, C.; Pretzl, R.; Sihorsch, K.

2003-07-01

The authors report on air pollution control systems of biomass burners: dedusting, centrifugal filtration, cyclone separators, electric filters (wet and dry), fabric filters, wet scrubbers. Technical specifications and methods of heat recovery are given. (uke)

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Residual heat removal during accidental situations

International Nuclear Information System (INIS)

Depond, M.; Sureau, H.; Tellier, N.

1983-07-01

Existing emergency procedures, whose purpose is residual heat removal and a safe recovery are based on sequential analysis and initiating event diagnosis. This approach was found in some cases inappropriate and inefficient, specially in case of out-of-design accidents corresponding to multiple equipment failure or simultaneous human failures. To cope with these situations, a new approach was necessary. Parallel studies performed in France at Framatome (the designer) and Electricity de France (the utility) gave a new method, called NSSS physical states approach. Prior to the implementation of this method which necessitates further studies and developments, some improvements in the existing operating procedures derived from the NSSS physical states have already been achieved: that is the case for the safety injection control and the development of an emergency procedure called ''U1''. This paper will briefly physical states approach and present the ''U1'' procedure. The tools which will be used to chack these methods are also mentioned

A decay heat removal system requiring no external energy

International Nuclear Information System (INIS)

Costes, D.; Fermandjian, J.

1983-12-01

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

Self-drying: a gecko's innate ability to remove water from wet toe pads.

Directory of Open Access Journals (Sweden)

Alyssa Y Stark

Full Text Available When the adhesive toe pads of geckos become wet, they become ineffective in enabling geckos to stick to substrates. This result is puzzling given that many species of gecko are endemic to tropical environments where water covered surfaces are ubiquitous. We hypothesized that geckos can recover adhesive capabilities following exposure of their toe pads to water by walking on a dry surface, similar to the active self-cleaning of dirt particles. We measured the time it took to recover maximum shear adhesion after toe pads had become wet in two groups, those that were allowed to actively walk and those that were not. Keeping in mind the importance of substrate wettability to adhesion on wet surfaces, we also tested geckos on hydrophilic glass and an intermediately wetting substrate (polymethylmethacrylate; PMMA. We found that time to maximum shear adhesion recovery did not differ in the walking groups based on substrate wettability (22.7±5.1 min on glass and 15.4±0.3 min on PMMA but did have a significant effect in the non-walking groups (54.3±3.9 min on glass and 27.8±2.5 min on PMMA. Overall, we found that by actively walking, geckos were able to self-dry their wet toe pads and regain maximum shear adhesion significantly faster than those that did not walk. Our results highlight a unexpected property of the gecko adhesive system, the ability to actively self-dry and recover adhesive performance after being rendered dysfunctional by water.

Large wet-type cooling towers and their influence on the environment

International Nuclear Information System (INIS)

Schiffers, A.

1977-01-01

Large wet-type cooling towers with natural draft are said to be ecologically beneficial today, especially concerning the heat emission from power plants. A description is given of the influence of such cooling towers on the environment and the possible climatic influences are considered in detail. Recent investigations have shown that wet-type cooling towers represent no danger of any kind for fauna and flora as to the bacterial radiation. Physical studies have shown that neither the emitted water vapour nor the heat emitted into the atmosphere, can significantly change the macroclimate and microclimate. At present, wet-type cooling towers cannot be replaced by dry-type or so-called hybrid-type cooling towers, the technical development of which for large units being not yet guaranteed. (orig.) [de

Description and cost analysis of a deluge dry/wet cooling system.

Energy Technology Data Exchange (ETDEWEB)

Wiles, L.E.; Bamberger, J.A.; Braun, D.J.; Braun, D.J.; Faletti, D.W.; Willingham, C.E.

1978-06-01

The use of combined dry/wet cooling systems for large base-load power plants offers the potential for significant water savings as compared to evaporatively cooled power plants and significant cost savings in comparison to dry cooled power plants. The results of a detailed engineering and cost study of one type of dry/wet cooling system are described. In the ''deluge'' dry/wet cooling method, a finned-tube heat exchanger is designed to operate in the dry mode up to a given ambient temperature. To avoid the degradation of performance for higher ambient temperatures, water (the delugeate) is distributed over a portion of the heat exchanger surface to enhance the cooling process by evaporation. The deluge system used in this study is termed the HOETERV system. The HOETERV deluge system uses a horizontal-tube, vertical-plate-finned heat exchanger. The delugeate is distributed at the top of the heat exchanger and is allowed to fall by gravity in a thin film on the face of the plate fin. Ammonia is used as the indirect heat transfer medium between the turbine exhaust steam and the ambient air. Steam is condensed by boiling ammonia in a condenser/reboiler. The ammonia is condensed in the heat exchanger by inducing airflow over the plate fins. Various design parameters of the cooling system have been studied to evaluate their impact on the optimum cooling system design and the power-plant/utility-system interface. Annual water availability was the most significant design parameter. Others included site meteorology, heat exchanger configuration and air flow, number and size of towers, fan system design, and turbine operation. It was concluded from this study that the HOETERV deluge system of dry/wet cooling, using ammonia as an intermediate heat transfer medium, offers the potential for significant cost savings compared with all-dry cooling, while achieving substantially reduced water consumption as compared to an evaporatively cooled power plant. (LCL)

The transfer function model for dynamic response of wet cooling coils

International Nuclear Information System (INIS)

Yao Ye; Liu Shiqing

2008-01-01

This paper mainly concerned about the dynamic response model of wet cooling coils that is developed by the Laplace transform method. The theoretic equations are firstly established based on the theory of energy conservation. Then, the transfer functions on the transient responses of wet cooling coils have been deduced using the method of Laplace transform. The transfer functions reveal the dynamic relationships between the inlet variables and the outlet ones of the cooling coils. Partial-fraction method and Newton-Raphson method are both used in the inversion of the transfer functions from the s-domain to τ-domain. To make the dynamic model of wet cooling coils more adaptive, RBFNN method is employed to determine the coefficients of heat and mass transfer. Experiments have been done and manifested that the coefficients of heat and mass transfer by RBFNN will be of great value to the validity of the transient response model of wet cooling coils in this study

THERMAL TRANSFERS IN WET HYPERBARIC ENVIRONMENT

Directory of Open Access Journals (Sweden)

Tamara STANCIU

2014-05-01

Full Text Available The heat losses of human body are greater in underwater environment than in dry, normal atmosphere, due to the great heat capacity of water. Body temperature of divers in immersion was studied taking into account the pressure the divers are subjected to. The theoretic equation that describes the total heat transfer- at both levels: skin and respiratory system- was established, considering conduction, convection and respiratory gas heating and humidification. The body temperature of the divers was measured in a series of dives at different depths of immersion, conducted in the wet simulator of the Diving Center, in Constanta. The experimental results were in good accordance with the temperature predicted by the mathematical model.

Experimental study on heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

Energy Technology Data Exchange (ETDEWEB)

Islam, M.S.; Monde, Masanori [Saga Univ. (Japan); Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio

1997-07-01

Frictional pressure drop and heat transfer performance in a very narrow rectangular channel having one-sided constant heat flux and repeated-ribs for turbulent flow have been investigated experimentally, and their experimental correlations were obtained using the least square method. The rib pitch-to-height ratios(p/k) were 10 and 20 while holding the rib height constant at 0.2mm, the Reynolds number(Re) from 2,414 to 98,458 under different channel heights of 1.2mm, 2.97mm, and 3.24mm, the rib height-to-channel equivalent diameter(k/De) of 0.03, 0.04, and 0.09 respectively. The results show that the rib-roughened surface augments heat transfer 2-3 times higher than that of the smooth surface with the expense of 2.8-4 times higher frictional pressure drop under Re=5000-10{sup 5}, p/k=10, and H=1.2mm. Experimental results obtained by channel height, H=1.2mm shows a little bit higher heat transfer and friction factor performance than the higher channel height, H=3.24mm. The effect of fin and consequently higher turbulence intensity are responsible for producing higher heat transfer rates. The obtained correlations could be used to design the cooling passages between the target plates to remove high heat flux up to 12MW/m{sup 2} generated at target plates in a high-intensity proton accelerator system. (author). 54 refs.

Experimental study on heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

International Nuclear Information System (INIS)

Islam, M.S.; Monde, Masanori; Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio.

1997-07-01

Frictional pressure drop and heat transfer performance in a very narrow rectangular channel having one-sided constant heat flux and repeated-ribs for turbulent flow have been investigated experimentally, and their experimental correlations were obtained using the least square method. The rib pitch-to-height ratios(p/k) were 10 and 20 while holding the rib height constant at 0.2mm, the Reynolds number(Re) from 2,414 to 98,458 under different channel heights of 1.2mm, 2.97mm, and 3.24mm, the rib height-to-channel equivalent diameter(k/De) of 0.03, 0.04, and 0.09 respectively. The results show that the rib-roughened surface augments heat transfer 2-3 times higher than that of the smooth surface with the expense of 2.8-4 times higher frictional pressure drop under Re=5000-10 5 , p/k=10, and H=1.2mm. Experimental results obtained by channel height, H=1.2mm shows a little bit higher heat transfer and friction factor performance than the higher channel height, H=3.24mm. The effect of fin and consequently higher turbulence intensity are responsible for producing higher heat transfer rates. The obtained correlations could be used to design the cooling passages between the target plates to remove high heat flux up to 12MW/m 2 generated at target plates in a high-intensity proton accelerator system. (author). 54 refs

Reliability assessment on decay heat removal system of a fast reactor

International Nuclear Information System (INIS)

Hioki, Kazumasa

1991-01-01

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

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

Heat removal capability of divertor coaxial tube assembly

International Nuclear Information System (INIS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications. (author)

Enhanced Control of Mercury and other HAPs by Innovative Modifications to Wet FGD Processes

International Nuclear Information System (INIS)

Hargrove, O.W.; Carey, T.R.; Richardson, C.F.; Skarupa, R.C.; Meserole, F.B.; Rhudy, R.G.; Brown, Thomas D.

1997-01-01

The overall objective of this project was to learn more about controlling emissions of hazardous air pollutants (HAPs) from coal-fired power plants that are equipped with wet flue gas desulfurization (FGD) systems. The project was included by FETC as a Phase I project in its Mega-PRDA program. Phase I of this project focused on three research areas. These areas in order of priority were: (1) Catalytic oxidation of vapor-phase elemental mercury; (2) Enhanced particulate-phase HAPs removal by electrostatic charging of liquid droplets; and (3) Enhanced mercury removal by addition of additives to FGD process liquor. Mercury can exist in two forms in utility flue gas--as elemental mercury and as oxidized mercury (predominant form believed to be HgCl 2 ). Previous test results have shown that wet scrubbers effectively remove the oxidized mercury from the gas but are ineffective in removing elemental mercury. Recent improvements in mercury speciation techniques confirm this finding. Catalytic oxidation of vapor-phase elemental mercury is of interest in cases where a wet scrubber exists or is planned for SO 2 control. If a loW--cost process could be developed to oxidize all of the elemental mercury in the flue gas, then the maximum achievable mercury removal across the existing or planned wet scrubber would increase. Other approaches for improving control of HAPs included a method for improving particulate removal across the FGD process and the use of additives to increase mercury solubility. This paper discusses results related only to catalytic oxidation of elemental mercury

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Quasi One-Dimensional Model of Natural Draft Wet-Cooling Tower Flow, Heat and Mass Transfer

Directory of Open Access Journals (Sweden)

Hyhlík Tomáš

2015-01-01

Full Text Available The article deals with the development of CFD (Computational Fluid Dynamics model of natural draft wet-cooling tower flow, heat and mass transfer. The moist air flow is described by the system of conservation laws along with additional equations. Moist air is assumed to be homogeneous mixture of dry air and water vapour. Liquid phase in the fill zone is described by the system of ordinary differential equations. Boundary value problem for the system of conservation laws is discretized in space using Kurganov-Tadmor central scheme and in time using strong stability preserving Runge-Kutta scheme. Initial value problems in the fill zone is solved by using standard fourth order Runge-Kutta scheme. The interaction between liquid water and moist air is done by source terms in governing equations.

