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.

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.

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

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

Passive heat removal characteristics of SMART

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

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.

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)

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

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.

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

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

Assessment of alternate ion exchange resins for improved antimony removal from the primary heat transport system

Energy Technology Data Exchange (ETDEWEB)

Burany, R.; Suryanarayan, S.; Husain, A. [Kinectrics, Inc., Toronto, ON (Canada)

2015-07-01

Radiation fields around the CANDU heat transport system are a major contributor to worker dose during inspection, maintenance and refurbishment activities. While Co-60 is typically the dominant contributor to radiation fields in CANDU reactors, Sb-124, an activation product of antimony, is also a significant contributor, accounting for 5-20% of the radiation fields. The goal of this research project was to investigate resins for improved removal of antimony under both oxidizing and reducing conditions.Several candidate resins were tested and short-listed through a sequence of iterative testing. The results of the laboratory testing have identified potential candidates for improved antimony removal. Further testing is required to ensure compatibility with existing station resin specifications. (author)

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

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 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.

Milk protein-gum tragacanth mixed gels: effect of heat-treatment sequence.

Science.gov (United States)

Hatami, Masoud; Nejatian, Mohammad; Mohammadifar, Mohammad Amin; Pourmand, Hanieh

2014-01-30

The aim of this study was to investigate the role of the heat-treatment sequence of biopolymer mixtures as a formulation parameter on the acid-induced gelation of tri-polymeric systems composed of sodium caseinate (Na-caseinate), whey protein concentrate (WPC), and gum tragacanth (GT). This was studied by applying four sequences of heat treatment: (A) co-heating all three biopolymers; (B) heating the milk-protein dispersion and the GT dispersion separately; (C) heating the dispersion containing Na-caseinate and GT together and heating whey protein alone; and (D) co-heating whey protein with GT and heating Na-caseinate alone. According to small-deformation rheological measurements, the strength of the mixed-gel network decreased in the order: C>B>D>A samples. SEM micrographs show that the network of sample C is much more homogenous, coarse and dense than sample A, while the networks of samples B and D are of intermediate density. The heat-treatment sequence of the biopolymer mixtures as a formulation parameter thus offers an opportunity to control the microstructure and rheological properties of mixed gels. Copyright © 2013 Elsevier Ltd. All rights reserved.

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.)

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

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)

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

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.)

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

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

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

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

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)

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)

Removal and transformation of hexavalent chromium in sequencing ...

African Journals Online (AJOL)

The objectives of this study are to evaluate the efficiency of removal of hexavalent chromium (Cr(VI)) in a sequencing batch reactor (SBR) and to ascertain the fate of Cr(VI) in the treatment process. An SBR was operated with the FILL, REACT, SETTLE, DRAW and IDLE periods in the time ratio of 2:12:2:1.5:6.5 for a cycle ...

Operational strategies for nitrogen removal in granular sequencing batch reactor

International Nuclear Information System (INIS)

Chen, Fang-yuan; Liu, Yong-Qiang; Tay, Joo-Hwa; Ning, Ping

2011-01-01

This study investigated the effects of different operational strategies for nitrogen removal by aerobic granules with mean granule sizes of 1.5 mm and 0.7 mm in a sequencing batch reactor (SBR). With an alternating anoxic/oxic (AO) operation mode without control of dissolve oxygen (DO), the granular sludge with different size achieved the total inorganic nitrogen (TIN) removal efficiencies of 67.8-71.5%. While under the AO condition with DO controlled at 2 mg/l at the oxic phase, the TIN removal efficiency was improved up to 75.0-80.4%. A novel operational strategy of alternating anoxic/oxic combined with the step-feeding mode was developed for nitrogen removal by aerobic granules. It was found that nitrogen removal efficiencies could be further improved to 93.0-95.9% with the novel strategy. Obviously, the alternating anoxic/oxic strategy combined with step-feeding is the optimal way for TIN removal by granular sludge, which is independent of granule size.

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)

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)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

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.

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

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

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

Energy Technology Data Exchange (ETDEWEB)

Murav’ev, V. P., E-mail: murval1@mail.ru

2016-07-15

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

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.

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

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

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)

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)

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.

Expressed sequence tags from heat-shocked seagrass Zostera noltii (Hornemann) from its southern distribution range.

Science.gov (United States)

Massa, Sónia I; Pearson, Gareth A; Aires, Tânia; Kube, Michael; Olsen, Jeanine L; Reinhardt, Richard; Serrão, Ester A; Arnaud-Haond, Sophie

2011-09-01

Predicted global climate change threatens the distributional ranges of species worldwide. We identified genes expressed in the intertidal seagrass Zostera noltii during recovery from a simulated low tide heat-shock exposure. Five Expressed Sequence Tag (EST) libraries were compared, corresponding to four recovery times following sub-lethal temperature stress, and a non-stressed control. We sequenced and analyzed 7009 sequence reads from 30min, 2h, 4h and 24h after the beginning of the heat-shock (AHS), and 1585 from the control library, for a total of 8594 sequence reads. Among 51 Tentative UniGenes (TUGs) exhibiting significantly different expression between libraries, 19 (37.3%) were identified as 'molecular chaperones' and were over-expressed following heat-shock, while 12 (23.5%) were 'photosynthesis TUGs' generally under-expressed in heat-shocked plants. A time course analysis of expression showed a rapid increase in expression of the molecular chaperone class, most of which were heat-shock proteins; which increased from 2 sequence reads in the control library to almost 230 in the 30min AHS library, followed by a slow decrease during further recovery. In contrast, 'photosynthesis TUGs' were under-expressed 30min AHS compared with the control library, and declined progressively with recovery time in the stress libraries, with a total of 29 sequence reads 24h AHS, compared with 125 in the control. A total of 4734 TUGs were screened for EST-Single Sequence Repeats (EST-SSRs) and 86 microsatellites were identified. Copyright © 2011 Elsevier B.V. All rights reserved.

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.

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.

RAMBO-K: Rapid and Sensitive Removal of Background Sequences from Next Generation Sequencing Data.

Directory of Open Access Journals (Sweden)

Simon H Tausch

Full Text Available The assembly of viral or endosymbiont genomes from Next Generation Sequencing (NGS data is often hampered by the predominant abundance of reads originating from the host organism. These reads increase the memory and CPU time usage of the assembler and can lead to misassemblies.We developed RAMBO-K (Read Assignment Method Based On K-mers, a tool which allows rapid and sensitive removal of unwanted host sequences from NGS datasets. Reaching a speed of 10 Megabases/s on 4 CPU cores and a standard hard drive, RAMBO-K is faster than any tool we tested, while showing a consistently high sensitivity and specificity across different datasets.RAMBO-K rapidly and reliably separates reads from different species without data preprocessing. It is suitable as a straightforward standard solution for workflows dealing with mixed datasets. Binaries and source code (java and python are available from http://sourceforge.net/projects/rambok/.

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

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.

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)

Analysis of the Integral Response of CAREM Reactor and the Residual Heat Removal System During a Failure of the Steam Generators Feed Water System

International Nuclear Information System (INIS)

Gimenez, Marcelo; Zanocco, Pablo; Schlamp, Miguel

2000-01-01

A global analysis of the behavior of Carem-25 Reactor and Residual Heat Removal System (RHRS) to mitigate a loss of heat sink accident is done in the present work.The proposed RHRS removes 2 MW of power and is duplicated to fulfill the redundancy criteria.It consists of two condensers with two tubes in a parallel array.Each tube has 2 S CH 160 TP 347 SS and 2 m 2 of area.The RHRS design requierements (for this accidental sequence) are: Short-term: primary circuit pressure must remain below the safety valves opening set point and the condensers must not flood in order to avoid instabilities. Long-term: reach hot-shutdown condition (primary circuit pressure below 2.3 MPa) at least before 48 hrs. Short-term reactor behavior is simulated using RELAP5 with a detail nodalization of the primary circuit and RHRS.Long term performance is simulated with a simple and conservative model, assuming a saturated primary circuit. This condition is expected during RHRS operation

Reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

International Nuclear Information System (INIS)

Schilling, R.

1983-01-01

Requirements made by Federal German licensing authorities called for the analysis of the secondary-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems identified, remedial measures of a system-specific and test-strategic nature presented and their contribution to improving system availability quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

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

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.

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 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)

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

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

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.

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

Probabilistic reliability analyses to detect weak points in secondary-side residual heat removal systems of KWU PWR plants

International Nuclear Information System (INIS)

Schilling, R.

1984-01-01

Requirements made by Federal German licensing authorities called for the analysis of the second-side residual heat removal systems of new PWR plants with regard to availability, possible weak points and the balanced nature of the overall system for different incident sequences. Following a description of the generic concept and the process and safety-related systems for steam generator feed and main steam discharge, the reliability of the latter is analyzed for the small break LOCA and emergency power mode incidents, weak points in the process systems are identified, remedial measures of a system-specific and test-strategic nature are presented and their contribution to improving system availability is quantified. A comparison with the results of the German Risk Study on Nuclear Power Plants (GRS) shows a distinct reduction in core meltdown frequency. (orig.)

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

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)

Isolation, screening and molecular identification of novel bacterial strain removing methylene blue from water solutions

Science.gov (United States)

Kilany, Mona

2017-11-01

The potentially deleterious effects of methylene blue (MB) on human health drove the interest in its removal promptly. Bioremediation is an effective and eco friendly for removing MB. Soil bacteria were isolated and examined for their potential to remove MB. The most potent bacterial candidate was characterized and identified using 16S rRNA sequence technique. The evolutionary history of the isolate was conducted by maximum likelihood method. Some physiochemical parameters were optimized for maximum decolorization. Decolorization mechanism and microbial toxicity study of MB (100 mg/l) and by-products were investigated. Participation of heat killed bacteria in color adsorption have been investigated too. The bacterial isolate was identified as Stenotrophomonas maltophilia strain Kilany_MB 16S ribosomal RNA gene with 99% sequence similarity. The sequence was submitted to NCBI (Accession number = KU533726). Phylogeny depicted the phylogenetic relationships between 16S ribosomal RNA gene, partial sequence (1442 bp), of the isolated strain and other strains related to Stenotrophomonas maltophilia in the GenBank database. The optimal conditions were investigated to be pH 5 at 30 °C, after 24 h using 5 mg/l MB showing optimum decolorization percentage (61.3%). Microbial toxicity study demonstrated relative reduction in the toxicity of MB decolorized products on test bacteria. Mechanism of color removal was proved by both biosorption and biodegradation, where heat-killed and live cells showed 43 and 52% of decolorization, respectively, as a maximum value after 24-h incubation. It was demonstrated that the mechanism of color removal is by adsorption. Therefore, good performance of S maltophilia in MB color removal reinforces the exploitation of these bacteria in environmental clean-up and restoration of the ecosystem.