Heat removal tests for pressurized water reactor containment spray by largescale facility

International Nuclear Information System (INIS)

Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

1983-01-01

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

Possible design of PBR for passive decay heat removal

International Nuclear Information System (INIS)

Sambuu, Odmaa; Obara, Toru

2016-01-01

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

Hot water extraction with in situ wet oxidation: Kinetics of PAHs removal from soil

International Nuclear Information System (INIS)

Dadkhah, Ali A.; Akgerman, Aydin

2006-01-01

Finding environmentally friendly and cost-effective methods to remediate soils contaminated with polycyclic aromatic hydrocarbons (PAHs) is currently a major concern of researchers. In this study, a series of small-scale semi-continuous extractions - with and without in situ wet oxidation - were performed on soils polluted with PAHs, using subcritical water (i.e. liquid water at high temperatures and pressures, but below the critical point) as the removal agent. Experiments were performed in a 300 mL reactor using an aged soil sample. To find the desorption isotherms and oxidation reaction rates, semi-continuous experiments with residence times of 1 and 2 h were performed using aged soil at 250 deg. C and hydrogen peroxide as oxidizing agent. In all combined extraction and oxidation flow experiments, PAHs in the remaining soil after the experiments were almost undetectable. In combined extraction and oxidation no PAHs could be detected in the liquid phase after the first 30 min of the experiments. Based on these results, extraction with hot water, if combined with oxidation, should reduce the cost of remediation and can be used as a feasible alternative technique for remediating contaminated soils and sediments

Modular Micromachined Si Heat Removal (MOMS Heat Removal): Electronic Integration and System Test

National Research Council Canada - National Science Library

Brown, Elliott

2003-01-01

...: (1) insulated-gated bipolar transistors (IGBTs), and (2) laterally-diffused (LD) MOSFETs. Heat pipes were found to provide little or no advantage over conventional copper-based heat spreaders in both device applications...

Heat release, time required, and cleaning ability of MTwo R and ProTaper universal retreatment systems in the removal of filling material.

Science.gov (United States)

Bramante, Clovis Monteiro; Fidelis, Natasha Siqueira; Assumpção, Tatiana Santos; Bernardineli, Norberti; Garcia, Roberto Brandão; Bramante, Alexandre Silva; de Moraes, Ivaldo Gomes

2010-11-01

This ex vivo study evaluated the heat release, time required, and cleaning efficacy of MTwo (VDW, Munich, Germany) and ProTaper Universal Retreatment systems (Dentsply/Maillefer, Ballaigues, Switzerland) and hand instrumentation in the removal of filling material. Sixty single-rooted human teeth with a single straight canal were obturated with gutta-percha and zinc oxide and eugenol-based cement and randomly allocated to 3 groups (n = 20). After 30-day storage at 37 °C and 100% humidity, the root fillings were removed using ProTaper UR, MTwo R, or hand files. Heat release, time required, and cleaning efficacy data were analyzed statistically (analysis of variance and the Tukey test, α = 0.05). None of the techniques removed the root fillings completely. Filling material removal with ProTaper UR was faster but caused more heat release. Mtwo R produced less heat release than the other techniques but was the least efficient in removing gutta-percha/sealer. ProTaper UR and MTwo R caused the greatest and lowest temperature increase on root surface, respectively; regardless of the type of instrument, more heat was released in the cervical third. Pro Taper UR needed less time to remove fillings than MTwo R. All techniques left filling debris in the root canals. Copyright © 2010 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

Failure Modes and Effects Analysis (FMEA) of the Residual Heat Removal System

International Nuclear Information System (INIS)

Eggleston, F.T.

1976-01-01

The Residual Heat Removal System (RHRS) transfer heat from the Reactor Coolant System (RCS) to the reactor plant Component Cooling System (CCS) to reduce the temperature of the RCS at a controlled rate during the second part of normal plant cooldown and maintains the desired temperature until the plant is restarted. By the use of an analytic tool, the Failure Modes and Effects Analysis, it is shown that the RHRS, because of its redundant two train design, is able to accommodate any credible component single failure with the only effect being an extension in the required cooldown time, thus demonstrating the reliability of the RHRS to perform its intended function

Effect of SUS316L stainless steel surface conditions on the wetting of molten multi-component oxides ceramic

Energy Technology Data Exchange (ETDEWEB)

Wang, Jin, E-mail: wangjinustb@gmail.com [Graduate School of Life Science and Systems Engineering, Kyushu Institute of Technology, Fukuoka, 808-0196 (Japan); Matsuda, Nozomu [Bar and Wire Product Unit, Nippon steel and Sumitomo Metal Corporation, Fukuoka, 802-8686 (Japan); Shinozaki, Nobuya [Graduate School of Life Science and Systems Engineering, Kyushu Institute of Technology, Fukuoka, 808-0196 (Japan); Miyoshi, Noriko [The Center for Instrumental Analysis, Kyushu Institute of Technology, Fukuoka, 804-8550 (Japan); Shiraishi, Takanobu [Graduate School of Biomedical Sciences, Nagasaki University, Nagasaki, 852-8588 (Japan)

2015-02-01

Highlights: • Multi-component oxides had a good wetting on stainless substrates with pretreatments. • Various substrates surface roughness caused the difference of final contact angles. • The wetting rate was slow on polished substrate due to the slow surface oxidation. - Abstract: A study on the effect of SUS316L stainless steel surface conditions on the wetting behavior of molten multi-component oxides ceramic was performed and aimed to contribute to the further understanding of the application of oxides ceramic in penetration treatment of stainless steel coatings and the deposition of stainless steel cermet coatings. The results show that at 1273 K, different surface pre-treatments (polishing and heating) had an important effect on the wetting behavior. The molten multi-component oxides showed good wettability on both stainless steel substrates, however, the wetting process on the polished substrate was significantly slower than that on the heated substrates. The mechanism of the interfacial reactions was discussed based on the microscopic and thermodynamic analysis, the substrates reacted with oxygen generated from the decomposition of the molten multi-component oxides and oxygen contained in the argon atmosphere, and the oxide film caused the molten multi-component oxides ceramic to spread on the substrates surfaces. For the polished substrate, more time was required for the surface oxidation to reach the surface composition of Heated-S, which resulted in relatively slow spreading and wetting rates. Moreover, the variance of the surface roughness drove the final contact angles to slightly different values following the sequence Polished-S > Heated-S.

System Study: Residual Heat Removal 1998-2014

Energy Technology Data Exchange (ETDEWEB)

Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-12-01

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trends were identified in the RHR results. A highly statistically significant decreasing trend was observed for the RHR injection mode start-only unreliability. Statistically significant decreasing trends were observed for RHR shutdown cooling mode start-only unreliability and RHR shutdown cooling model 24-hour unreliability.

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

International Nuclear Information System (INIS)

Loimann, E.; Reutler, H.

1977-01-01

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

High-rate oil removing scouring agent. Koyubun jokyoyo seirenzai ni tsuite

Energy Technology Data Exchange (ETDEWEB)

Ishihara, K.; Sato, Y. (Dai-Ichi Kogyo Seiyaku Co. Ltd., Kyoto (Japan))

1991-11-01

Fiber forming, scutching and knitting processes in recent years are performed three to five times faster than in the conventional processes. Associated therewith, oil solutions are taken importantly for their stability and workability, such as smoothing properties, heat resistance and abrasion resistance. On the other hand, difficulty is increasing in removing the oils after scutching and knitting. This paper explains basic rinsing activities required in oil removal, and describes various test characteristics and compatibility of various high-rate oil removing scouring agents. An oil-in-water rinsing mechanism relies upon comprehensive actions of a surfactant in wetting, permeation, emulsified dispersion and solubilization. The most importantly taken among them is the emulsifying action, which largely depends upon its chemical structure. Therefore, for a high-rate oil removing scouring agent, creation of activators is required that make the above basic characteristics and activities compatible for various applications. For example, the above product covers a great variety of kinds for diverse applications, based on non-ionic and anion-based activators. 6 figs., 20 tabs.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Photocatalytic oxidation removal of Hg"0 using ternary Ag/AgI-Ag_2CO_3 hybrids in wet scrubbing process under fluorescent light

International Nuclear Information System (INIS)

Zhang, Anchao; Zhang, Lixiang; Chen, Xiaozhuan; Zhu, Qifeng; Liu, Zhichao; Xiang, Jun

2017-01-01

Highlights: • Ag/AgI-Ag_2CO_3 hybrids were employed for Hg"0 removal under fluorescent light. • Superoxide radical (·O_2"−) played a key role in Hg"0 removal. • NO exhibited a significant effect on Hg"0 removal in comparison to SO_2. • The mechanism for enhanced Hg"0 removal over Ag/AgI-Ag_2CO_3 was proposed. - Abstract: A series of ternary Ag/AgI-Ag_2CO_3 photocatalysts synthesized using a facile coprecipitation method were employed to investigate their performances of Hg"0 removal in a wet scrubbing reactor. The hybrids were characterized by N_2 adsorption-desorption, XRD, SEM-EDS, HRTEM, XPS, DRS and ESR. The photocatalytic activities of Hg"0 removal were evaluated under fluorescent light. The results showed that AgI content, fluorescent light irradiation, reaction temperature all showed significant influences on Hg"0 removal. NO exhibited significant effect on Hg"0 removal in comparison to SO_2. Among these ternary Ag/AgI-Ag_2CO_3 hybrids, Ag/AgI(0.1)-Ag_2CO_3 showed the highest Hg"0 removal efficiency, which could be ascribed to the effective separation of photogenerated electron-hole pairs between AgI and Ag_2CO_3 and the surface plasmon resonance (SPR) effect in the visible region by metallic silver nanoparticles (Ag"0 NPs). The trapping studies of reactive radicals showed that the superoxide radicals (·O_2"−) may play a key role in Hg"0 removal under fluorescent light. According to the experimental and characterization results, a possible photocatalytic oxidation mechanism for enhanced Hg"0 removal over Ag/AgI(0.1)-Ag_2CO_3 hybrid under fluorescent light was proposed.

Wet water glass production plant

Directory of Open Access Journals (Sweden)

Stanković Mirjana S.

2003-01-01

Full Text Available The IGPC Engineering Department designed basic projects for a wet hydrate dissolution plant, using technology developed in the IGPC laboratories. Several projects were completed: technological, machine, electrical, automation. On the basis of these projects, a production plant of a capacity of 75,000 t/y was manufactured, at "Zeolite Mira", Mira (VE, Italy, in 1997. and 1998, increasing detergent zeolite production, from 50,000 to 100,000 t/y. Several goals were realized by designing a wet hydrate dissolution plant. The main goal was increasing the detergent zeolite production. The technological cycle of NaOH was closed, and no effluents emitted, and there is no pollution (except for the filter cake. The wet water glass production process is fully automatized, and the product has uniform quality. The production process can be controlled manually, which is necessary during start - up, and repairs. By installing additional process equipment (centrifugal pumps and heat exchangers technological bottlenecks were overcome, and by adjusting the operation of autoclaves, and water glass filters and also by optimizing the capacities of process equipment.

Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

International Nuclear Information System (INIS)

Venker, Jeanne

2015-01-01

The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO 2 ) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO 2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO 2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO 2 Brayton cycles and to evaluate the introduced heat removal system

Dry purification of aspirational air in coke-sorting systems with wet slaking of coke

Energy Technology Data Exchange (ETDEWEB)

T.F. Trembach; A.G. Klimenko [Giprokoks, the State Institute for the Design of Coke-Industry Enterprises, Kharkov (Ukraine)

2009-07-15

Coke transportation after wet slaking is accompanied by the release of dust in the production building and in the surrounding atmosphere. Wet methods are traditionally used to purify very humid air. Giprokoks has developed designs for highly efficient dry dust-removal methods in such conditions.

Analysis of hybrid electric/thermofluidic inputs for wet shape memory alloy actuators

Science.gov (United States)

Flemming, Leslie; Mascaro, Stephen

2013-01-01

A wet shape memory alloy (SMA) actuator is characterized by an SMA wire embedded within a compliant fluid-filled tube. Heating and cooling of the SMA wire produces a linear contraction and extension of the wire. Thermal energy can be transferred to and from the wire using combinations of resistive heating and free/forced convection. This paper analyzes the speed and efficiency of a simulated wet SMA actuator using a variety of control strategies involving different combinations of electrical and thermofluidic inputs. A computational fluid dynamics (CFD) model is used in conjunction with a temperature-strain model of the SMA wire to simulate the thermal response of the wire and compute strains, contraction/extension times and efficiency. The simulations produce cycle rates of up to 5 Hz for electrical heating and fluidic cooling, and up to 2 Hz for fluidic heating and cooling. The simulated results demonstrate efficiencies up to 0.5% for electric heating and up to 0.2% for fluidic heating. Using both electric and fluidic inputs concurrently improves the speed and efficiency of the actuator and allows for the actuator to remain contracted without continually delivering energy to the actuator, because of the thermal capacitance of the hot fluid. The characterized speeds and efficiencies are key requirements for implementing broader research efforts involving the intelligent control of electric and thermofluidic networks to optimize the speed and efficiency of wet actuator arrays.

Wet-air oxidation cleans up black wastewater

Energy Technology Data Exchange (ETDEWEB)

1993-09-01

Sterling Organics produces the analgesic paracetamol (acetaminophen) at its Dudley, England, plant. The wastewater from the batch process contains intermediates such as para-aminophenol (PAP) and byproducts such as thiosulfates, sulfites and sulfides. To stay ahead of increasingly strict environmental legislation, Sterling Organics installed a wet-air oxidation system at the Dudley facility in August 1992. The system is made by Zimpro Environmental Inc. (Rothschild, Wis.). Zimpro's wet-air oxidation system finds a way around the limitations of purely chemical or physical processes. In the process, compressed air at elevated temperature and pressure oxidizes the process intermediates and byproducts and removes the color from the wastewater.

Cyclic process for producing methane in a tubular reactor with effective heat removal

Science.gov (United States)

Frost, Albert C.; Yang, Chang-Lee

1986-01-01

Carbon monoxide-containing gas streams are converted to methane by a cyclic, essentially two-step process in which said carbon monoxide is disproportionated to form carbon dioxide and active surface carbon deposited on the surface of a catalyst, and said carbon is reacted with steam to form product methane and by-product carbon dioxide. The exothermic heat of reaction generated in each step is effectively removed during each complete cycle so as to avoid a build up of heat from cycle-to-cycle, with particularly advantageous techniques being employed for fixed bed, tubular and fluidized bed reactor operations.

Experimental and analytical studies on the passive residual heat removal system for the advanced integral type reactor

International Nuclear Information System (INIS)

Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki

2004-01-01

An experiment on the thermal-hydraulic characteristics of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART-P, has been performed, and its experimental results have been analyzed using a best-estimated system analysis code, MARS. The experiment is performed to investigate the performance of the passive residual heat removal system using the high temperature and high pressure thermal-hydraulic test facility (VISTA) which simulates the SMART-P. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfer through the PRHRS heat exchanger in the emergency cooldown tank is sufficient enough to enable a natural circulation of the coolant. Analysis on a typical PRHRS test has been carried out using the MARS code. The overall trends of the calculated flow rate, pressure, temperature, and heat transfer rate in the PRHRS are similar to the experimental data. There is good agreement between the experimental data and the calculated one for the fluid temperature in the PRHRS steam line. However, the calculated fluid temperature in the PRHRS condensate line is higher, the calculated coolant outlet temperature is lower, and the heat transfer rate through the PRHRS heat exchanger is lower than the experimental data. It seems that it is due to an insufficient heat transfer modeling in the pool such as the emergency cooldown tank in the MARS calculation. (author)

Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

Energy Technology Data Exchange (ETDEWEB)

Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

2016-02-15

Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Development of evaluation method for heat removal design of dry storage facilities. Pt. 4. Numerical analysis on vault storage system of cross flow type

International Nuclear Information System (INIS)

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

1999-01-01

On the basis of the result of the heat removal test on vault storage system of cross flow type using the 1/5 scale model, an evaluation method for the heat removal design was established. It was composed of the numerical analysis for the convection phenomena of air flow inside the whole facility and that for the natural convection and the detailed turbulent mechanism near the surface of the storage tube. In the former analysis, air temperature distribution in the storage area obtained by the calculation gave good agreement within ±3degC with the test result. And fine turbulence models were introduced in the latter analysis to predict the separation flow in the boundary layer near the surface of the storage tube and the buoyant flow generated by the heat from the storage tube. Furthermore, the properties of removing the heat in a designed full-scale storage facility, such as flow pattern in the storage area, temperature and heat transfer rate of the storage tubes, were evaluated by using each of three methods, which were the established numerical analysis method, the experimental formula demonstrated in the heat removal test and the conventional evaluation method applied to the past heat removal design. As a result, the safety margin and issues included in the methods were grasped, and the measures to make a design more rational were proposed. (author)

Simultaneous removal of NO and SO2 using vacuum ultraviolet light (VUV)/heat/peroxymonosulfate (PMS).

Science.gov (United States)

Liu, Yangxian; Wang, Yan; Wang, Qian; Pan, Jianfeng; Zhang, Jun

2018-01-01

Simultaneous removal process of SO 2 and NO from flue gas using vacuum ultraviolet light (VUV)/heat/peroxymonosulfate (PMS) in a VUV spraying reactor was proposed. The key influencing factors, active species, reaction products and mechanism of SO 2 and NO simultaneous removal were investigated. The results show that vacuum ultraviolet light (185 nm) achieves the highest NO removal efficiency and yield of and under the same test conditions. NO removal is enhanced at higher PMS concentration, light intensity and oxygen concentration, and is inhibited at higher NO concentration, SO 2 concentration and solution pH. Solution temperature has a double impact on NO removal. CO 2 concentration has no obvious effect on NO removal. and produced from VUV-activation of PMS play a leading role in NO removal. O 3 and ·O produced from VUV-activation of O 2 also play an important role in NO removal. SO 2 achieves complete removal under all experimental conditions due to its very high solubility in water and good reactivity. The highest simultaneous removal efficiency of SO 2 and NO reaches 100% and 91.3%, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

Wetting characteristics of 3-dimensional nanostructured fractal surfaces

Energy Technology Data Exchange (ETDEWEB)

Davis, Ethan, E-mail: ethan.davis4@huskers.unl.edu [Nano & Microsystems Research Laboratory, Department of Mechanical and Materials Engineering, University of Nebraska-Lincoln, W342 Nebraska Hall, Lincoln, NE 68588-0526 (United States); Liu, Ying; Jiang, Lijia; Lu, Yongfeng [Laser Assisted Nano Engineering Lab, Department of Electrical and Computer Engineering, University of Nebraska-Lincoln, 209N Scott Engineering Center, Lincoln, NE 68588-0511 (United States); Ndao, Sidy, E-mail: sndao2@unl.edu [Nano & Microsystems Research Laboratory, Department of Mechanical and Materials Engineering, University of Nebraska-Lincoln, W342 Nebraska Hall, Lincoln, NE 68588-0526 (United States)

2017-01-15

Highlights: • Hierarchically structured surfaces were fabricated on the micro/nano-scale. • These structures reduced the contact angle of the inherently hydrophilic material. • Similar surfaces have applications in two-phase heat transfer and microfluidics. - Abstract: This article reports the fabrication and wetting characteristics of 3-dimensional nanostructured fractal surfaces (3DNFS). Three distinct 3DNFS surfaces, namely cubic, Romanesco broccoli, and sphereflake were fabricated using two-photon direct laser writing. Contact angle measurements were performed on the multiscale fractal surfaces to characterize their wetting properties. Average contact angles ranged from 66.8° for the smooth control surface to 0° for one of the fractal surfaces. The change in wetting behavior was attributed to modification of the interfacial surface properties due to the inclusion of 3-dimensional hierarchical fractal nanostructures. However, this behavior does not exactly obey existing surface wetting models in the literature. Potential applications for these types of surfaces in physical and biological sciences are also discussed.

Thermo-fluid dynamic analysis of wet compression process

Energy Technology Data Exchange (ETDEWEB)

Mohan, Abhay; Kim, Heuy Dong [School of Mechanical Engineering, Andong National University, Andong (Korea, Republic of); Chidambaram, Palani Kumar [FMTRC, Daejoo Machinery Co. Ltd., Daegu (Korea, Republic of); Suryan, Abhilash [Dept. of Mechanical Engineering, College of Engineering Trivandrum, Kerala (India)

2016-12-15

Wet compression systems increase the useful power output of a gas turbine by reducing the compressor work through the reduction of air temperature inside the compressor. The actual wet compression process differs from the conventional single phase compression process due to the presence of latent heat component being absorbed by the evaporating water droplets. Thus the wet compression process cannot be assumed isentropic. In the current investigation, the gas-liquid two phase has been modeled as air containing dispersed water droplets inside a simple cylinder-piston system. The piston moves in the axial direction inside the cylinder to achieve wet compression. Effects on the thermodynamic properties such as temperature, pressure and relative humidity are investigated in detail for different parameters such as compression speeds and overspray. An analytical model is derived and the requisite thermodynamic curves are generated. The deviations of generated thermodynamic curves from the dry isentropic curves (PV{sup γ} = constant) are analyzed.