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)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Heat 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)

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

2014-01-01

the efficiency of machinery systems. The wet sulphuric acid process is an effective way of removing flue gas sulphur oxides from land-based coal-fired power plants. Moreover, organic Rankine cycles (ORC) are suitable for heat to power conversion for low temperature heat sources. This paper describes the design...... that an ORC 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 the combined cycle thermal efficiency by 2.6%. The findings indicate that the technology has potential in marine applications regarding both energy...... and modeling of a highly efficient machinery system which includes the removal of exhaust gas sulphur oxides. The system consists of a two-stroke diesel engine, the wet sulphuric process for sulphur removal, a conventional steam Rankine cycle and an ORC. Results of numerical modeling efforts suggest...

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

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.

Analysis of loss of decay heat removal sequences at Browns Ferry Unit One: Chapter 17

International Nuclear Information System (INIS)

Harrington, R.M.

1983-01-01

This paper summarizes the Oak Ridge National Laboratory (ORNL) report ''Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis'' (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA), using Unit One of the Browns Ferry Nuclear Plant as the model design. Each unit of this three-unit plant has a maximum authorized power of 3293 MW(t) or 1067 net MW(e). The primary containments are of the Mark I pressure suppression pool type and the three units share a secondary containment of the controlled leakage, elevated release design. Each unit occupies a separate reactor building located in one structure underneath the common refueling floor

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

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.

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

Efficiency of SBR Process with a Six Sequence Aerobic-Anaerobic Cycle for Phosphorus and Organic Material Removal from Municipal Wastewater

Directory of Open Access Journals (Sweden)

Nadiya Shahandeh

2018-02-01

Full Text Available Background: Various chemical, physical and biologic treatment methods are being used to remove nitrogen and phosphorus from wastewater. Sequencing batch reactor (SBR is a modified activated sludge process that removes phosphorus and organic material from sanitary wastewater, biologically. Methods: This study was conducted in 2016.The performance of an aerobic-anaerobic SBR pilot device, located at Ahwaz West Wastewater Treatment Plant, Ahwaz, southern Iran in phosphorus and organic material removal was evaluated to determine the effect of the aerobic-anaerobic step time on the efficiency of nitrogen and phosphorus removal, the effect of changing the sequence of steps and the effect of time ratio on phosphorus removal efficiency. A reactor of 8 L was used. Influent contained 397 and 10.7 mg/l COD and phosphorus, respectively. The pilot plant started with a 24 h cycle including four cycles of 6 h, as follows: 1- Loading (15 min, 2-Anaerobic (2 h-Aerobic (2 h, 3- Settling (1 h, Idleness (30 min and 5- decant (15 min. Results: After reaching steady conditions (6 months, Removal percentages of phosphorus, BOD5, COD, and TSS in The SBR over a period of 6 months was 79%, 86%, 89% and 83%, respectively. Conclusion: Result of this study can be used for designing and optimum operation of sequencing batch reactors.

Simultaneous nitrification-denitrification and phosphorus removal in a fixed bed sequencing batch reactor (FBSBR)

International Nuclear Information System (INIS)

Rahimi, Yousef; Torabian, Ali; Mehrdadi, Naser; Shahmoradi, Behzad

2011-01-01

Research highlights: → Sludge production in FSBR reactor is 20-30% less than SBR reactor. → FSBR reactor showed more nutrient removal rate than SBR reactor. → FSBR reactor showed less VSS/TSS ratio than SBR reactor. - Abstract: Biological nutrient removal (BNR) was investigated in a fixed bed sequencing batch reactor (FBSBR) in which instead of activated sludge polypropylene carriers were used. The FBSBR performance on carbon and nitrogen removal at different loading rates was significant. COD, TN, and phosphorus removal efficiencies were at range of 90-96%, 60-88%, and 76-90% respectively while these values at SBR reactor were 85-95%, 38-60%, and 20-79% respectively. These results show that the simultaneous nitrification-denitrification (SND) is significantly higher than conventional SBR reactor. The higher total phosphorus (TP) removal in FBSBR correlates with oxygen gradient in biofilm layer. The influence of fixed media on biomass production yield was assessed by monitoring the MLSS concentrations versus COD removal for both reactors and results revealed that the sludge production yield (Y obs ) is significantly less in FBSBR reactors compared with SBR reactor. The FBSBR was more efficient in SND and phosphorus removal. Moreover, it produced less excess sludge but higher in nutrient content and stabilization ratio (less VSS/TSS ratio).

Simultaneous nitrification-denitrification and phosphorus removal in a fixed bed sequencing batch reactor (FBSBR)

Energy Technology Data Exchange (ETDEWEB)

Rahimi, Yousef, E-mail: you.rahimi@gmail.com [Department of Civil and Environmental Engineering, Graduate Faculty of Environment, University of Tehran, No. 25 Qods St., Enghelab Ave, Tehran (Iran, Islamic Republic of); Torabian, Ali, E-mail: atorabi@ut.ac.ir [Department of Civil and Environmental Engineering, Graduate Faculty of Environment, University of Tehran, No. 25 Qods St., Enghelab Ave, Tehran (Iran, Islamic Republic of); Mehrdadi, Naser, E-mail: mehrdadi@ut.ac.ir [Department of Civil and Environmental Engineering, Graduate Faculty of Environment, University of Tehran, No. 25 Qods St., Enghelab Ave, Tehran (Iran, Islamic Republic of); Shahmoradi, Behzad, E-mail: bshahmorady@gmail.com [Department of Environmental Science, University of Mysore, MGM-06 Mysore (India)

2011-01-30

Research highlights: {yields} Sludge production in FSBR reactor is 20-30% less than SBR reactor. {yields} FSBR reactor showed more nutrient removal rate than SBR reactor. {yields} FSBR reactor showed less VSS/TSS ratio than SBR reactor. - Abstract: Biological nutrient removal (BNR) was investigated in a fixed bed sequencing batch reactor (FBSBR) in which instead of activated sludge polypropylene carriers were used. The FBSBR performance on carbon and nitrogen removal at different loading rates was significant. COD, TN, and phosphorus removal efficiencies were at range of 90-96%, 60-88%, and 76-90% respectively while these values at SBR reactor were 85-95%, 38-60%, and 20-79% respectively. These results show that the simultaneous nitrification-denitrification (SND) is significantly higher than conventional SBR reactor. The higher total phosphorus (TP) removal in FBSBR correlates with oxygen gradient in biofilm layer. The influence of fixed media on biomass production yield was assessed by monitoring the MLSS concentrations versus COD removal for both reactors and results revealed that the sludge production yield (Y{sub obs}) is significantly less in FBSBR reactors compared with SBR reactor. The FBSBR was more efficient in SND and phosphorus removal. Moreover, it produced less excess sludge but higher in nutrient content and stabilization ratio (less VSS/TSS ratio).

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.

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.

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)

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.)

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)

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.

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

2017-03-15

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

International Nuclear Information System (INIS)

Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

2017-01-01

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

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

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

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.

Viral Bacterial Artificial Chromosomes: Generation, Mutagenesis, and Removal of Mini-F Sequences

Directory of Open Access Journals (Sweden)

B. Karsten Tischer

2012-01-01

Full Text Available Maintenance and manipulation of large DNA and RNA virus genomes had presented an obstacle for virological research. BAC vectors provided a solution to both problems as they can harbor large DNA sequences and can efficiently be modified using well-established mutagenesis techniques in Escherichia coli. Numerous DNA virus genomes of herpesvirus and pox virus were cloned into mini-F vectors. In addition, several reverse genetic systems for RNA viruses such as members of Coronaviridae and Flaviviridae could be established based on BAC constructs. Transfection into susceptible eukaryotic cells of virus DNA cloned as a BAC allows reconstitution of recombinant viruses. In this paper, we provide an overview on the strategies that can be used for the generation of virus BAC vectors and also on systems that are currently available for various virus species. Furthermore, we address common mutagenesis techniques that allow modification of BACs from single-nucleotide substitutions to deletion of viral genes or insertion of foreign sequences. Finally, we review the reconstitution of viruses from BAC vectors and the removal of the bacterial sequences from the virus genome during this process.

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)

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)

CNE (Embalse nuclear power plant): probabilistic safety study. Electric power supply. Events sequence

International Nuclear Information System (INIS)

Figueroa, N.

1987-01-01

The plant response to the occurrence of the starting event 'total loss of electric power supply to class IV and class III' is analyzed. This involves the study of automatical actions of safety and process systems as well as the operator actions. The probabilistic evaluation of starting event frequency is performed through fault-tree techniques. The frequency of occurrence 'loss of electric power supply to class IV (λIV = 0.56/year) and the probability of failure to demand of 'reserve' generating groups (Pd III 6.79 x 10 -3 ) contribute to the mentioned frequency. As soon as the starting event occurs, the reactor power must be reduced to 0%, the fuel must be cooled through the thermo siphon and decay heat has to be removed. The events sequence analysis leads to the conclusion that the non shutting down of the reactor with any of the shutdown systems is 'incredible' (10 -6 /year). In all cases the fuel is cooled by building the thermo siphon except when a substantial inventory loss exist due to a closure failure of some valve of pressure and inventory control system. The order of magnitude of the failure of decay heat removal through the steam generators is 4 x 10 -4 . This removal would be assured by the emergency water system. Therefore, the frequency of the sequence of possible core meltdown, when the reactor does not shut down is: λ = 5 x 10 -9 /year and for the failure of heat removal: λ = 2 x 10 -6 /year. (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.

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.

Event sequence quantification for a loss of shutdown cooling accident in the GCFR

International Nuclear Information System (INIS)

Frank, M.; Reilly, J.

1979-10-01

A summary is presented of the core-wide sequence of events of a postulated total loss of forced and natural convection decay heat removal in a shutdown Gas-Cooled Fast Reactor (GCFR). It outlines the analytical methods and results for the progression of the accident sequence. This hypothetical accident proceeds in the distinct phases of cladding melting, assembly wall melting and molten steel relocation into the interassembly spacing, and fuel relocation. It identifies the key phenomena of the event sequence and the concerns and mechanisms of both recriticality and recriticality prevention

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)

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

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...