Thermo-fluid dynamic analysis of wet compression process

International Nuclear Information System (INIS)

Mohan, Abhay; Kim, Heuy Dong; Chidambaram, Palani Kumar; Suryan, Abhilash

2016-01-01

Wet compression systems increase the useful power output of a gas turbine by reducing the compressor work through the reduction of air temperature inside the compressor. The actual wet compression process differs from the conventional single phase compression process due to the presence of latent heat component being absorbed by the evaporating water droplets. Thus the wet compression process cannot be assumed isentropic. In the current investigation, the gas-liquid two phase has been modeled as air containing dispersed water droplets inside a simple cylinder-piston system. The piston moves in the axial direction inside the cylinder to achieve wet compression. Effects on the thermodynamic properties such as temperature, pressure and relative humidity are investigated in detail for different parameters such as compression speeds and overspray. An analytical model is derived and the requisite thermodynamic curves are generated. The deviations of generated thermodynamic curves from the dry isentropic curves (PV γ = constant) are analyzed

Wetting characteristics of 3-dimensional nanostructured fractal surfaces

International Nuclear Information System (INIS)

Davis, Ethan; Liu, Ying; Jiang, Lijia; Lu, Yongfeng; Ndao, Sidy

2017-01-01

Highlights: • Hierarchically structured surfaces were fabricated on the micro/nano-scale. • These structures reduced the contact angle of the inherently hydrophilic material. • Similar surfaces have applications in two-phase heat transfer and microfluidics. - Abstract: This article reports the fabrication and wetting characteristics of 3-dimensional nanostructured fractal surfaces (3DNFS). Three distinct 3DNFS surfaces, namely cubic, Romanesco broccoli, and sphereflake were fabricated using two-photon direct laser writing. Contact angle measurements were performed on the multiscale fractal surfaces to characterize their wetting properties. Average contact angles ranged from 66.8° for the smooth control surface to 0° for one of the fractal surfaces. The change in wetting behavior was attributed to modification of the interfacial surface properties due to the inclusion of 3-dimensional hierarchical fractal nanostructures. However, this behavior does not exactly obey existing surface wetting models in the literature. Potential applications for these types of surfaces in physical and biological sciences are also discussed.

Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity when heat removing from a power transformer of combined heat and power station

Directory of Open Access Journals (Sweden)

Nurpeiis Atlant

2017-01-01

Full Text Available Numerical analyses of the effect of a biphasic thermosyphon vapor channel sizes on the heat transfer intensity was conducted when heat removing from an oil tank of a power transformer of combined heat and power station (CHP. The power transformer cooling system by the closed biphasic thermosyphon was proposed. The mathematical modeling of heat transfer and phase transitions of coolant in the thermosyphon was performed. The problem of heat transfer is formulated in dimensionless variables “velocity vorticity vector – current function – temperature” and solved by finite difference method. As a result of numerical simulation it is found that an increase in the vapor channel length from 0.15m to 1m leads to increasing the temperature difference by 3.5 K.

Dispersant Application during SG Wet Layup at SK Unit 1

Energy Technology Data Exchange (ETDEWEB)

Kwon, Hyukchul; Lee, Dooho; Sung, Kibang [KHNP Central Institute, Daejeon (Korea, Republic of)

2014-05-15

The corrosion products in the feedwater are deposited onto the steam generators (SGs) despite the effort to control them within limit of impurity. This deposit is one of causes for occurrence of SCC (Stress Corrosion Cracking), water level fluctuation and further corrosion of SGs. To minimize corrosion and remove deposit, the nuclear power plants apply high pH to the secondary system and SG chemical cleaning, respectively. But these methods can be costly and carry risks of extended outages or incomplete cleaning. Another method is an on-line dispersant application. The role of dispersant is to make deposit suspended in the SGs. Then, the suspended deposit is discharged to the blowdown system. The iron removal is increased in the blowdown system during the dispersant application. Additional significant benefit in the form of reduced corrosion product transport may be obtained through applying dispersant in the SGs wet lay operational mode. This method helps to reduce the total SGs loading without affecting critical outage activities and with minimal additional effort on the part of the utilities. This study provides the results of the dispersant application trial during the SG wet layup at SK Unit 1. As the PAA concentrations were increased, the corrosion rates of Alloy 690 and SA 106 Gr.B were increased. The corrosion rate of Alloy 690 was 2 times less than that of SA 106 Gr.B at 100 ppm of PAA based on the electrochemical experimental. There were no significant feasibility problems with application of PAA during the SG wet layup. The reasonable estimation of the additional mass removed by the presence of PAA during SGs wet layup is 460 g. The iron removal depended on PAA concentration injected based on the comparative results of the SK Unit 1 and TMI-1. It is expected that injection of PAA into the SG result in a significant decrease in the amount of iron transported to the SGs during the startup.

Numerical simulation of flow field in cooling tower of passive residual heat removal system of HTGR

International Nuclear Information System (INIS)

Li Xiaowei; Zhang Li; Wu Xinxin; He Shuyan

2011-01-01

Environmental wind will influence the working conditions of natural convection cooling tower. The velocity and temperature fields in the natural convection cooling tower of the HTGR residual heat removal system at different environmental wind velocities were numerically simulated. The results show that, if there is no wind baffle, the flow in the cooling tower is blocked when environmental wind velocity is higher than 6 m/s, residual heat can hardly be removed, and when wind velocity is higher than 9 m/s, the air even flow downwards in the tower, so wind baffle is very necessary. With the wind baffle installed, the cooling tower works well at the wind speed even higher than 9 m/s. The optimum baffle size and positions are also analyzed. (authors)

Cosmetic Cleansing Oil Absorption by Soft Contact Lenses in Dry and Wet Conditions.

Science.gov (United States)

Tsukiyama, Junko; Miyamoto, Yuko; Kodama, Aya; Fukuda, Masahiko; Shimomura, Yoshikazu

2017-09-01

Previous reports showed that cosmetic cleansing oil for removing makeup, which contains mineral oil and surfactant, can deform some silicone hydrogel contact lenses (SHCLs) when applied directly to the lenses, although plasma-coated SHCLs (lotrafilcon A and B) were not affected. In the present study, we investigated hydrogel lenses and SHCLs in both wet and dry conditions. Several brands of hydrogel and SHCLs were immersed in a cleansing oil solution containing Sudan Black B for 5 min under wet and dry conditions. The lenses under the wet condition were simply picked up from the saline, whereas those under the dry condition were blotted with paper wipes. After immersing, the excess solution remaining on the lenses was removed by finger rubbing with a multipurpose solution. The lenses were then examined using a stereomicroscope, and their mean brightness was measured and compared. The cosmetic cleansing oil was not absorbed by the hydrogel lenses under wet or dry conditions. However, four of seven brands of SHCLs absorbed the cosmetic cleansing oil under both conditions (dry and wet), whereas asmofilcon A absorbed it only under the dry condition. Lotrafilcon B and delefilcon A did not absorb cleansing oil even under the dry condition. Hydrogel lenses resist cosmetic cleansing oil. However, SHCLs have different degrees of resistance depending on the lens material. Some SHCLs absorbed cosmetic cleansing oil more under dry conditions than under wet conditions.

Reliability analysis on passive residual heat removal of AP1000 based on Grey model

Energy Technology Data Exchange (ETDEWEB)

Qi, Shi; Zhou, Tao; Shahzad, Muhammad Ali; Li, Yu [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Jiang, Guangming [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Laboratory

2017-06-15

It is common to base the design of passive systems on the natural laws of physics, such as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can affect the function of the system. The reliability of the passive residual heat removal is analysed.

Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Venker, Jeanne

2015-03-31

The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO{sub 2}) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO{sub 2} Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO{sub 2} Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO{sub 2} Brayton cycles and to evaluate the introduced

A value/impact assessment for alternative decay heat removal systems

International Nuclear Information System (INIS)

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

1984-01-01

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

Carbonaceous Materials Obtained from Sewage Sludge for NO2 Removal under Wet Conditions at Room Temperature

International Nuclear Information System (INIS)

Pietrzak, R.

2010-01-01

The effect of the processes of carbonisation and activation on adsorbents obtained from sewage sludge and their sorption properties towards NO 2 were studied. Carbonaceous adsorbents were obtained by carbonisation of sewage sludge at 600 o C for four different times 30, 60, 90 and 120 min followed by activation of the carbonisates by CO 2 at 800 o C for 60 min. Adsorption of NO 2 was carried out in wet air. It has been shown that by appropriate thermal and chemical treatment of sludge, mesoporous adsorbents capable of NO 2 removal can be obtained. The sorption abilities of the carbonised and activated samples to adsorb NO 2 have been shown to increase with increased time of carbonisation and reach maximum for the carbonisation maintained for 90 min. Further increase in this time causes a decrease in the adsorption abilities of the samples. The sorption properties of the carbonisates have been proved to be determined by the chemical character of the surface, while those of the activated samples - by the porous structure. (author)

A mechanism for corrosion product deposition on the carbon steel piping in the residual heat removal system of BWRs

International Nuclear Information System (INIS)

Aizawa, Motohiro; Chiba, Yoshinori; Hosokawa, Hideyuki; Ohsumi, Katsumi; Uchida, Shunsuke; Ishizawa, Noboru

2002-01-01

The dose rate of the residual heat removal (RHR) piping has been considered to be caused by accumulation of insoluble (crud) radioactive corrosion products on carbon steel surfaces. Soft shutdown procedures (i.e., plant shutdown with moderate coolant temperature reduction rate) used to be applied to reduce crud radioactivity release from the fuel surface, but these are no longer used because of the need for shorter plant shutdown times. In order to apply other suitable countermeasures to reduce RHR dose rate, assessment of plant data, experiments on deposition of crud and ion species on carbon steel, and mass balance evaluation of radioactive corrosion products based on plant and laboratory data were carried out and the following findings were made. (1) Deposits of ion species on carbon steel surfaces of the RHR piping was much more numerous than for crud. (2) Ion species accumulation behavior on RHR piping, which is temperature dependent, can be evaluated with the calculation model used for the dehydration reaction of corrosion products generated during the wet lay-up period. (3) Deposition amounts could be reduced to 1/2.5 when the starting RHR system operation temperature was lowered from 155degC to 120degC. (author)

Wetting behaviour and reactivity between liquid Gd and ZrO2 substrate

Directory of Open Access Journals (Sweden)

Turalska P.

2017-01-01

Full Text Available The wetting behavior and reactivity between molten pure Gd and polycrystalline 3YSZ substrate (ZrO2 stabilized with 3 wt% of Y2O3were experimentally determined by a sessile drop method using a classical contact heating coupled with drop pushing procedure. The test was performed under an inert flowing gas atmosphere (Ar at two temperatures of 1362°C and 1412°C. Immediately after melting (Tm=1341°C, liquid Gd did not wet the substrate forming a contact angle of θ=141°. The non-wetting to wetting transition (θ < 90° took place after about 110 seconds of interaction and was accompanied by a sudden decrease in the contact angle value to 67°. Further heating of the couple to 1412 °C did not affect wetting (θ=67°±1°. The solidified Gd/3YSZ couple was studied by means of optical microscopy and scanning electron microscopy coupled with X-ray energy dispersive spectroscopy. Structural investigations revealed that the wettability in the Gd/3YSZ system is of a reactive nature associated with the formation of a continuous layer of a wettable reaction product Gd2Zr2O7.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

De Wild, P.J.; Nyqvist, R.G.; De Bruijn, F.A.; Stobbe, E.R. [ECN Hydrogen and Clean Fossil Fuels, Petten (Netherlands)

2006-02-15

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC)) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20 ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG.