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

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)

New Aspects of a Lid-Removal Mechanism in the Onset of a SEP-Producing Eruption Sequence

Science.gov (United States)

Sterling, Alphonse C.; Moore, Ronald L.; Falconer, David; Knox, Javon M

2014-06-01

We examine a sequence of two ejective eruptions from a single active region on 2012 January 23, using magnetograms and EUV images from SDO/HMI and SDO/AIA, and EUV images from STEREO. Cheng et al. (2013) showed that the first eruption's (``Eruption 1'') flux rope was apparent only in ``hotter'' AIA channels, and that it removed overlying field that allowed the second eruption (``Eruption 2'') to begin via ideal MHD instability; here we say Eruption 2 began via a ``lid removal'' mechanism. We show that during Eruption-1's onset, its flux rope underwent ``tether weakening'' (TW) reconnection with the field of an adjacent active region. Standard flare loops from Eruption 1 developed over Eruption-2's flux rope and enclosed filament, but these overarching new loops were unable to confine that flux rope/filament. Eruption-1's flare loops, from both TW reconnection and standard-flare-model internal reconnection, were much cooler than Eruption-2's flare loops (GOES thermal temperatures of ~9 MK compared to ~14 MK). This eruption sequence produced a strong solar energetic particle (SEP) event (10 MeV protons, >10^3 pfu for 43 hrs), apparently starting when Eruption-2's CME blasted through Eruption-1's CME at 5---10 R_s. This occurred because the two CMEs originated in close proximity and in close time sequence: Eruption-1's fast rise started soon after the TW reconnection; the lid removal by Eruption-1's ejection triggered the slow onset of Eruption 2; and Eruption-2's CME, which started ~1 hr later, was three times faster than Eruption-1's CME.

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)

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

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)

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

Passive decay heat removal by natural air convection after severe accidents

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Overview of BWR Severe Accident Sequence Analyses at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Hodge, S.A.

1983-01-01

Since its inception in October 1980, the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory (ORNL) has completed four studies including Station Blackout, Scram Discharge Volume Break, Loss of Decay Heat Removal, and Loss of Injection accident sequences for the Browns Ferry Nuclear Plant. The accident analyses incorporated in a SASA study provide much greater detail than that practically achievable in a Probabilistic Risk Assessment (PRA). When applied to the candidate dominant accident sequences identified by a PRA, the detailed SASA results determine if factors neglected by the PRA would have a significant effect on the order of dominant sequences. Ongoing SASA work at ORNL involves the analysis of Anticipated Transients Without Scram (ATWS) sequences for Browns Ferry

Time evolution simulation of heat removal in a small water tank by natural convection

Energy Technology Data Exchange (ETDEWEB)

Freitas, Carlos Alberto de, E-mail: carlos.freitas1950@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), Nilopolis, RJ (Brazil); Jachic, Joao; Moreira, Maria de Lourdes, E-mail: jjachic@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2013-07-01

One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

Time evolution simulation of heat removal in a small water tank by natural convection

International Nuclear Information System (INIS)

Freitas, Carlos Alberto de; Jachic, Joao; Moreira, Maria de Lourdes

2013-01-01

One of the cooling modes for any source of heat such as in a shutdown nuclear core is the natural convection. The design specifications of any cooling pool can only be done when the removal heat rate and the corresponding mass flow rate is reasonably established. In our simulation scheme, we assumed that the body forces acting in the cubic water cell are: the weight, the drag force and the integrated pressure forces on the horizontal surfaces, the viscosity shear forces on the vertical surfaces and also a special viscosity drag force due to the mass dislocation along a Bernoulli type current tube outside the motive region. For a suitable time step, the uprising convection velocity is determined by an implicit and also by an explicit solution algorithm. The resulting differential equation depends on updating specific mass, dynamic viscosity and constant pressure heat coefficient with the last known temperature in the cell that absorbed heat. Numerical calculation software was performed using MATLAB’s technical computing language and then applied for a heat generation plate simulating a spent fuel assembler from a shutdown nuclear core. The results show time evolution of convection, terminal velocity and water temperature distribution. Pool dimension as well as pool level decrement are also determined for various air exhausting system conditions and heat rate of the spent fuel plate being cooled. (author)

Biological nutrient removal from municipal wastewater in sequencing batch biofilm reactors

Energy Technology Data Exchange (ETDEWEB)

Arnz, P

2001-07-01

Enhanced biological phosphorus removal (EBPR) has only been put into practice in activated sludge systems. In recent years, the Sequencing Batch Biofilm Reactor (SBBR) has emerged as an alternative allowing EBPR to be achieved in a biofilm reactor. High efficiency of phosphate removal was demonstrated in a SBBR fed with synthetic wastewater containing acetate. The aim of this study was to investigate EBPR from municipal wastewater in semi full-scale and laboratory-scale SBBRs. The focus of the investigation in the semi full-scale reactor was on determination of achievable reaction rates and effluent concentrations under varying influent conditions throughout all seasons of a year. Interactions between nitrogen and phosphorus removal and the influence of backwashing on the reactor performance was examined. Summing up, it can be stated that the SBBR proved to be an attractive alternative to activated sludge systems. Phosphorus elimination efficiency was comparable to common systems but biomass sedimentation problems were avoided. In order to further exploit the potential of the SBBR and to achieve reactor performances superior to those of existing systems designing a special biofilm carrier material may allow to increase the phenomenon of simultaneous nitrification/denitrification while maintaining EBPR activity. (orig.) [German] Die vermehrte biologische Phosphorelimination (Bio-P) aus Abwasser wurde bisher nur in Belebtschlammsystemen praktiziert. In den letzten Jahren konnte jedoch gezeigt werden, dass sich durch die Anwendung des Sequencing Batch Biofilm Reactor (SBBR) - Verfahrens auch in Biofilmreaktoren Bio-P verwirklichen laesst. Versuche in Laboranlagen haben ergeben, dass sich eine weitgehende Phosphorelimination aufrecht erhalten laesst, wenn die Reaktoren mit einem ideal zusammengesetzten, synthetischen Abwasser beschickt werden. Ziel dieser Arbeit war es, Bio-P aus kommunalem Abwasser in SBBR-Versuchsanlagen im halbtechnischen und im Labormassstab zu

Mathematical modelling for magnetite (crude removal from primary heat transfer loop by ion-exchange resins

Directory of Open Access Journals (Sweden)

Zeeshan Nawaz

2009-04-01

Full Text Available The present research focuses to develop mathematical model for the removal of iron (magnetite by ion-exchange resin from primary heat transfer loop of process industries. This mathematical model is based on operating capacities (that’s provide more effective design as compared to loading capacity from static laboratory tests. Results showed non-steady state distribution of external Fe2+ and limitations imposed on operating conditions, these conditions includes; loading and elution cycle time, flow rate, concentration of both loading and removal, volume of resin required. Number of generalized assumptions was made under shortcut modeling techniques to overcome the gap of theoretical and actual process design.

Identification and Removal of Contaminant Sequences From Ribosomal Gene Databases: Lessons From the Census of Deep Life.

Science.gov (United States)

Sheik, Cody S; Reese, Brandi Kiel; Twing, Katrina I; Sylvan, Jason B; Grim, Sharon L; Schrenk, Matthew O; Sogin, Mitchell L; Colwell, Frederick S

2018-01-01

Earth's subsurface environment is one of the largest, yet least studied, biomes on Earth, and many questions remain regarding what microorganisms are indigenous to the subsurface. Through the activity of the Census of Deep Life (CoDL) and the Deep Carbon Observatory, an open access 16S ribosomal RNA gene sequence database from diverse subsurface environments has been compiled. However, due to low quantities of biomass in the deep subsurface, the potential for incorporation of contaminants from reagents used during sample collection, processing, and/or sequencing is high. Thus, to understand the ecology of subsurface microorganisms (i.e., the distribution, richness, or survival), it is necessary to minimize, identify, and remove contaminant sequences that will skew the relative abundances of all taxa in the sample. In this meta-analysis, we identify putative contaminants associated with the CoDL dataset, recommend best practices for removing contaminants from samples, and propose a series of best practices for subsurface microbiology sampling. The most abundant putative contaminant genera observed, independent of evenness across samples, were Propionibacterium , Aquabacterium , Ralstonia , and Acinetobacter . While the top five most frequently observed genera were Pseudomonas , Propionibacterium , Acinetobacter , Ralstonia , and Sphingomonas . The majority of the most frequently observed genera (high evenness) were associated with reagent or potential human contamination. Additionally, in DNA extraction blanks, we observed potential archaeal contaminants, including methanogens, which have not been discussed in previous contamination studies. Such contaminants would directly affect the interpretation of subsurface molecular studies, as methanogenesis is an important subsurface biogeochemical process. Utilizing previously identified contaminant genera, we found that ∼27% of the total dataset were identified as contaminant sequences that likely originate from DNA

Summary report for Group X6: Heat removal system and system analysis

Energy Technology Data Exchange (ETDEWEB)

Leung, W

2005-12-15

This report is a summary of the activities of the X6 design support for the Heat Removal System (HRS) of MEGAPIE. It can be divided into two main parts: The first part is about the design and manufacturing of he cooling loop (the first 3 chapters), and the second part is dealing with the thermal hydraulic analysis of the overall HRS. This also reflects the change of the X6 activities from design to operation support. The activities of this group are more or less driven by the needs rather than a complete set of tasks given at the start of the project. The first part chronicles the system development. Some of the arguments are probably outdated but are kept in the original form to illustrate the evolution of concepts. The main objective is, of course, to design a heat removal system that can cool the liquid metal spallation target for a 1 MW proton beam i.e. 1.74 mA in 575 MeV). It is also reckoned that the liquid metal, BE (lead-bismuth-eutectic), must be kept liquid even when the proton beam was switched off. This requires either that the cooling system can be shut down or the operating temperature of the coolant be higher than the freezing point of LBE. As for safety concerns, the HRS system must not exert a pressure that exceeds the design pressure of the target beam window in case of a break at the target heat exchanger (THX); this limits the cover gas pressure to about 4 bar(a). These are the basic design principles that carry through the conceptual and engineering design of he system. The organic coolant Diphyl THT was then chosen, because of its wide range of operating temperature (i.e. from 0 to 340 degC) and high boiling point, and a proven record in industrial applications. (author)

Summary report for Group X6: Heat removal system and system analysis

International Nuclear Information System (INIS)

Leung, W.

2005-12-01

This report is a summary of the activities of the X6 design support for the Heat Removal System (HRS) of MEGAPIE. It can be divided into two main parts: The first part is about the design and manufacturing of he cooling loop (the first 3 chapters), and the second part is dealing with the thermal hydraulic analysis of the overall HRS. This also reflects the change of the X6 activities from design to operation support. The activities of this group are more or less driven by the needs rather than a complete set of tasks given at the start of the project. The first part chronicles the system development. Some of the arguments are probably outdated but are kept in the original form to illustrate the evolution of concepts. The main objective is, of course, to design a heat removal system that can cool the liquid metal spallation target for a 1 MW proton beam i.e. 1.74 mA in 575 MeV). It is also reckoned that the liquid metal, BE (lead-bismuth-eutectic), must be kept liquid even when the proton beam was switched off. This requires either that the cooling system can be shut down or the operating temperature of the coolant be higher than the freezing point of LBE. As for safety concerns, the HRS system must not exert a pressure that exceeds the design pressure of the target beam window in case of a break at the target heat exchanger (THX); this limits the cover gas pressure to about 4 bar(a). These are the basic design principles that carry through the conceptual and engineering design of he system. The organic coolant Diphyl THT was then chosen, because of its wide range of operating temperature (i.e. from 0 to 340 degC) and high boiling point, and a proven record in industrial applications. (author)

Expressed sequence tags from heat-shocked seagrass Zostera noltii (Hornemann) from its southern distribution range

NARCIS (Netherlands)

Massa, Sonia I.; Pearson, Gareth A.; Aires, Tania; Kube, Michael; Olsen, Jeanine L.; Reinhardt, Richard; Serrao, Ester A.; Arnaud-Haond, Sophie

Predicted global climate change threatens the distributional ranges of species worldwide. We identified genes expressed in the intertidal seagrass Zostera midi during recovery from a simulated low tide heat-shock exposure. Five Expressed Sequence Tag (EST) libraries were compared, corresponding to

Non-linear effects in vortex viscous flow in superconductors-role of finite heat removal velocity

International Nuclear Information System (INIS)

Bezuglyj, A.I.; Shklovskij, V.A.

1991-01-01

The role of finite heat removal velocity in experiments on non-linear effects in vortex viscous flow in superconducting films near critical temperature was investigated. It was shown that the account of thermal effects permits to explain the experimentally observed dependence of electron energy relaxation time and current break-down in voltage-current characteristic from magnetic field value. 5 refs.; 1 fig. (author)

Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

International Nuclear Information System (INIS)

Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

1988-07-01

This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

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

International Nuclear Information System (INIS)

Natta, M.; Bloch, M.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

The mechanism and design of sequencing batch reactor systems for nutrient removal--the state of the art.

Science.gov (United States)

Artan, N; Wilderer, P; Orhon, D; Morgenroth, E; Ozgür, N

2001-01-01

The Sequencing Batch Reactor (SBR) process for carbon and nutrient removal is subject to extensive research, and it is finding a wider application in full-scale installations. Despite the growing popularity, however, a widely accepted approach to process analysis and modeling, a unified design basis, and even a common terminology are still lacking; this situation is now regarded as the major obstacle hindering broader practical application of the SBR. In this paper a rational dimensioning approach is proposed for nutrient removal SBRs based on scientific information on process stoichiometry and modelling, also emphasizing practical constraints in design and operation.

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-08-15

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

Optimization of residual heat removal pump axial thrust and axial bearing

International Nuclear Information System (INIS)

Schubert, F.

1996-01-01

The residual heat removal (RHR) pumps of German 1300 megawatt pressurized-water reactor (PWR) power plants are of the single stage end suction type with volute casing or with diffuser and forged circular casing. Due to the service conditions the pumps have to cover the full capacity range as well as a big variation in suction static pressure. This results in a big difference in the axial thrust that has to be borne by the axial bearing. Because these pumps are designed to operate without auxiliary systems (things that do not exist can not fail), they are equipped with antifriction bearings and sump oil lubrication. To minimize the heat production within the bearing casing, a number of PWR plants have pumps with combined axial/radial bearings of the ball type. Due to the fact that the maximum axial thrust caused by static pressure and hydrodynamic forces on the impeller is too big to be borne by that type of axial bearing, the impellers were designed to produce a hydrodynamic axial force that counteracts the static axial force. Thus, the resulting axial thrust may change direction when the static pressure varies

Optimization of residual heat removal pump axial thrust and axial bearing

Energy Technology Data Exchange (ETDEWEB)

Schubert, F.

1996-12-01

The residual heat removal (RHR) pumps of German 1300 megawatt pressurized-water reactor (PWR) power plants are of the single stage end suction type with volute casing or with diffuser and forged circular casing. Due to the service conditions the pumps have to cover the full capacity range as well as a big variation in suction static pressure. This results in a big difference in the axial thrust that has to be borne by the axial bearing. Because these pumps are designed to operate without auxiliary systems (things that do not exist can not fail), they are equipped with antifriction bearings and sump oil lubrication. To minimize the heat production within the bearing casing, a number of PWR plants have pumps with combined axial/radial bearings of the ball type. Due to the fact that the maximum axial thrust caused by static pressure and hydrodynamic forces on the impeller is too big to be borne by that type of axial bearing, the impellers were designed to produce a hydrodynamic axial force that counteracts the static axial force. Thus, the resulting axial thrust may change direction when the static pressure varies.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

In situ detection of a heat-shock regulatory element binding protein using a soluble short synthetic enhancer sequence

Energy Technology Data Exchange (ETDEWEB)

Harel-Bellan, A; Brini, A T; Farrar, W L [National Cancer Institute, Frederick, MD (USA); Ferris, D K [Program Resources, Inc., Frederick, MD (USA); Robin, P [Institut Gustave Roussy, Villejuif (France)

1989-06-12

In various studies, enhancer binding proteins have been successfully absorbed out by competing sequences inserted into plasmids, resulting in the inhibition of the plasmid expression. Theoretically, such a result could be achieved using synthetic enhancer sequences not inserted into plasmids. In this study, a double stranded DNA sequence corresponding to the human heat shock regulatory element was chemically synthesized. By in vitro retardation assays, the synthetic sequence was shown to bind specifically a protein in extracts from the human T cell line Jurkat. When the synthetic enhancer was electroporated into Jurkat cells, not only the enhancer was shown to remain undegraded into the cells for up to 2 days, but also its was shown to bind intracellularly a protein. The binding was specific and was modulated upon heat shock. Furthermore, the binding protein was shown to be of the expected molecular weight by UV crosslinking. However, when the synthetic enhancer element was co-electroporated with an HSP 70-CAT reporter construct, the expression of the reporter plasmid was consistently enhanced in the presence of the exogenous synthetic enhancer.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Passive decay heat removal by sump cooling after core meltdown

International Nuclear Information System (INIS)

Knebel, J.U.; Mueller, U.

1996-01-01

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

Features of an emergency heat-conducting path in reactors about lead-bismuth and lead heat-carriers

International Nuclear Information System (INIS)

Beznosov, A.V.; Bokova, T.A.; Molodtsov, A.A.

2006-01-01

The reactor emergency heat removal systems should transfer heat from the surface of reactor core fuel element claddings to the primary circuit followed by heat transfer to the environment. One suggests three design approaches for emergency heat removal systems in lead-bismuth and lead cooled reactor circuits that take account of the peculiar nature of their features. Application of the discussed systems for emergency heat removal improves safety of lead-bismuth and lead cooled reactor plants [ru

Loss of residual heat removal system: Diablo Canyon, Unit 2, April 10, 1987

International Nuclear Information System (INIS)

1987-06-01

This report presents the findings of an NRC Augmented Inspection Team (AIT) investigation into the circumstances associated with the loss of residual heat removal (RHR) system capability for a period of approximately one and one-half hours at the Diablo Canyon, Unit 2 reactor facility on April 10, 1987. This event occurred while the Diablo Canyon, Unit 2, a pressurized water reactor, was shutdown with the reactor coolant system (RCS) water level drained to approximately mid-level of the hot leg piping. The reactor containment building equipment hatch was removed at the time of the event, and plant personnel were in the process of removing the primary side manways to gain access into the steam generator channel head areas. Thus, two fission product barriers were breached throughout the event. The RCS temperature increased from approximately 87 0 F to bulk boiling conditions without RCS temperature indication available to the plant operators. The RCS was subsequently pressurized to approximately 7 to 10 psig. The NRC AIT members concluded that the Diablo Canyon, Unit 2 plant was, at the time of the event, in a condition not previously analyzed by the NRC staff. The AIT findings from this event appear significant and generic to other pressurized water reactor facilities licensed by the NRC

Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

International Nuclear Information System (INIS)

Siniscalchi, Marcio Rezende

2005-01-01

This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

Dinoflagellate phylogeny as inferred from heat shock protein 90 and ribosomal gene sequences.

Directory of Open Access Journals (Sweden)

Mona Hoppenrath

2010-10-01

Full Text Available Interrelationships among dinoflagellates in molecular phylogenies are largely unresolved, especially in the deepest branches. Ribosomal DNA (rDNA sequences provide phylogenetic signals only at the tips of the dinoflagellate tree. Two reasons for the poor resolution of deep dinoflagellate relationships using rDNA sequences are (1 most sites are relatively conserved and (2 there are different evolutionary rates among sites in different lineages. Therefore, alternative molecular markers are required to address the deeper phylogenetic relationships among dinoflagellates. Preliminary evidence indicates that the heat shock protein 90 gene (Hsp90 will provide an informative marker, mainly because this gene is relatively long and appears to have relatively uniform rates of evolution in different lineages.We more than doubled the previous dataset of Hsp90 sequences from dinoflagellates by generating additional sequences from 17 different species, representing seven different orders. In order to concatenate the Hsp90 data with rDNA sequences, we supplemented the Hsp90 sequences with three new SSU rDNA sequences and five new LSU rDNA sequences. The new Hsp90 sequences were generated, in part, from four additional heterotrophic dinoflagellates and the type species for six different genera. Molecular phylogenetic analyses resulted in a paraphyletic assemblage near the base of the dinoflagellate tree consisting of only athecate species. However, Noctiluca was never part of this assemblage and branched in a position that was nested within other lineages of dinokaryotes. The phylogenetic trees inferred from Hsp90 sequences were consistent with trees inferred from rDNA sequences in that the backbone of the dinoflagellate clade was largely unresolved.The sequence conservation in both Hsp90 and rDNA sequences and the poor resolution of the deepest nodes suggests that dinoflagellates reflect an explosive radiation in morphological diversity in their recent

Radiation detector system having heat pipe based cooling

Science.gov (United States)

Iwanczyk, Jan S.; Saveliev, Valeri D.; Barkan, Shaul

2006-10-31

A radiation detector system having a heat pipe based cooling. The radiation detector system includes a radiation detector thermally coupled to a thermo electric cooler (TEC). The TEC cools down the radiation detector, whereby heat is generated by the TEC. A heat removal device dissipates the heat generated by the TEC to surrounding environment. A heat pipe has a first end thermally coupled to the TEC to receive the heat generated by the TEC, and a second end thermally coupled to the heat removal device. The heat pipe transfers the heat generated by the TEC from the first end to the second end to be removed by the heat removal device.