Removal of sulphur-containing odorants from fuel gases for fuel cell-based combined heat and power applications

Energy Technology Data Exchange (ETDEWEB)

de Wild, P.J.; Nyqvist, R.G.; de Bruijn, F.A.; Stobbe, E.R. [Energy Research Centre of The Netherlands ECN, P.O. Box 1, 1755 ZG Petten (Netherlands)

2006-09-22

Natural gas (NG) and liquefied petroleum gas (LPG) are important potential feedstocks for the production of hydrogen for fuel cell-based (e.g. proton exchange membrane fuel cells (PEMFC) or solid oxide fuel Cells (SOFC) combined heat and power (CHP) applications. To prevent detrimental effects on the (electro)catalysts in fuel cell-based combined heat and power installations (FC-CHP), sulphur removal from the feedstock is mandatory. An experimental bench-marking study of adsorbents has identified several candidates for the removal of sulphur containing odorants at low temperature. Among these adsorbents a new material has been discovered that offers an economically attractive means to remove TetraHydroThiophene (THT), the main European odorant, from natural gas at ambient temperature. The material is environmentally benign, easy to use and possesses good activity (residual sulphur levels below 20ppbv) and capacity for the common odorant THT in natural gas. When compared to state-of-the-art metal-promoted active carbon the new material has a THT uptake capacity that is up to 10 times larger, depending on temperature and pressure. Promoted versions of the new material have shown potential for the removal of THT at higher temperatures and/or for the removal of other odorants such as mercaptans from natural gas or from LPG. (author)

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

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2014-01-01

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

Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

International Nuclear Information System (INIS)

Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

2015-01-01

The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

Removal kinetics for gaseous NO and SO2by an aqueous NaClO2solution mist in a wet electrostatic precipitator

KAUST Repository

Park, Hyun-Woo

2016-07-26

Removal kinetics for NO and SO2 by NaClO2 solution mist were investigated in a wet electrostatic precipitator. By varying the molar concentrations of NO, SO2, and NaClO2, the removal rates of NO and SO2 confirmed to range from 34.8 to 72.9 mmol/m3 s and 36.6 to 84.7 mmol/m3 s, respectively, at a fixed gas residence time of 0.25 s. The rate coefficients of NO and SO2 were calculated to be 0.679 (mmol/m3)−0.33 s−1 and 1.401 (mmol/m3)−0.1 s−1 based on the rates of the individual removal of NO and SO2. Simultaneous removal of NO and SO2 investigated after the evaluation of removal rates for their individual treatment was performed. At a short gas residence time, SO2 gas removed more quickly by a mist of NaClO2 solution than NO gas in simultaneous removal experiments. This is because SO2 gas, which has a relatively high solubility in solution, was absorbed more rapidly at the gas–liquid interface than NO gas. NO and SO2 gases were absorbed as nitrite (Formula presented.) and sulfite (Formula presented.) ions, respectively, by the NaClO2 solution mist at the gas–liquid interface. Then, (Formula presented.) and (Formula presented.) were oxidized to nitrate (Formula presented.) and sulfate (Formula presented.), respectively, by reactions with (Formula presented.), ClO2, HClO, and ClO in the liquid phase. © 2016 Informa UK Limited, trading as Taylor & Francis Group

Aerosol removal due to precipitation and wind forcings in Milan urban area

Science.gov (United States)

Cugerone, Katia; De Michele, Carlo; Ghezzi, Antonio; Gianelle, Vorne

2018-01-01

Air pollution represents a critical issue in Milan urban area (Northern Italy). Here, the levels of fine particles increase, overcoming the legal limits, mostly in wintertime, due to favourable calm weather conditions and large heating and vehicular traffic emissions. The main goal of this work is to quantify the aerosol removal effect due to precipitation at the ground. At first, the scavenging coefficients have been calculated for aerosol particles with diameter between 0.25 and 3 μm. The average values of this coefficient vary between 2 ×10-5 and 5 ×10-5 s-1. Then, the aerosol removal induced separately by precipitation and wind have been compared through the introduction of a removal index. As a matter of fact, while precipitation leads to a proper wet scavenging of the particles from the atmosphere, high wind speeds cause enhanced particle dispersion and dilution, that locally bring to a tangible decrease of aerosol particles' number. The removal triggered by these two forcings showed comparable average values, but different trends. The removal efficiency of precipitation lightly increases with the increase of particle diameters and vice versa happens with strong winds.

Study on grey theoretical model of passive residual heat removal system

International Nuclear Information System (INIS)

Zhou Tao; Yang Ruichang; Su, G.H.; Jia Dounan; Sugiyama, K.

2004-01-01

Natural Circulation Passive Residual Heat Removal System is treated as a Grey System by taking into account of its complexity and uncertainty of effect for factors each other. The magnitude and degree of some factors are confirmed by grey incidence analysis method; The one-one relationship of some variables is built by GM (1, 1) model; The relationship between key factor and other effect factors is built (1, 4) model. Grey model shows its more advantage of precision through comparing with multivariate model. (author)

Influence of wick properties in a vertical LHP on remove waste heat from electronic equipment

International Nuclear Information System (INIS)

Smitka, Martin; Nemec, Patrik; Malcho, Milan

2014-01-01

The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work is to develop porous wick of sintered nickel powder with different grain sizes. These porous wicks were used in LHP and there were performed a series of measurements to remove waste heat from the insulated gate bipolar transistor (IGBT)

Influence of wick properties in a vertical LHP on remove waste heat from electronic equipment

Energy Technology Data Exchange (ETDEWEB)

Smitka, Martin, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk; Nemec, Patrik, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk; Malcho, Milan, E-mail: martin.smitka@fstroj.uniza.sk, E-mail: patrik.nemec@fstroj.uniza.sk, E-mail: milan.malcho@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engeneering, Univerzitna 1, 010 26 Žilina (Slovakia)

2014-08-06

The loop heat pipe is a vapour-liquid phase-change device that transfers heat from evaporator to condenser. One of the most important parts of the LHP is the porous wick structure. The wick structure provides capillary force to circulate the working fluid. To achieve good thermal performance of LHP, capillary wicks with high permeability and porosity and fine pore radius are expected. The aim of this work is to develop porous wick of sintered nickel powder with different grain sizes. These porous wicks were used in LHP and there were performed a series of measurements to remove waste heat from the insulated gate bipolar transistor (IGBT)

Effect of short-term heat acclimation training on kinetics of lactate removal following maximal exercise.

Science.gov (United States)

Dileo, Tsavis D; Powell, Jeffrey B; Kang, Hyoung K; Roberge, Raymond J; Coca, Aitor; Kim, Jung-Hyun

2016-01-01

Heat acclimation (HA) evokes numerous physiological adaptations, improves heat tolerance and has also been shown to enhance lactate (LA) responses during exercise, similar to that seen with endurance training. The purpose of this study was to examine whether HA improves the body's ability to remove LA during recovery following maximal exercise. Ten healthy men completed two trials of maximal treadmill exercise (pre- and post-HA) separated by 5 days of HA. Each day of HA consisted of two 45 minute periods of cycling at ~50% VO2max separated by a 15min rest period in an environmental chamber (T(db) 45° C, RH 20%). In pre-/post-HA trials, venous blood was collected during 60 minutes of recovery to determine LA concentrations and removal kinetics (A2: amplitude and y2: velocity constant) using bi-exponential curve fitting. Physiological adaptation to heat was significantly developed during HA, as evidenced by end-exercise T(re) (DAY1 vs. 5) (38.89±0.56 vs. 38.66±0.44° C), T(sk) (38.07±0.51 vs. 37.66±0.48° C), HR (175.0±9.9 vs. 165.0±18.5 beats·min(-1)), and sweat rate (1.24 ±.26 vs. 1.47 ±0.27 L·min(-1)) (PLA concentrations (LA(0min): 8.78±1.08 vs. 8.69±1.23; LA(peak): 10.97±1.77 vs. 10.95±1.46; and La(60min); 2.88±0.82 vs. 2.96±0.93 mmol·L(-1)) or removal kinetics (A2: -13.05±7.05 vs -15.59±7.90 mmol.L(-1) and y2: 0.02±0.01 vs. 0.03±.01 min(-1)). The present study concluded that, while effective in inducing thermo-physiological adaptations to heat stress, short-term HA does not improve the body's ability to remove LA following maximal exercise. Therefore, athletes and workers seeking faster LA recovery from intense physical activity may not benefit from short-term HA.

Wetting Behavior and Reactivity of Molten Silicon with h-BN Substrate at Ultrahigh Temperatures up to 1750 °C

Science.gov (United States)

Polkowski, Wojciech; Sobczak, Natalia; Nowak, Rafał; Kudyba, Artur; Bruzda, Grzegorz; Polkowska, Adelajda; Homa, Marta; Turalska, Patrycja; Tangstad, Merete; Safarian, Jafar; Moosavi-Khoonsari, Elmira; Datas, Alejandro

2017-12-01

For a successful implementation of newly proposed silicon-based latent heat thermal energy storage systems, proper ceramic materials that could withstand a contact heating with molten silicon at temperatures much higher than its melting point need to be developed. In this regard, a non-wetting behavior and low reactivity are the main criteria determining the applicability of ceramic as a potential crucible material for long-term ultrahigh temperature contact with molten silicon. In this work, the wetting of hexagonal boron nitride (h-BN) by molten silicon was examined for the first time at temperatures up to 1750 °C. For this purpose, the sessile drop technique combined with contact heating procedure under static argon was used. The reactivity in Si/h-BN system under proposed conditions was evaluated by SEM/EDS examinations of the solidified couple. It was demonstrated that increase in temperature improves wetting, and consequently, non-wetting-to-wetting transition takes place at around 1650 °C. The contact angle of 90° ± 5° is maintained at temperatures up to 1750 °C. The results of structural characterization supported by a thermodynamic modeling indicate that the wetting behavior of the Si/h-BN couple during heating to and cooling from ultrahigh temperature of 1750 °C is mainly controlled by the substrate dissolution/reprecipitation mechanism.

Wet-etch sequence optimisation incorporating time dependent chemical maintenance

NARCIS (Netherlands)

Kruif, B.J. de

2015-01-01

Wafer fabrication is the major cost contributor in semiconductor manufacturing. One of the steps in the fabrication is the removal of exposed layers in an automatic wet-etch station with chemicals. In time, these chemicals get polluted and their effectiveness decreases. Therefore, the chemicals in

Dynamic Temperature Rise Mechanism and Some Controlling Factors of Wet Clutch Engagement

Directory of Open Access Journals (Sweden)

Zhang Zhigang

2016-01-01

Full Text Available The friction transmission model of wet clutch is established to analyze the friction transmission mechanism of its engagement. The model is developed by applying both the average flow model and the elastic contact model between the friction disk and separator plate. The key components during wet clutch engagement are the separator plate, friction disk, and lubricant. The one-dimension transient models of heat transfer in radial direction for the three components are built on the basis of the heat transfer theory and the conservation law of energy. The friction transmission model and transient heat transfer models are coupled and solved by using the Runge-Kutta numerical method, and the radial temperature distribution and their detailed parametric study for the three components are conducted separately. The simulation results show that the radial temperature for the three components rises with the increase of radius in engagement. The changes in engagement pressure, lubricant viscosity, friction lining permeability, combined surface roughness RMS, equivalent elasticity modulus, difference between dynamic and static friction coefficients, and lubricant flow have important influence on the temperature rise characteristics. The proposed models can get better understanding of the dynamic temperature rise characteristics of wet clutch engagement.

Design considerations for wet flue gas desulfurization systems - wet scrubber hardware issues

Energy Technology Data Exchange (ETDEWEB)

Hurwitz, H.