Steady-state heat and particle removal with the actively cooled Phase III outboard pump limiter in Tore Supra

International Nuclear Information System (INIS)

Nygren, R.; Koski, J.; Lutz, T.; McGrath; Miller, J.; Watkins, J.; Guilhem, D.; Chappuis, P.; Cordier, J.; Loarer, T.

1995-01-01

Tore Supra's Phase III outboard pump limiter (OPL) is a modular actively-cooled mid-plane limiter, designed for heat and particle removal during long pulse operation. During its initial operation in 1993, the OPL successfully removed about 1 MW of power during ohmicly heated shots of up to 10 s duration and reached (steady state) thermal equilibrium. The particle pumping of the Phase III OPL was found to be about 50% greater than the Phase II OPL which had a radial distance between the last closed flux surface and the entrance of the pumping throat of 3.5 cm compared with only 2.5 cm for the Phase III OPL. This paper gives examples of power distribution over the limiter from IR measurements of surface temperature and from extensively calorimetry (34 thermocouples and 10 flow meters) and compares the distributions with values predicted by a 3D model (HF3D) with a detailed magnetic configuration (e.g., includes field ripple). ((orig.))

Modelling of decay heat removal using large water pools

International Nuclear Information System (INIS)

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

1992-01-01

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

Removal of veterinary antibiotics from anaerobically digested swine wastewater using an intermittently aerated sequencing batch reactor.

Science.gov (United States)

Zheng, Wei; Zhang, Zhenya; Liu, Rui; Lei, Zhongfang

2018-03-01

A lab-scale intermittently aerated sequencing batch reactor (IASBR) was applied to treat anaerobically digested swine wastewater (ADSW) to explore the removal characteristics of veterinary antibiotics. The removal rates of 11 veterinary antibiotics in the reactor were investigated under different chemical organic demand (COD) volumetric loadings, solid retention times (SRT) and ratios of COD to total nitrogen (TN) or COD/TN. Both sludge sorption and biodegradation were found to be the major contributors to the removal of veterinary antibiotics. Mass balance analysis revealed that greater than 60% of antibiotics in the influent were biodegraded in the IASBR, whereas averagely 24% were adsorbed by sludge under the condition that sludge sorption gradually reached its equilibrium. Results showed that the removal of antibiotics was greatly influenced by chemical oxygen demand (COD) volumetric loadings, which could achieve up to 85.1%±1.4% at 0.17±0.041kgCOD/m -3 /day, while dropped to 75.9%±1.3% and 49.3%±12.1% when COD volumetric loading increased to 0.65±0.032 and 1.07±0.073kgCOD/m -3 /day, respectively. Tetracyclines, the dominant antibiotics in ADSW, were removed by 87.9% in total at the lowest COD loading, of which 30.4% were contributed by sludge sorption and 57.5% by biodegradation, respectively. In contrast, sulfonamides were removed about 96.2%, almost by biodegradation. Long SRT seemed to have little obvious impact on antibiotics removal, while a shorter SRT of 30-40day could reduce the accumulated amount of antibiotics and the balanced antibiotics sorption capacity of sludge. Influent COD/TN ratio was found not a key impact factor for veterinary antibiotics removal in this work. Copyright © 2017. Published by Elsevier B.V.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Analytical studies on the impact of using repeated-rib roughness in LMR [Liquid Metal Reactor] decay heat removal systems

International Nuclear Information System (INIS)

Obot, N.T.; Tessier, J.H.; Pedersen, D.R.

1988-01-01

A numerical study was carried out to determine the effects of roughness on the thermal performance of Liquid Metal Reactor (LMR) decay heat removal systems for a range of possible design configurations and operating conditions. The ranges covered for relative rib height (e/D/sub h/), relative pitch (p/e) and flow attack angle were 0.026--0.103, 5--20 and 0--90 degrees, successively. The heat flux was varied between 1.1 and 21.5 kW/m 2 (0.1 and 2.0 kW/ft 2 ). Calculations were made for three cases: smooth duct with no ribs, ribs on both the guard vessel and collector wall, and ribs on the collector wall only. The results indicate that significant benefits, amounting to nearly two-fold reductions in guard vessel and collector wall temperatures, can be realized by placing repeated ribs on both the guard vessel and the collector wall. The magnitudes of the reduction in the reactor vessel temperature are considerably smaller. In general, the level of improvement, be it with respect to temperature or heat flux, is only mildly affected by changes in rib height or pitch but exhibits greater sensitivity to the assumed value for the system form loss. When the ribs are placed only on the collector wall, the heat removal capability is substantially reduced

CACHE: an extended BASIC program which computes the performance of shell and tube heat exchangers

International Nuclear Information System (INIS)

Tallackson, J.R.

1976-03-01

An extended BASIC program, CACHE, has been written to calculate steady state heat exchange rates in the core auxiliary heat exchangers, (CAHE), designed to remove afterheat from High-Temperature Gas-Cooled Reactors (HTGR). Computationally, these are unbaffled counterflow shell and tube heat exchangers. The computational method is straightforward. The exchanger is subdivided into a user-selected number of lengthwise segments; heat exchange in each segment is calculated in sequence and summed. The program takes the temperature dependencies of all thermal conductivities, viscosities and heat capacities into account providing these are expressed algebraically. CACHE is easily adapted to compute steady state heat exchange rates in any unbaffled counterflow exchanger. As now used, CACHE calculates heat removal by liquid weight from high-temperature helium and helium mixed with nitrogen, oxygen and carbon monoxide. A second program, FULTN, is described. FULTN computes the geometrical parameters required as input to CACHE. As reported herein, FULTN computes the internal dimensions of the Fulton Station CAHE. The two programs are chained to operate as one. Complete user information is supplied. The basic equations, variable lists, annotated program lists, and sample outputs with explanatory notes are included

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

International Nuclear Information System (INIS)

Kim, Kyung Mo; Bang, In Cheol

2017-01-01

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

Mechanisms of material removal and mass transport in focused ion beam nanopore formation

Energy Technology Data Exchange (ETDEWEB)

Das, Kallol, E-mail: das7@illinois.edu; Johnson, Harley T., E-mail: htj@illinois.edu [Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 1206 West Green Street, MC-244, Urbana, Illinois 61801 (United States); Freund, Jonathan B., E-mail: jbfreund@illinois.edu [Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 1206 West Green Street, MC-244, Urbana, Illinois 61801 (United States); Department of Aerospace Engineering, University of Illinois at Urbana-Champaign, 306 Talbot Laboratory, MC-236, 104 South Wright Street Urbana, Illinois 61801 (United States)

2015-02-28

Despite the widespread use of focused ion beam (FIB) processing as a material removal method for applications ranging from electron microscope sample preparation to nanopore processing for DNA sequencing, the basic material removal mechanisms of FIB processing are not well understood. We present the first complete atomistic simulation of high-flux FIB using large-scale parallel molecular dynamics (MD) simulations of nanopore fabrication in freestanding thin films. We focus on the root mechanisms of material removal and rearrangement and describe the role of explosive boiling in forming nanopores. FIB nanopore fabrication is typically understood to occur via sputter erosion. This can be shown to be the case in low flux systems, where individual ion impacts are sufficiently separated in time that they may be considered as independent events. But our detailed MD simulations show that in high flux FIB processing, above a threshold level at which thermal effects become significant, the primary mechanism of material removal changes to a significantly accelerated, thermally dominated process. Under these conditions, the target is heated by the ion beam faster than heat is conducted away by the material, leading quickly to melting, and then continued heating to nearly the material critical temperature. This leads to explosive boiling of the target material with spontaneous bubble formation and coalescence. Mass is rapidly rearranged at the atomistic scale, and material removal occurs orders of magnitude faster than would occur by simple sputtering. While the phenomenology is demonstrated computationally in silicon, it can be expected to occur at lower beam fluxes in other cases where thermal conduction is suppressed due to material properties, geometry, or ambient thermal conditions.

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Progress in methodology for probabilistic assessment of accidents: timing of accident sequences

International Nuclear Information System (INIS)

Lanore, J.M.; Villeroux, C.; Bouscatie, F.; Maigret, N.

1981-09-01

There is an important problem for probabilistic studies of accident sequences using the current event tree techniques. Indeed this method does not take into account the dependence in time of the real accident scenarios, involving the random behaviour of the systems (lack or delay in intervention, partial failures, repair, operator actions ...) and the correlated evolution of the physical parameters. A powerful method to perform the probabilistic treatment of these complex sequences (dynamic evolution of systems and associated physics) is Monte-Carlo simulation, very rare events being treated with the help of suitable weighting and biasing techniques. As a practical example the accident sequences related to the loss of the residual heat removal system in a fast breeder reactor has been treated with that method

Palm oil mill effluent and municipal wastewater co-treatment by zeolite augmented sequencing batch reactors: Turbidity removal

Science.gov (United States)

Farraji, Hossein; Zaman, Nastaein Qamaruz; Aziz, Hamidi Abdul; Sa'at, Siti Kamariah Md

2017-10-01

Palm oil mill effluent (POME) is the largest wastewater in Malaysia. Of the 60 million tons of POME produced annually, 2.4-3 million tons are total solids. Turbidity is caused by suspended solids, and 75% of total suspended solids are organic matter. Coagulation and flocculation are popular treatments for turbidity removal. Traditional commercial treatments do not meet discharge standards. This study evaluated natural zeolite and municipal wastewater (MWW)-augmented sequencing batch reactor as a microbiological digestion method for the decontamination of POME in response surface methodology. Aeration, contact time, and MWW/POME ratio were selected as response factors for turbidity removal. Results indicated that turbidity removal varied from 96.7% (MWW/POME ratio=50 %, aeration flow=0.5 L/min, and contact time=12) to 99.31% (MWW/POME ratio=80%, aeration flow 4L/min, and contact time 12 h). This study is the first to present MWW augmentation as a suitable microorganism supplier for turbidity biodegradation in high-strength agroindustrial wastewater.

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

International Nuclear Information System (INIS)

1975-10-01

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

Nuclear reactor equipped with a flooding tank and a residual heat removal and emergency cooling system

International Nuclear Information System (INIS)

Schabert, H.P.; Winkler, F.

1975-01-01

A description is given of a nuclear reactor such as a pressurized-water reactor or the like which is equipped with a flooding tank and a residual heat removal and emergency cooling system. The flooding tank is arranged within the containment shell at an elevation above the upper edge of the reactor core and contains a liquid for flooding the reactor core in the event of a loss of coolant

Heat-Assisted Machining for Material Removal Improvement

Science.gov (United States)

Mohd Hadzley, A. B.; Hafiz, S. Muhammad; Azahar, W.; Izamshah, R.; Mohd Shahir, K.; Abu, A.

2015-09-01

Heat assisted machining (HAM) is a process where an intense heat source is used to locally soften the workpiece material before machined by high speed cutting tool. In this paper, an HAM machine is developed by modification of small CNC machine with the addition of special jig to hold the heat sources in front of the machine spindle. Preliminary experiment to evaluate the capability of HAM machine to produce groove formation for slotting process was conducted. A block AISI D2 tool steel with100mm (width) × 100mm (length) × 20mm (height) size has been cut by plasma heating with different setting of arc current, feed rate and air pressure. Their effect has been analyzed based on distance of cut (DOC).Experimental results demonstrated the most significant factor that contributed to the DOC is arc current, followed by the feed rate and air pressure. HAM improves the slotting process of AISI D2 by increasing distance of cut due to initial cutting groove that formed during thermal melting and pressurized air from the heat source.

Biological phosphorus and nitrogen removal in sequencing batch reactors: effects of cycle length, dissolved oxygen concentration and influent particulate matter.

Science.gov (United States)

Ginige, Maneesha P; Kayaalp, Ahmet S; Cheng, Ka Yu; Wylie, Jason; Kaksonen, Anna H

2013-01-01

Removal of phosphorus (P) and nitrogen (N) from municipal wastewaters is required to mitigate eutrophication of receiving water bodies. While most treatment plants achieve good N removal using influent carbon (C), the use of influent C to facilitate enhanced biological phosphorus removal (EBPR) is poorly explored. A number of operational parameters can facilitate optimum use of influent C and this study investigated the effects of cycle length, dissolved oxygen (DO) concentration during aerobic period and influent solids on biological P and N removal in sequencing batch reactors (SRBs) using municipal wastewaters. Increasing cycle length from 3 to 6 h increased P removal efficiency, which was attributed to larger portion of N being removed via nitrite pathway and more biodegradable organic C becoming available for EBPR. Further increasing cycle length from 6 to 8 h decreased P removal efficiencies as the demand for biodegradable organic C for denitrification increased as a result of complete nitrification. Decreasing DO concentration in the aerobic period from 2 to 0.8 mg L(-1) increased P removal efficiency but decreased nitrification rates possibly due to oxygen limitation. Further, sedimented wastewater was proved to be a better influent stream than non-sedimented wastewater possibility due to the detrimental effect of particulate matter on biological nutrient removal.

Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

Energy Technology Data Exchange (ETDEWEB)

Hiroshi, Takahashi [Brookhaven National Lab., Upton, NY (United States)

2001-07-01

To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

International Nuclear Information System (INIS)

Hiroshi, Takahashi

2001-01-01

To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

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

International Nuclear Information System (INIS)

Schubert, B.K.

1984-07-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

An estimation of core damage frequency of a pressurized water reactor during midloop operation due to loss of residual heat removal

International Nuclear Information System (INIS)

Chao, C.C.; Chen, C.T.; Lee, M.

1995-01-01

The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 x 10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation

Removing lead from metallic mixture of waste printed circuit boards by vacuum distillation: factorial design and removal mechanism.

Science.gov (United States)

Li, Xingang; Gao, Yujie; Ding, Hui

2013-10-01

The lead removal from the metallic mixture of waste printed circuit boards by vacuum distillation was optimized using experimental design, and a mathematical model was established to elucidate the removal mechanism. The variables studied in lead evaporation consisted of the chamber pressure, heating temperature, heating time, particle size and initial mass. The low-level chamber pressure was fixed at 0.1 Pa as the operation pressure. The application of two-level factorial design generated a first-order polynomial that agreed well with the data for evaporation efficiency of lead. The heating temperature and heating time exhibited significant effects on the efficiency, which was validated by means of the copper-lead mixture experiments. The optimized operating conditions within the region studied were the chamber pressure of 0.1 Pa, heating temperature of 1023 K and heating time of 120 min. After the conditions were employed to remove lead from the metallic mixture of waste printed circuit boards, the efficiency was 99.97%. The mechanism of the effects was elucidated by mathematical modeling that deals with evaporation, mass transfer and condensation, and can be applied to a wider range of metal removal by vacuum distillation. Copyright © 2013 Elsevier Ltd. All rights reserved.

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Hu, Rui; Lisowski, Darius; Kraus, Adam

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.

Containment heat removal system

International Nuclear Information System (INIS)

Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

1992-01-01

This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

The influence of heat treatments on several types of base-metal removable partial denture alloys.

Science.gov (United States)

Morris, H F; Asgar, K; Rowe, A P; Nasjleti, C E

1979-04-01

Four removable partial denture alloys, Vitallium (Co-Cr alloy), Dentillium P.D. (Fe-Cr alloy), Durallium L.G. (Co-Cr-Ni alloy), and Ticonium 100 (Ni-Cr alloy), were evaluated in the as-cast condition and after heat treatment for 15 minutes at 1,300 degrees, 1,600 degrees, 1,900 degrees, and 2,200 degrees F followed by quenching in water. The following properties were determined and compared for each alloy at each heat treatment condition: the yield strengths at 0.01%, 0.1%, and 0.2% offsets, the ultimate tensile strength, the percent elongation, the modulus of elasticity, and the Knoop microhardness. The results were statistically analyzed. Photomicrographs were examined for each alloy and test condition. The following conclusions were made: 1. The "highest values" were exhibited by the as-cast alloy. 2. Heat treatment of the partial denture alloys tested resulted in reductions in strength, while the elongations varied. This study demonstrates that, in practice, one should avoid (a) prolonged "heat-soaking" while soldering and (b) grinding or polishing of the casting until the alloy is "red hot". 3. Durallium L.G. was the least affected by the various heat treatment conditions. 4. Conventional reporting of the yield strength at 0.2% offset, the ultimate tensile strength, and percent elongation are not adequate to completely describe and compare the mechanical behavior of alloys. The reporting of the yield strength at 0.01% offset, in addition to the other reported properties, will provide a more complete description of the behavior of the dental alloys.

The effect of different aspect ratio and bottom heat flux towards contaminant removal using numerical analysis

International Nuclear Information System (INIS)

Saadun, M N A; Manaf, M Z A; Zakaria, M S; Hafidzal, M H M; Azwadi, C S Nor; Malek, Z A A

2013-01-01

Cubic Interpolated Pseudo-particle (CIP) numerical simulation scheme has been anticipated to predict the interaction involving fluids and solid particles in an open channel with rectangular shaped cavity flow. The rectangular shaped cavity is looking by different aspect ratio in modelling the real pipeline joints that are in a range of sizes. Various inlet velocities are also being applied in predicting various fluid flow characteristics. In this paper, the constant heat flux is introduced at the bottom wall, showing the buoyancy effects towards the contaminant's removal rate. In order to characterize the fluid flow, the numerical scheme alone is initially tested and validated in a lid driven cavity with a single particle. The study of buoyancy effects and different aspect ratio of rectangular geometry were carried out using a MATLAB govern by Navier-Stokes equation. CIP is used as a model for a numerical scheme solver for fluid solid particles interaction. The result shows that the higher aspect ratio coupled with heated bottom wall give higher percentage of contaminant's removal rate. Comparing with the benchmark results has demonstrated the applicability of the method to reproduce fluid structure which is complex in the system. Despite a slight deviation of the formations of vortices from some of the literature results, the general pattern is considered to be in close agreement with those published in the literature

Post-accident heat removal ''information exchange''

International Nuclear Information System (INIS)

Plein, H.G.; Carlson, G.A.

1975-01-01

The in-core molten pool experiments are designed to produce a pool of fission heated temperature and flow patterns of such pools, and evaluate the barrier melt-through potential of the molten UO 2 . The first experiments, to be conducted this fiscal year in the Annular Core Pulse Reactor, will be uncomplicated and multiply-contained to prove containment design and to provide initial information on fission heated molten pool characteristics. Concurrent with the in-core experiments, high temperature ultrasonic techniques are being developed to measure UO 2 temperatures up to and above the melting point for use in later more definitive experiments scheduled for FY77

Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

Energy Technology Data Exchange (ETDEWEB)

Lisowski, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Kilsdonk, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bremer, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Aeschlimann, R. W. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-06-01

The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m² to accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.

Sana experiments for self-acting removal of the after-heat in reactors with pebble bed fuel and their interpretation

International Nuclear Information System (INIS)

Niessen, H.F.; Stoecker, Bernd; Amoignon, Olivier; Zuying, Gao; Jie, Liu

1997-01-01

For the confirmation of self-acting afterheat removal under hypothetical accident conditions from pebble bed reactors at the Research Center Juelich a test facility with an electrical heating input up to 30kW was erected and operated. A description of the test facility is given. Within the different tests the pebble diameter, the pebble material, the gas in the pebble bed, the heating-power and the arrangement of the heating were changed. Parts of the data were used within an IAEA Co-ordinated Research Program as benchmark problems for the code validation. All computer codes could simulate the test results with a sufficient good agreement, when the tests were executed with helium. For the tests with nitrogen the natural convection has to be taken into account. (author)

Quantitative trait loci mapping of heat tolerance in broccoli (Brassica oleracea var. italica) using genotyping-by-sequencing.

Science.gov (United States)

Branham, Sandra E; Stansell, Zachary J; Couillard, David M; Farnham, Mark W

2017-03-01

Five quantitative trait loci and one epistatic interaction were associated with heat tolerance in a doubled haploid population of broccoli evaluated in three summer field trials. Predicted rising global temperatures due to climate change have generated a demand for crops that are resistant to yield and quality losses from heat stress. Broccoli (Brassica oleracea var. italica) is a cool weather crop with high temperatures during production decreasing both head quality and yield. Breeding for heat tolerance in broccoli has potential to both expand viable production areas and extend the growing season but breeding efficiency is constrained by limited genetic information. A doubled haploid (DH) broccoli population segregating for heat tolerance was evaluated for head quality in three summer fields in Charleston, SC, USA. Multiple quantitative trait loci (QTL) mapping of 1,423 single nucleotide polymorphisms developed through genotyping-by-sequencing identified five QTL and one positive epistatic interaction that explained 62.1% of variation in heat tolerance. The QTL identified here can be used to develop markers for marker-assisted selection and to increase our understanding of the molecular mechanisms underlying plant response to heat stress.