1994-12-31

About 20 years ago the first wet flue gas desulfurization systems installed on coal fired utility boilers in the United States were experiencing extreme operating problems. In addition to their failure to achieve the necessary SO{sub 2} removal efficiencies, these FGD systems required a major investment in maintenance, both material and labor, just to remain operational. These first generation systems demonstrated that a lack of understanding of the chemistry and operating conditions of wet flue gas desulfurization can lead to diastrous results. As the air pollution control industry developed, both in the United States and in Japan, a second generation of FGD systems was introduced. These designs incorporated major improvements in both system chemistry control and in the equipment utilized in the process. Indeed, the successful introduction of utility gas desulfurization systems in Germany was possible only through the transfer of the technology improvements developed in the US and in Japan. Today, technology has evolved to a third generation of wet flue gas desulfurication systems and these systems are now offered worldwide through a series of international licensing agreements. The rapid economic growth and development in Asia and the Pacific Rim combined with existing problems in ambient air quality in these same geographic areas, has resulted in the use of advanced air pollution control systems; including flue gas desulfurization both for new utility units and for many retrofit projects. To meet the requirements of the utility industry, FGD systems must meet high standards of reliability, operability and performance. Key components in achieving these objectives are: FGD System reliability/operability/performance; FGD system supplier qualifications; process design; equipment selection. This paper will discuss each of the essential factors with a concentration on the equipment selection and wet scrubber hardware issues.

Investigation of bubble flow regimes in nucleate boiling of highly-wetting liquids

International Nuclear Information System (INIS)

Tong, W.; Bar-Cohen, A.; Simon, T.W.

1991-01-01

This paper describes an investigation of the bubble flow regimes in nucleate boiling of FC-72, a highly-wetting liquid. Theoretically analysis of vapor bubble generation and departure from the heated surface reveals that the heat fluxes required for the merging of consecutive bubbles, for highly-wetting liquids, lie in the upper range of the nucleate boiling heat flux. A visual and photographic study of nucleate boiling from sputtered platinum surfaces has supported the theoretical results and shown that the isolated bubble behavior extends to at least 50-80% of the critical heat flux, considerably higher than observed by others with water. Lateral coalescence of adjacent bubbles has been found to be a more likely cause of the termination of the isolated bubble regime. These findings suggest that thermal transport models which are based on isolated bubble behavior may be applicable to nearly the entire range of nucleate boiling of electronic cooling fluids

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

SASSYS validation with the EBR-II shutdown heat removal tests

International Nuclear Information System (INIS)

Herzog, J.P.

1989-01-01

SASSYS is a coupled neutronic and thermal hydraulic code developed for the analysis of transients in liquid metal cooled reactors (LMRs). The code is especially suited for evaluating of normal reactor transients -- protected (design basis) and unprotected (anticipated transient without scram) transients. Because SASSYS is heavily used in support of the IFR concept and of innovative LMR designs, such as PRISM, a strong validation base for the code must exist. Part of the validation process for SASSYS is analysis of experiments performed on operating reactors, such as the metal fueled Experimental Breeder Reactor -- II (EBR-II). During the course of a series of historic whole-plant experiments, EBR-II illustrated key safety features of metal fueled LMRs. These experiments, the Shutdown Heat Removal Tests (SHRT), culminated in unprotected loss of flow and loss of heat sink transients from full power and flow. Analysis of these and earlier SHRT experiments constitutes a vital part of SASSYS validation, because it facilitates scrutiny of specific SASSYS models and of integrated code capability. 12 refs., 11 figs

Full scale calcium bromide injection with subsequent mercury oxidation and removal within wet flue gas desulphurization system: Experience at a 700 MW coal-fired power facility

Science.gov (United States)

Berry, Mark Simpson

The Environmental Protection Agency promulgated the Mercury and Air Toxics Standards rule, which requires that existing power plants reduce mercury emissions to meet an emission rate of 1.2 lb/TBtu on a 30-day rolling average and that new plants meet a 0.0002 lb/GWHr emission rate. This translates to mercury removals greater than 90% for existing units and greater than 99% for new units. Current state-of-the-art technology for the control of mercury emissions uses activated carbon injected upstream of a fabric filter, a costly proposition. For example, a fabric filter, if not already available, would require a 200M capital investment for a 700 MW size unit. A lower-cost option involves the injection of activated carbon into an existing cold-side electrostatic precipitator. Both options would incur the cost of activated carbon, upwards of 3M per year. The combination of selective catalytic reduction (SCR) reactors and wet flue gas desulphurization (wet FGD) systems have demonstrated the ability to substantially reduce mercury emissions, especially at units that burn coals containing sufficient halogens. Halogens are necessary for transforming elemental mercury to oxidized mercury, which is water-soluble. Plants burning halogen-deficient coals such as Power River Basin (PRB) coals currently have no alternative but to install activated carbon-based approaches to control mercury emissions. This research consisted of investigating calcium bromide addition onto PRB coal as a method of increasing flue gas halogen concentration. The treated coal was combusted in a 700 MW boiler and the subsequent treated flue gas was introduced into a wet FGD. Short-term parametric and an 83-day longer-term tests were completed to determine the ability of calcium bromine to oxidize mercury and to study the removal of the mercury in a wet FGD. The research goal was to show that calcium bromine addition to PRB coal was a viable approach for meeting the Mercury and Air Toxics Standards rule

Design and transient analyses of emergency passive residual heat removal system of CPR1000

International Nuclear Information System (INIS)

Zhang, Y.P.; Qiu, S.Z.; Su, G.H.; Tian, W.X.

2012-01-01

Highlights: ► Designing an EPRHRs for CPR1000. ► Developing a RELAP model of the EPRHRs. ► The EPRHRs could take away the decay heat effectively. - Abstract: The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the

Modelling Particulate Removal in Tubular Wet Electrostatic Precipitators Using a Modified Drift Flux Model

Directory of Open Access Journals (Sweden)

S Ramechecandane

2016-09-01

Full Text Available Tubular electrostatic precipitators (ESP have been used in a number of chemical processing industries. The tubular ESPs have many advantages over conventional plate-plate and wire-plate ESPs. The present study is concerned with the numerical modeling of particulate removal in a tubular wet single-stage electrostatic precipitator (wESP. The geometric parameters of a model wESP and the corresponding inlet gas velocities for the wESP are chosen from available experimental data. In addition to the RNG k - ε model for the mean turbulent flow field inside the wESP, the Poisson equation for the electric field, the charge continuity equation and the concentration equation are solved sequentially to obtain a full-fledged solution to the problem under investigation. The proposed drift flux model is implemented in the opensource CFD code OpenFOAM®. The paper discusses the influence of the number of charges acquired by the particles and the corresponding inlet gas velocities on particle concentration distribution within the wESP. Two representative cases with monodispersed particles of 1 μm and 10 μm diameter are considered for the numerical analysis. It is seen from the present analysis that the number of units of charge on particles, the particle size and the inlet gas velocities play a vital role in determining the efficiency of electrostatic precipitation.

Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

An experimental investigation on the airside performance of fin-and-tube heat exchangers having slit fins under wet condition

International Nuclear Information System (INIS)

Kim, Nae Hyun; Kim, Tae Hun

2015-01-01

In this study, the heat transfer and friction characteristics of the 5.3 mm O.D. slit-finned heat exchangers under wet condition have been experimentally investigated. Plain-finned heat exchangers having the same 5.3 mm O.D. tubes are also tested for comparison purpose. The effect of fin pitch on j and f factor is negligible. Slit fin samples yield higher j and f factors than plain fin samples. For one row configuration, the average f factor ratio between slit fin sample and plain fin sample is 2.18. The ratio increases to 2.41 for two row configuration, and to 2.65 for three row configuration. As for the j factor, the ratios are approximately the same (1.61, 1.70 and 1.71 for one, two and three row configuration). Both j and f factor increase as the number of tube row decreases. The same trend is observed for the plain fin samples. At high Reynolds numbers, the j/f ratios of the slit fin are approximately the same as those of the plain fin. At low Reynolds numbers, the j/f ratios of the slit fin are smaller than those of plain fin. Data are compared with existing correlations.

Heat exchanger device and method for heat removal or transfer

Science.gov (United States)

Koplow, Jeffrey P

2013-12-10

Systems and methods for a forced-convection heat exchanger are provided. In one embodiment, heat is transferred to or from a thermal load in thermal contact with a heat conducting structure, across a narrow air gap, to a rotating heat transfer structure immersed in a surrounding medium such as air.

GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

International Nuclear Information System (INIS)

Ramachandran, S.; Krause, M.; Nguyen, T.

2003-01-01

This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-02-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

International Nuclear Information System (INIS)

Gautier, G.M.

1995-01-01

In the framework of research on diversified means for removing the residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system), which includes integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: it is independent of the state of the loops, even if the volume of water in the primary circuit is small, it is compatible with either a passive or an active operation mode, and compatible with any other decay heat removal systems. An evaluation is presented here of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of another system. The results of this evaluation show the interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system, no need for the use of a high pressure safety injection system. (author). 4 refs., 7 figs., 1 tab

Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Gautier, G.M. [Commissariat a l`Energie Atomique, Saint-Paul-Lez-Durance (France)

1995-09-01

In the framework of research on diversified means for removing residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system). This system consists of integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: such as independent from the state of the loops, low volume of water in the primary circuit, compatibility with either a passive or an active operation mode, and compatibility with any other decay heat removal systems. This paper presents an evaluation of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of any another system. The results of this evaluation show the potential interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system and no need for the use of a high pressure safety injection system.

Wetting properties of liquid lithium on lithium compounds

Energy Technology Data Exchange (ETDEWEB)

Krat, S.A., E-mail: stepan.krat@gmail.com [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Popkov, A.S. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Gasparyan, Yu. M.; Pisarev, A.A. [National Research Nuclear University MEPhI, Moscow (Russian Federation); Fiflis, Peter; Szott, Matthew; Christenson, Michael; Kalathiparambil, Kishor; Ruzic, David N. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States)

2017-04-15

Highlights: • Contact angles of liquid lithium and Li{sub 3}N, Li{sub 2}O, Li{sub 2}CO{sub 3} were measured. • Liquid lithium wets lithium compounds at relatively low temperatures: Li{sub 3}N at 257 °C, Li{sub 2}O at 259 °C, Li{sub 2}CO{sub 3} at 323 °C. • Li wets Li{sub 2}O and Li{sub 3}N better than previously measured fusion-relevant materials (W, Mo, Ta, TZM, stainless steel). • Li wets Li{sub 2}CO{sub 3} better than most previously measured fusion-relevant materials (W, Mo, Ta). - Abstract: Liquid metal plasma facing components (LMPFC) have shown a potential to supplant solid plasma facing components materials in the high heat flux regions of magnetic confinement fusion reactors due to the reduction or elimination of concerns over melting, wall damage, and erosion. To design a workable LMPFC, one must understand how liquid metal interacts with solid underlying structures. Wetting is an important factor in such interaction, several designs of LMPFC require liquid metal to wet the underlying solid structures. The wetting of lithium compounds (lithium nitride, oxide, and carbonate) by 200 °C liquid lithium at various surface temperature from 230 to 330 °C was studied by means of contact angle measurements. Wetting temperatures, defined as the temperature above which the contact angle is less than 90°, were measured. The wetting temperature was 257 °C for nitride, 259 °C for oxide, and 323 °C for carbonate. Surface tensions of solid lithium compounds were calculated from the contact angle measurements.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Mitigation Measures Following a Loss-of-Residual-Heat-Removal Event During Shutdown

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

2000-01-01

The transient following a loss-of-residual-heat-removal event during shutdown was analyzed to determine the containment closure time (CCT) to prevent uncontrolled release of fission products and the gravity-injection path and rate (GIPR) for effective core cooling using the RELAP5/MOD3.2 code. The plant conditions of Yonggwang Units 3 and 4, a pressurized water reactor (PWR) of 2815-MW(thermal) power in Korea, were reviewed, and possible event sequences were identified. From the CCT analysis for the five cases of typical plant configurations, it was estimated for the earliest CCT to be 40 min after the event in a case with a large cold-leg opening and emptied steam generators (SGs). However, the case with water-filled SGs significantly delayed the CCT through the heat removal to the secondary side. From the GIPR analysis for the six possible gravity-injection paths from the refueling water storage tank (RWST), the case with the injection point and opening on the other leg side was estimated to be the most suitable path to avoid core boiling. In addition, from the sensitivity study, it was evaluated for the plant to be capable of providing the core cooling for the long-term transient if nominal RWST water is available. As a result, these analysis methods and results will provide useful information in understanding the plant behavior and preparing the mitigation measures after the event, especially for Combustion Engineering-type PWR plants. However, to directly apply the analysis results to the emergency procedure for such an event, additional case studies are needed for a wide range of operating conditions such as reactor coolant inventory, RWST water temperature, and core decay heat rate

A three-region conduction-controlled rewetting analysis by the Heat Balance Integral Method

International Nuclear Information System (INIS)

Sahu, S.K.; Das, P.K.; Bhattacharyya, S.