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.)

Intergenic sequence between Arabidopsis caseinolytic protease B-cytoplasmic/heat shock protein100 and choline kinase genes functions as a heat-inducible bidirectional promoter.

Science.gov (United States)

Mishra, Ratnesh Chandra; Grover, Anil

2014-11-01

In Arabidopsis (Arabidopsis thaliana), the At1g74310 locus encodes for caseinolytic protease B-cytoplasmic (ClpB-C)/heat shock protein100 protein (AtClpB-C), which is critical for the acquisition of thermotolerance, and At1g74320 encodes for choline kinase (AtCK2) that catalyzes the first reaction in the Kennedy pathway for phosphatidylcholine biosynthesis. Previous work has established that the knockout mutants of these genes display heat-sensitive phenotypes. While analyzing the AtClpB-C promoter and upstream genomic regions in this study, we noted that AtClpB-C and AtCK2 genes are head-to-head oriented on chromosome 1 of the Arabidopsis genome. Expression analysis showed that transcripts of these genes are rapidly induced in response to heat stress treatment. In stably transformed Arabidopsis plants harboring this intergenic sequence between head-to-head oriented green fluorescent protein and β-glucuronidase reporter genes, both transcripts and proteins of the two reporters were up-regulated upon heat stress. Four heat shock elements were noted in the intergenic region by in silico analysis. In the homozygous transfer DNA insertion mutant Salk_014505, 4,393-bp transfer DNA is inserted at position -517 upstream of ATG of the AtClpB-C gene. As a result, AtCk2 loses proximity to three of the four heat shock elements in the mutant line. Heat-inducible expression of the AtCK2 transcript was completely lost, whereas the expression of AtClpB-C was not affected in the mutant plants. Our results suggest that the 1,329-bp intergenic fragment functions as a heat-inducible bidirectional promoter and the region governing the heat inducibility is possibly shared between the two genes. We propose a model in which AtClpB-C shares its regulatory region with heat-induced choline kinase, which has a possible role in heat signaling. © 2014 American Society of Plant Biologists. All Rights Reserved.

Sequencing Batch Reactor (SBR) for the removal of Hg2+ and Cd2+ from synthetic petrochemical factory wastewater

International Nuclear Information System (INIS)

Malakahmad, Amirhossein; Hasani, Amirhesam; Eisakhani, Mahdieh; Isa, Mohamed Hasnain

2011-01-01

Highlights: → We assessed SBR performances to treat synthetic wastewater containing Hg 2+ and Cd 2+ . → SBR was able to remove 76-90% of Hg 2+ and 96-98% of Cd 2+ . → COD removal efficiency and MLVSS was affected by Hg 2+ and Cd 2+ concentrations. → Removal was not only biological process but also by biosorption process of sludge. - Abstract: Petrochemical factories which manufacture vinyl chloride monomer and poly vinyl chloride (PVC) are among the largest industries which produce wastewater contains mercury and cadmium. The objective of this research is to evaluate the performance of a lab-scale Sequencing Batch Reactor (SBR) to treat a synthetic petrochemical wastewater containing mercury and cadmium. After acclimatization of the system which lasted 60 days, the SBR was introduced to mercury and cadmium in low concentrations which then was increased gradually to 9.03 ± 0.02 mg/L Hg and 15.52 ± 0.02 mg/L Cd until day 110. The SBR performance was assessed by measuring Chemical Oxygen Demand, Total and Volatile Suspended Solids as well as Sludge Volume Index. At maximum concentrations of the heavy metals, the SBR was able to remove 76-90% of Hg 2+ and 96-98% of Cd 2+ . The COD removal efficiency and MLVSS (microorganism population) in the SBR was affected by mercury and cadmium concentrations in influent. Different species of microorganisms such as Rhodospirilium-like bacteria, Gomphonema-like algae, and sulfate reducing-like bacteria were identified in the system. While COD removal efficiency and MLVSS concentration declined during addition of heavy metals, the appreciable performance of SBR in removal of Hg 2+ and Cd 2+ implies that the removal in SBR was not only a biological process, but also by the biosorption process of the sludge.

Removing the bottleneck in whole genome sequencing of Mycobacterium tuberculosis for rapid drug resistance analysis: a call to action

Directory of Open Access Journals (Sweden)

Ruth McNerney

2017-03-01

Full Text Available Whole genome sequencing (WGS can provide a comprehensive analysis of Mycobacterium tuberculosis mutations that cause resistance to anti-tuberculosis drugs. With the deployment of bench-top sequencers and rapid analytical software, WGS is poised to become a useful tool to guide treatment. However, direct sequencing from clinical specimens to provide a full drug resistance profile remains a serious challenge. This article reviews current practices for extracting M. tuberculosis DNA and possible solutions for sampling sputum. Techniques under consideration include enzymatic digestion, physical disruption, chemical degradation, detergent solubilization, solvent extraction, ligand-coated magnetic beads, silica columns, and oligonucleotide pull-down baits. Selective amplification of genomic bacterial DNA in sputum prior to WGS may provide a solution, and differential lysis to reduce the levels of contaminating human DNA is also being explored. To remove this bottleneck and accelerate access to WGS for patients with suspected drug-resistant tuberculosis, it is suggested that a coordinated and collaborative approach be taken to more rapidly optimize, compare, and validate methodologies for sequencing from patient samples.

OligoHeatMap (OHM): an online tool to estimate and display hybridizations of oligonucleotides onto DNA sequences.

Science.gov (United States)

Croce, Olivier; Chevenet, François; Christen, Richard

2008-07-01

The efficiency of molecular methods involving DNA/DNA hybridizations depends on the accurate prediction of the melting temperature (T(m)) of the duplex. Many softwares are available for T(m) calculations, but difficulties arise when one wishes to check if a given oligomer (PCR primer or probe) hybridizes well or not on more than a single sequence. Moreover, the presence of mismatches within the duplex is not sufficient to estimate specificity as it does not always significantly decrease the T(m). OHM (OligoHeatMap) is an online tool able to provide estimates of T(m) for a set of oligomers and a set of aligned sequences, not only as text files of complete results but also in a graphical way: T(m) values are translated into colors and displayed as a heat map image, either stand alone or to be used by softwares such as TreeDyn to be included in a phylogenetic tree. OHM is freely available at http://bioinfo.unice.fr/ohm/, with links to the full source code and online help.

Heat pipes for ground heating and cooling

Energy Technology Data Exchange (ETDEWEB)

Vasiliev, L L

1988-01-01

Different versions of heat pipe ground heating and cooling devices are considered. Solar energy, biomass, ground stored energy, recovered heat of industrial enterprises and ambient cold air are used as energy and cold sources. Heat pipe utilization of air in winter makes it possible to design accumulators of cold and ensures deep freezing of ground in order to increase its mechanical strength when building roadways through the swamps and ponds in Siberia. Long-term underground heat storage systems are considered, in which the solar and biomass energy is accumulated and then transferred to heat dwellings and greenhouses, as well as to remove snow from roadways with the help of heat pipes and solar collectors.

De novo RNA sequencing transcriptome of Rhododendron obtusum identified the early heat response genes involved in the transcriptional regulation of photosynthesis.

Directory of Open Access Journals (Sweden)

Linchuan Fang

Full Text Available Rhododendron spp. is an important ornamental species that is widely cultivated for landscape worldwide. Heat stress is a major obstacle for its cultivation in south China. Previous studies on rhododendron principally focused on its physiological and biochemical processes, which are involved in a series of stress tolerance. However, molecular or genetic properties of rhododendron's response to heat stress are still poorly understood. The phenotype and chlorophyll fluorescence kinetics parameters of four rhododendron cultivars were compared under normal or heat stress conditions, and a cultivar with highest heat tolerance, "Yanzhimi" (R. obtusum was selected for transcriptome sequencing. A total of 325,429,240 high quality reads were obtained and assembled into 395,561 transcripts and 92,463 unigenes. Functional annotation showed that 38,724 unigenes had sequence similarity to known genes in at least one of the proteins or nucleotide databases used in this study. These 38,724 unigenes were categorized into 51 functional groups based on Gene Ontology classification and were blasted to 24 known cluster of orthologous groups. A total of 973 identified unigenes belonged to 57 transcription factor families, including the stress-related HSF, DREB, ZNF, and NAC genes. Photosynthesis was significantly enriched in the Kyoto Encyclopedia of Genes and Genomes pathway, and the changed expression pattern was illustrated. The key pathways and signaling components that contribute to heat tolerance in rhododendron were revealed. These results provide a potentially valuable resource that can be used for heat-tolerance breeding.

De novo RNA sequencing transcriptome of Rhododendron obtusum identified the early heat response genes involved in the transcriptional regulation of photosynthesis

Science.gov (United States)

Tong, Jun; Dong, Yanfang; Xu, Dongyun; Mao, Jing; Zhou, Yuan

2017-01-01

Rhododendron spp. is an important ornamental species that is widely cultivated for landscape worldwide. Heat stress is a major obstacle for its cultivation in south China. Previous studies on rhododendron principally focused on its physiological and biochemical processes, which are involved in a series of stress tolerance. However, molecular or genetic properties of rhododendron’s response to heat stress are still poorly understood. The phenotype and chlorophyll fluorescence kinetics parameters of four rhododendron cultivars were compared under normal or heat stress conditions, and a cultivar with highest heat tolerance, “Yanzhimi” (R. obtusum) was selected for transcriptome sequencing. A total of 325,429,240 high quality reads were obtained and assembled into 395,561 transcripts and 92,463 unigenes. Functional annotation showed that 38,724 unigenes had sequence similarity to known genes in at least one of the proteins or nucleotide databases used in this study. These 38,724 unigenes were categorized into 51 functional groups based on Gene Ontology classification and were blasted to 24 known cluster of orthologous groups. A total of 973 identified unigenes belonged to 57 transcription factor families, including the stress-related HSF, DREB, ZNF, and NAC genes. Photosynthesis was significantly enriched in the Kyoto Encyclopedia of Genes and Genomes pathway, and the changed expression pattern was illustrated. The key pathways and signaling components that contribute to heat tolerance in rhododendron were revealed. These results provide a potentially valuable resource that can be used for heat-tolerance breeding. PMID:29059200

Natural Circulation in the Blanket Heat Removal System During a Loss-of-Pumping Accident (LOFA) Based on Initial Conceptual Design

International Nuclear Information System (INIS)

Hamm, L.L.