2009-01-01

Conduction-controlled rewetting of two-dimensional objects is analyzed by the Heat Balance Integral Method (HBIM) considering three distinct regions: a dry region ahead of wet front, the sputtering region immediately behind the wet front and a continuous film region further upstream. The HBIM yields solutions for wet front velocity, sputtering length and temperature field with respect to wet front. Employing this method, it is seen that heat transfer mechanism is dependent upon two temperature parameters. One of them characterizes the initial wall temperature while the other specifies the range of temperature for sputtering region. Additionally, the mechanism of heat transfer is found to be dependent on two Biot numbers comprising a convective heat transfer in the wet region and a boiling heat transfer in the sputtering region. The present solution exactly matches with the one-dimensional analysis of K.H. Sun, G.E. Dix, C.L. Tien [Cooling of a very hot vertical surface by falling liquid film, ASME J. Heat Transf. 96 (1974) 126-131] for low Biot numbers. Good agreement with experimental results is also observed. (authors)

PANDA passive decay heat removal transient test results

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

Estimating Wet Bulb Globe Temperature Using Standard Meteorological Measurements

International Nuclear Information System (INIS)

Hunter, C.H.

1999-01-01

The heat stress management program at the Department of Energy''s Savannah River Site (SRS) requires implementation of protective controls on outdoor work based on observed values of wet bulb globe temperature (WBGT). To ensure continued compliance with heat stress program requirements, a computer algorithm was developed which calculates an estimate of WBGT using standard meteorological measurements. In addition, scripts were developed to generate a calculation every 15 minutes and post the results to an Intranet web site

SASSYS-1 modelling of RVACS/RACS heat removal in an LMR

International Nuclear Information System (INIS)

Dunn, F.E.

1987-01-01

The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This air-side RVACS/RACS model is coupled with the sodium-side primary loop thermal hydraulics model in SASSYS-1 to give a complete treatment of the problem. Application of this model to an unprotected loss-of-flow event in the PRISM rector shows that in the long run the RVACS cooling is sufficient to prevent unacceptably high system temperatures in this case

Evaluation of wet bulb globe temperature index for estimation of heat strain in hot/humid conditions in the Persian Gulf.

Science.gov (United States)

Dehghan, Habibolah; Mortazavi, Seyed Bagher; Jafari, Mohammad J; Maracy, Mohammad R

2012-12-01

Heat exposure among construction workers in the Persian Gulf region is a serious hazard for health. The aim of this study was to evaluate the performance of wet bulb globe temperature (WBGT) Index for estimation of heat strain in hot/humid conditions by the use of Physiological Strain Index (PSI) as the gold standard. This cross-sectional study was carried out on 71 workers of two Petrochemical Companies in South of Iran in 2010 summer. The WBGT index, heart rate, and aural temperature were measured by Heat Stress Monitor (Casella Microtherm WBGT), Heart Rate Monitor (Polar RS100), and Personal Heat Strain Monitor (Questemp II), respectively. The obtained data were analyzed with descriptive statistics and Pearson correlation analysis. The mean (SD) of WBGT values was 33.1 (2.7). The WBGT values exceed from American Conference of Governmental Industrial Hygienists (ACGIH) standard (30°C) in 96% work stations, whereas the PSI values were more than 5.0 (moderate strain) in 11% of workstations. The correlation between WBGT and PSI values was 0.61 (P = 0.001). When WBGT values were less and more than 34°C, the mean of PSI was 2.6 (low strain) and 5.2 (moderate strain), respectively. In the Persian Gulf weather, especially hot and humid in the summer months, due to the WBGT values exceeding 30°C (in 96% of cases) and weak correlation between WBGT and PSI, the work/rest cycles of WBGT Index is not suitable for heat stress management. Therefore, in Persian Gulf weather, heat stress evaluation based on physiologic variables may have higher validity than WBGT index.

Evaluation of wet bulb globe temperature index for estimation of heat strain in hot/humid conditions in the Persian Gulf

Directory of Open Access Journals (Sweden)

Habibolah Dehghan

2012-01-01

Full Text Available Background: Heat exposure among construction workers in the Persian Gulf region is a serious hazard for health. The aim of this study was to evaluate the performance of wet bulb globe temperature (WBGT Index for estimation of heat strain in hot/humid conditions by the use of Physiological Strain Index (PSI as the gold standard. Material and Methods : This cross-sectional study was carried out on 71 workers of two Petrochemical Companies in South of Iran in 2010 summer. The WBGT index, heart rate, and aural temperature were measured by Heat Stress Monitor (Casella Microtherm WBGT, Heart Rate Monitor (Polar RS100, and Personal Heat Strain Monitor (Questemp II, respectively. The obtained data were analyzed with descriptive statistics and Pearson correlation analysis. Results: The mean (SD of WBGT values was 33.1 (2.7. The WBGT values exceed from American Conference of Governmental Industrial Hygienists (ACGIH standard (30°C in 96% work stations, whereas the PSI values were more than 5.0 (moderate strain in 11% of workstations. The correlation between WBGT and PSI values was 0.61 ( P = 0.001. When WBGT values were less and more than 34°C, the mean of PSI was 2.6 (low strain and 5.2 (moderate strain, respectively. Conclusion: In the Persian Gulf weather, especially hot and humid in the summer months, due to the WBGT values exceeding 30°C (in 96% of cases and weak correlation between WBGT and PSI, the work/rest cycles of WBGT Index is not suitable for heat stress management. Therefore, in Persian Gulf weather, heat stress evaluation based on physiologic variables may have higher validity than WBGT index.

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

International Nuclear Information System (INIS)

Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

2010-01-01

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

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

International Nuclear Information System (INIS)

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

1989-04-01

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

Effect of heat transfer in the fog region during core reflooding

International Nuclear Information System (INIS)

Rouai, N. M.; El-sawy, H. M.

1993-01-01

Core reflooding following a loss of coolant accident (LOCA) in a pressurized water reactor (PWR) received considerable attention during the past thirty years. In this paper a one dimensional model is used to study the effect of the heat transfer in the fog region ahead of the wet front reflooding rate of a cylindrical fuel element following a LOCA in a PWR. The heat conduction equation in the cladding is solved in coordinate system moving with the wet front under a variety of condition to investigate the effects of such parameters as the initial cladding surface temperature, the decay heat generation rate in the fuel and the mode of heat transfer prevailing. The cladding surface is divided into three axial regions according to the mechanism of heat transfer, namely, a boiling region behind the wet front, a fog region ahead of the wet front and a dry region further downstream of the wet front. The effect of changing the heat transfer coefficient in the fog region on the rewetting rate and on the fog length is investigated. The results of this simple model show that increasing the heat transfer in the fog region increases the rewetting velocity and consequently decreases the fog length. The results are in general agreement with a more accurate two-dimensional model and experimental data. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-07-01

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

Unavailability of the residual system heat removal of Angra 1 by Bayesian networks considering dependent failures

International Nuclear Information System (INIS)

Gomes, Many R.S.; Melo, Paulo F.F.F. e

2015-01-01

This work models by Bayesian networks the residual heat removal system (SRCR) of Angra I nuclear power plant, using fault tree mapping for systematically identifying all possible modes of occurrence caused by a large loss of coolant accident (large LOCA). The focus is on dependent events, such as the bridge system structure of the residual heat removal system and the occurrence of common-cause failures. We used the Netica™ tool kit, Norsys Software Corporation and Python 2.7.5 for modeling Bayesian networks and Microsoft Excel for modeling fault trees. Working with dependent events using Bayesian networks is similar to the solutions proposed by other models, beyond simple understanding and ease of application and modification throughout the analysis. The results obtained for the unavailability of the system were satisfactory, showing that in most cases the system will be available to mitigate the effects of an accident as described above. (author)

Unavailability of the residual system heat removal of Angra 1 by Bayesian networks considering dependent failures

Energy Technology Data Exchange (ETDEWEB)

Gomes, Many R.S.; Melo, Paulo F.F.F. e, E-mail: mgomes@con.ufrj.br, E-mail: frutuoso@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Programa de Pos-Graduacao em Engenharia Nuclear

2015-07-01

This work models by Bayesian networks the residual heat removal system (SRCR) of Angra I nuclear power plant, using fault tree mapping for systematically identifying all possible modes of occurrence caused by a large loss of coolant accident (large LOCA). The focus is on dependent events, such as the bridge system structure of the residual heat removal system and the occurrence of common-cause failures. We used the Netica™ tool kit, Norsys Software Corporation and Python 2.7.5 for modeling Bayesian networks and Microsoft Excel for modeling fault trees. Working with dependent events using Bayesian networks is similar to the solutions proposed by other models, beyond simple understanding and ease of application and modification throughout the analysis. The results obtained for the unavailability of the system were satisfactory, showing that in most cases the system will be available to mitigate the effects of an accident as described above. (author)

Effect of Heat Drawing Process on Mechanical Properties of Dry-Jet Wet Spun Fiber of Linear Low Density Polyethylene/Carbon Nanotube Composites

Directory of Open Access Journals (Sweden)

Jong Won Kim

2017-01-01

Full Text Available Polyethylene is one of the most commonly used polymer materials. Even though linear low density polyethylene (LLDPE has better mechanical properties than other kinds of polyethylene, it is not used as a textile material because of its plastic behavior that is easy to break at the die during melt spinning. In this study, LLDPE fibers were successfully produced with a new approach using a dry-jet wet spinning and a heat drawing process. The fibers were filled with carbon nanotubes (CNTs to improve the strength and reduce plastic deformation. The crystallinity, degree of orientation, mechanical properties (strength to yield, strength to break, elongation at break, and initial modulus, electrical conductivity, and thermal properties of LLDPE fibers were studied. The results show that the addition of CNTs improved the tensile strength and the degree of crystallinity. The heat drawing process resulted in a significant increase in the tensile strength and the orientation of the CNTs and polymer chains. In addition, this study demonstrates that the heat drawing process effectively decreases the plastic deformation of LLDPE.

User's manual for the BNW-II optimization code for dry/wet-cooled power plants

International Nuclear Information System (INIS)

Braun, D.J.; Bamberger, J.A.; Braun, D.J.; Faletti, D.W.; Wiles, L.E.

1978-05-01

The User's Manual describes how to operate BNW-II, a computer code developed by the Pacific Northwest Laboratory (PNL) as a part of its activities under the Department of Energy (DOE) Dry Cooling Enhancement Program. The computer program offers a comprehensive method of evaluating the cost savings potential of dry/wet-cooled heat rejection systems. Going beyond simple ''figure-of-merit'' cooling tower optimization, this method includes such items as the cost of annual replacement capacity, and the optimum split between plant scale-up and replacement capacity, as well as the purchase and operating costs of all major heat rejection components. Hence the BNW-II code is a useful tool for determining potential cost savings of new dry/wet surfaces, new piping, or other components as part of an optimized system for a dry/wet-cooled plant

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Enhanced Hg{sup 2+} removal and Hg{sup 0} re-emission control from wet fuel gas desulfurization liquors with additives

Energy Technology Data Exchange (ETDEWEB)

Tingmei Tang; Jiang Xu; Rongjie Lu; Jingjing Wo; Xinhua Xu [Zhejiang University, Hangzhou (China). Department of Environmental Engineering

2010-12-15

Secondary atmospheric pollutions may result from wet flue gas desulfurization (FGD) systems caused by the reduction of Hg{sup 2+} to Hg{sup 0}. The present study employed three agents: Na{sub 2}S, 2,4,6-trimercaptotiazine, trisodium salt nonahydrate (TMT) and sodium dithiocarbamate (DTCR) to precipitate aqueous Hg{sup 2+} in simulated desulfurization solutions. The effects of the precipitator's dosing quantity, the initial pH value, the reaction temperature, the concentrations of Cl{sup -} and other metal ions (e.g. Cu{sup 2+} and Pb{sup 2+}) on Hg{sup 2+} removal were studied. A linear relationship was observed between Hg{sup 2+} removal efficiency and the increasing precipitator's doses along with initial pH. The addition of chloride and metal ions impaired the Hg{sup 2+} removal from solutions due to the complexation of Cl{sup -} and Hg{sup 2+} as well as the chelating competition between Hg{sup 2+} and other metal ions. Based on a comprehensive comparison of the treatment effects, DTCR was found to be the most effective precipitating agent. Moreover, all the precipitating agents were potent enough to inhibit Hg{sup 2+} reduction as well as Hg{sup 0} re-emission from FGD liquors. More than 90% Hg{sup 2+} was captured by precipitating agents while Hg{sup 2+} reduction efficiency decreased from 54% to just less than 3%. The additives could efficiently control the secondary Hg{sup 0} pollution from FGD liquors. 21 refs., 6 figs.

New advances in wet scrubbing improvement efficiencies

Energy Technology Data Exchange (ETDEWEB)

Keen, A.R. [Altech Group, Toronto, ON (Canada)

2000-07-01

Wet scrubbing systems are the most versatile and cost efficient of all air pollution abatement technologies. This paper presented System REITHER{sup TM} which is a new generation of venturi scrubber. The advantages of this design are that it is simple and compact, has high removal efficiencies for sub-micron dusts or aerosols and it is flexible to handle any mass flow rate. It also provides high and constant reliability, is easy to control and has the potential to absorb gaseous pollutants. Another advantage is that it can handle corrosive streams through corrosion resistant materials. Innovations in wet scrubbing have made it possible to provide reliable and efficient separation of fine particles, corrosive aerosols and gases. New technology provides industrial engineers with a cost effective option when control air emissions is required. 1 fig.

Control technology for crystalline silica exposures in construction: wet abrasive blasting.

Science.gov (United States)

Golla, Vijay; Heitbrink, William

2004-03-01

This study was designed to document the effect that wet abrasive blasting has on reducing worker exposure to crystalline silica, which has been associated with silicosis and premature death. In this study, worker exposure to respirable crystalline silica was monitored during wet abrasive blasting on the exterior walls of a parking garage to remove surface concrete and expose the underlying aggregate. In this process a wet sand mix comprised of 80% dry sand and 20% water was used. Sampling and analysis revealed that the geometric mean respirable quartz concentration was 0.2 mg/m(3) for workers conducting abrasive blasting and 0.06 mg/m(3) for helpers. When abrasive blasting was conducted in areas that apparently had reduced natural ventilation, dust exposures appeared to increase. When compared with other published data, this case study suggests that wet abrasive blasting causes less exposure to crystalline silica than dry abrasive blasting.

Theory and design of heat exchanger : air cooled plate, spiral heat exchanger

International Nuclear Information System (INIS)

Min, Ui Dong

1960-02-01

This book deals with air cooled heat exchanger, which introduces heat rejection system, wet surface cooler in new from, explanation of structure and design, materials, basic design like plenums chambers and fan ring, finned tube fouling factor, airflow in forced draft and fan design. It also tells of plate heat exchanger and spiral heat exchanger giving descriptions of summary, selection, basic design, device and safety function, maintenance, structure of plate heat exchanger, frames and connector plate and, basic things of spiral tube heat exchanger.

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

International Nuclear Information System (INIS)

Hayashi, Takao; Ishi, Takayuki; Hayakawa, Hitoshi; Ohashi, Kazutaka

1997-01-01

Experiments on the separated heat pipe system of variable conductance type, which enclose non-condensible gas, have been carried out with intention of applying such system to passive decay heat removal of the modular reactors such as HTR plant. Basic experiments have been carried out on the experimental apparatus consisting of evaporator, vapor transfer tube, condenser tube and return tube which returns the condensed liquid back to the evaporator. Water and methanol were examined as the working fluids and nitrogen gas was enclosed as the non-condensible gas. The behaviors of the system were examined for the parametric changes of the heat input under the various pressures of nitrogen gas initially enclosed, including the case without enclosing N 2 gas for the comparison. The results of the experiments shows very clear features of self control characteristics. The self control mechanism was made clear, that is, in such system in which the condensing area in the condenser expands automatically in accordance with the increase of the heat input to keep the system temperature nearly constant. The working temperature of the system are clearly dependent on the pressure of the non-condensable gas initially enclosed, with higher system working temperature with higher initial gas pressure enclosed. The analyses were done on water and methanol as the working fluids, which show very good agreement with the experimental results. A lot of attractive applications are expected including the self switching feature with minimum heat loss during normal operation with maintaining the sufficient heat removal at accidents. (author)

Scrubber-Integrated Wet Electrostatic Precipitator; Skrubberintegrerat vaatt elektrofilter, WESP

Energy Technology Data Exchange (ETDEWEB)

Andersson, Sven; Baefver, Linda; Davidsson, Kent; Pettersson, Jens; Schmidt, Hans; Strand, Michael; Yngvesson, Johan

2011-07-01

Combustion processes for heat and power production are an important source of sub-micron particle emissions, which cause enhanced health risks and premature deaths. To meet future requirements of economical and robust dust cleaning equipment, the Wet Electrostatic Precipitation (WESP) technology has been further developed in this project. A pilot scale slip stream WESP unit, installed by Goetaverken Miljoe, has been successfully installed and tested at the Renova Waste-to-Energy plant in Goeteborg, Sweden. The particles in the gas are charged by an ionizing electrode and collected in a concentric cylinder geometry. The WESP pilot consists of a unique combination of several existing technologies: it is integrated with a packed bed scrubber which means an ideally uniformly distributed gas flow in the WESP inlet. Furthermore, the WESP unit has a water cooled condensing collector, which facilitates continuous formation of a water film. The downward flowing water film transports the collected dust counter current to the upward flowing flue gas in order to minimize particle re-entrainment. The WESP is equipped with a high frequency transformer for stable voltage output and is fabricated in electrically conductive corrosion resistant Fibre Reinforced Plastic (FRP). The concentration of dust upstream of the WESP unit varied between 6.2 and 28 mg/Nm{sup 3} dry gas. All measured outlet dust concentrations were below 0.3 mg/Nm{sup 3} (dry gas, 11% O{sub 2}), which equals 3% of the applicable emission limit. The dust removal efficiency has been higher than 97% in all the dust measurements. The mean value of all the dust measurements was 15.2 mg/Nm{sup 3} upstream and 0.14 mg/Nm{sup 3} in downstream (both as dry gas, 11% O{sub 2}), which gives an average removal efficiency of slightly more than 99%. The removal efficiency increased with increasing inlet dust concentration, SO{sub 2} concentration and {Delta}T of the collector cooling. Chlorine, potassium, sodium, silicon and

Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Romanato, Luiz Sergio, E-mail: romanato@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade

2011-07-01

Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2011-01-01

Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

International Nuclear Information System (INIS)

Krasikov, E. A.

2012-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated

Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

Energy Technology Data Exchange (ETDEWEB)

Krasikov, E. A. [National Research Centre Kurchatov Inst., 1, Kurchatov Sq., Moscow, 123182 (Russian Federation)

2012-07-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which

Reliability evaluation of power supply and distribution for special heat removal systems in nuclear power stations

International Nuclear Information System (INIS)

Jazbec, D.

1982-01-01

An example of the power supply and distribution of a Special Emergency Heat Removal System (SEHR) shows how an engineering organization may, with the aid of the analytical method of min-cut sets optimize the system reliability. Herein are given the necessary simple calculation methods. (Auth.)

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

International Nuclear Information System (INIS)

Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

1993-01-01

The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

1993-02-01

The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

The steady-state modeling and optimization of a refrigeration system for high heat flux removal

International Nuclear Information System (INIS)

Zhou Rongliang; Zhang Tiejun; Catano, Juan; Wen, John T.; Michna, Gregory J.; Peles, Yoav; Jensen, Michael K.

2010-01-01

Steady-state modeling and optimization of a refrigeration system for high heat flux removal, such as electronics cooling, is studied. The refrigeration cycle proposed consists of multiple evaporators, liquid accumulator, compressor, condenser and expansion valves. To obtain more efficient heat transfer and higher critical heat flux (CHF), the evaporators operate with two-phase flow only. This unique operating condition necessitates the inclusion of a liquid accumulator with integrated heater for the safe operation of the compressor. Due to the projected incorporation of microchannels into the system to enhance the heat transfer in heat sinks, the momentum balance equation, rarely seen in previous vapor compression cycle heat exchangers modeling efforts, is utilized in addition to the mass and energy balance equations to capture the expected significant microchannel pressure drop witnessed in previous experimental investigations. Using the steady-state model developed, a parametric study is performed to study the effect of various external inputs on the system performance. The Pareto optimization is applied to find the optimal system operating conditions for given heat loads such that the system coefficient of performance (COP) is optimized while satisfying the CHF and other system operation constraints. Initial validation efforts show the good agreement between the experimental data and model predictions.

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

International Nuclear Information System (INIS)

1994-08-01

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

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

International Nuclear Information System (INIS)

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

1995-12-01

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

Methods and apparatus for catalytic hydrothermal gasification of biomass

Science.gov (United States)

Elliott, Douglas C.; Butner, Robert Scott; Neuenschwander, Gary G.; Zacher, Alan H.; Hart, Todd R.

2012-08-14

Continuous processing of wet biomass feedstock by catalytic hydrothermal gasification must address catalyst fouling and poisoning. One solution can involve heating the wet biomass with a heating unit to a temperature sufficient for organic constituents in the feedstock to decompose, for precipitates of inorganic wastes to form, for preheating the wet feedstock in preparation for subsequent separation of sulfur contaminants, or combinations thereof. Treatment further includes separating the precipitates out of the wet feedstock, removing sulfur contaminants, or both using a solids separation unit and a sulfur separation unit, respectively. Having removed much of the inorganic wastes and the sulfur that can cause poisoning and fouling, the wet biomass feedstock can be exposed to the heterogeneous catalyst for gasification.

Passive decay heat removal by natural air convection after severe accidents