1998-01-01

A transient natural convection model of the APT blanket primary heat removal (HR) system was developed to demonstrate that the blanket could be cooled for a sufficient period of time for long term cooling to be established following a loss-of-flow accident (LOFA). The particular case of interest in this report is a complete loss-of-pumping accident. For the accident scenario in which pumps are lost in both the target and blanket HR systems, natural convection provides effective cooling of the blanket for approximately 68 hours, and, if only the blanket HR systems are involved, natural convection is effective for approximately 210 hours. The heat sink for both of these accident scenarios is the assumed stagnant fluid and metal on the secondary sides of the heat exchangers

Modeling of a heat sink and high heat flux vapor chamber

Science.gov (United States)

Vadnjal, Aleksander

An increasing demand for a higher heat flux removal capability within a smaller volume for high power electronics led us to focus on a novel cold plate design. A high heat flux evaporator and micro channel heat sink are the main components of a cold plate which is capable of removing couple of 100 W/cm2. In order to describe performance of such porous media device a proper modeling has to be addressed. A universal approach based on the volume average theory (VAT) to transport phenomena in porous media is shown. An approach on how to treat the closure for momentum and energy equations is addressed and a proper definition for friction factors and heat transfer coefficients are discussed. A numerical scheme using a solution to Navier-Stokes equations over a representative elementary volume (REV) and the use of VAT is developed to show how to compute friction factors and heat transfer coefficients. The calculation show good agreement with the experimental data. For the heat transfer coefficient closure, a proper average for both fluid and solid is investigated. Different types of heating are also investigated in order to determine how it influences the heat transfer coefficient. A higher heat fluxes in small area condensers led us to the micro channels in contrast to the classical heat fin design. A micro channel can have various shapes to enhance heat transfer, but the shape that will lead to a higher heat flux removal with a moderate pumping power needs to be determined. The standard micro-channel terminology is usually used for channels with a simple cross section, e.g. square, round, triangle, etc., but here the micro channel cross section is going to be expanded to describe more complicated and interconnected micro scale channel cross sections. The micro channel geometries explored are pin fins (in-line and staggered) and sintered porous micro channels. The problem solved here is a conjugate problem involving two heat transfer mechanisms; (1) porous media

Effect of local versus remote tonic heat pain during training on acquisition and retention of a finger-tapping sequence task.

Science.gov (United States)

Bilodeau, Marie-Claude; Roosink, Meyke; Mercier, Catherine

2016-02-01

Although pain is present in a large proportion of patients receiving rehabilitation, its impact on motor learning is still unclear, especially in the case of neuropathic pain that is not tightly linked to specific movements. The aim of this study was to determine the effect of local and remote tonic cutaneous heat pain applied during training on motor learning of a finger-tapping sequence task. Forty-five healthy participants, randomized to the control, local pain or remote pain groups, were trained to perform an explicit finger motor sequence of five items as fast as possible. During the 10 training blocks (30 s each), local pain and remote pain groups received a heat pain stimulus on the wrist or leg, respectively. Performance was tested in the absence of pain in all groups before (baseline), immediately after (post-immediate), 60 min after (post-60 min) and 24 h after training (post-24 h) to assess both acquisition and next-day retention. Speed increased over time from baseline to post-24 h (p pain during training. No changes were observed on error rates, which were very low even at baseline. These results with experimental heat pain suggest that the ability to relearn finger sequence should not be affected by concomitant neuropathic pain in neurorehabilitation. However, these results need to be validated in the context of chronic pain, by including pain as a co-variable in motor rehabilitation trials.

Evaluation of Heat Transfer to the Implant-Bone Interface During Removal of Metal Copings Cemented onto Titanium Abutments.

Science.gov (United States)

Cakan, Umut; Cakan, Murat; Delilbasi, Cagri

2016-01-01

The aim of this investigation was to measure the temperature increase due to heat transferred to the implant-bone interface when the abutment screw channel is accessed or a metal-ceramic crown is sectioned buccally with diamond or tungsten carbide bur using an air rotor, with or without irrigation. Cobalt-chromium copings were cemented onto straight titanium abutments. The temperature changes during removal of the copings were recorded over a period of 1 minute. The sectioning of coping with diamond bur and without water irrigation generated the highest temperature change at the cervical part of the implant. Both crown removal methods resulted in an increase in temperature at the implant-bone interface. However, this temperature change did not exceed 47°C, the potentially damaging threshold for bone reported in the literature.

The use of ferrofluids for heat removal: Advantage or disadvantage?

Energy Technology Data Exchange (ETDEWEB)

Krauzina, Marina T., E-mail: krauzina@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Bozhko, Aleksandra A., E-mail: bozhko@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Krauzin, Pavel V., E-mail: krauzin@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Suslov, Sergey A., E-mail: ssuslov@swin.edu.au [Department of Mathematics H38, Swinburne University of Technology, Hawthorn, Victoria 3122 (Australia)

2017-06-01

It is shown experimentally that, depending on the relative orientation of the gravity and the thermal gradient and on the pre-history of experiment, the application of a uniform external vertical magnetic field to a spherical cavity filled with magnetic ferrofluid can either enhance or suppress a convective heat transfer. - Highlights: • Conduction heat transfer in magnetic fluid heated from above is stronger than that in a fluid not containing nanoparticles. • The application of a uniform vertical magnetic field enhances heat transfer when magnetic fluid is heated from above. • Heat transfer in a magnetic fluid heated from below is weaker than that in a fluid not containing nanoparticles.

A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

International Nuclear Information System (INIS)

Sobajima, M.

1998-01-01

The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

Residual heat removal pump and low pressure safety injection pump retrofit program

International Nuclear Information System (INIS)

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

1992-01-01

Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation

International Nuclear Information System (INIS)

Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

1991-02-01

This paper identifies the topics 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 would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab

Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Allard, L.T.; Elerath, J.G.

1976-02-01

This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

SSH analysis of endosperm transcripts and characterization of heat stress regulated expressed sequence tags in bread wheat

Directory of Open Access Journals (Sweden)

Suneha Goswami

2016-08-01

Full Text Available Heat stress is one of the major problems in agriculturally important cereal crops, especially wheat. Here, we have constructed a subtracted cDNA library from the endosperm of HS-treated (42°C for 2 h wheat cv. HD2985 by suppression subtractive hybridization (SSH. We identified ~550 recombinant clones ranging from 200 to 500 bp with an average size of 300 bp. Sanger’s sequencing was performed with 205 positive clones to generate the differentially expressed sequence tags (ESTs. Most of the ESTs were observed to be localized on the long arm of chromosome 2A and associated with heat stress tolerance and metabolic pathways. Identified ESTs were BLAST search using Ensemble, TriFLD and TIGR databases and the predicted CDS were translated and aligned with the protein sequences available in pfam and InterProScan 5 databases to predict the differentially expressed proteins (DEPs. We observed eight different types of post-translational modifications (PTMs in the DEPs corresponds to the cloned ESTs—147 sites with phosphorylation, 21 sites with sumoylation, 237 with palmitoylation, 96 sites with S-nitrosylation, 3066 calpain cleavage sites, and 103 tyrosine nitration sites, predicted to sense the heat stress and regulate the expression of stress genes. Twelve DEPs were observed to have transmembrane helixes (TMH in their structure, predicted to play the role of sensors of HS. Quantitative Real-Time PCR of randomly selected ESTs showed very high relative expression of HSP17 under HS; up-regulation was observed more in wheat cv. HD2985 (thermotolerant, as compared to HD2329 (thermosusceptible during grain-filling. The abundance of transcripts was further validated through northern blot analysis. The ESTs and their corresponding DEPs can be used as molecular marker for screening or targeted precision breeding program. PTMs identified in the DEPs can be used to elucidate the thermotolerance mechanism of wheat – a novel step towards the development of �

Experience on sodium removal from various components

Energy Technology Data Exchange (ETDEWEB)

Kamei, M; Kanbe, M; Yagisawa, H; Sasaki, S; Kataoka, H; Fukada, T; Ishii, Y; Saito, R; Mimoto, Y [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.; Fukada, T.; Ishii, Y.; Saito, R.; Mimoto, Y.

1978-01-01

Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

Experience on sodium removal from various components

International Nuclear Information System (INIS)

Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.

1978-02-01

Since 1970, OEC (O-arai Engineering Center) has been investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of ''JOYO'' and Dummy fuel assembly of ''JOYO'', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of ''JOYO'', a sector model of Sodium-to-Air cooler of ''JOYO'' and a proto-type Isolation valve of ''JOYO'' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental subassemblies, the Fuel Handling Machine of ''MONJU'' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of Sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a ''JOYO'' prototype pump by reinstalling it after sodium removal five times. (author)

Thermal control system. [removing waste heat from industrial process spacecraft

Science.gov (United States)

Hewitt, D. R. (Inventor)

1983-01-01

The temperature of an exothermic process plant carried aboard an Earth orbiting spacecraft is regulated using a number of curved radiator panels accurately positioned in a circular arrangement to form an open receptacle. A module containing the process is insertable into the receptacle. Heat exchangers having broad exterior surfaces extending axially above the circumference of the module fit within arcuate spacings between adjacent radiator panels. Banks of variable conductance heat pipes partially embedded within and thermally coupled to the radiator panels extend across the spacings and are thermally coupled to broad exterior surfaces of the heat exchangers by flanges. Temperature sensors monitor the temperature of process fluid flowing from the module through the heat exchanges. Thermal conduction between the heat exchangers and the radiator panels is regulated by heating a control fluid within the heat pipes to vary the effective thermal length of the heat pipes in inverse proportion to changes in the temperature of the process fluid.

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

International Nuclear Information System (INIS)

Warnke, E.P.

1990-02-01

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

Removing Noise From Pyrosequenced Amplicons

Directory of Open Access Journals (Sweden)

Davenport Russell J

2011-01-01

Full Text Available Abstract Background In many environmental genomics applications a homologous region of DNA from a diverse sample is first amplified by PCR and then sequenced. The next generation sequencing technology, 454 pyrosequencing, has allowed much larger read numbers from PCR amplicons than ever before. This has revolutionised the study of microbial diversity as it is now possible to sequence a substantial fraction of the 16S rRNA genes in a community. However, there is a growing realisation that because of the large read numbers and the lack of consensus sequences it is vital to distinguish noise from true sequence diversity in this data. Otherwise this leads to inflated estimates of the number of types or operational taxonomic units (OTUs present. Three sources of error are important: sequencing error, PCR single base substitutions and PCR chimeras. We present AmpliconNoise, a development of the PyroNoise algorithm that is capable of separately removing 454 sequencing errors and PCR single base errors. We also introduce a novel chimera removal program, Perseus, that exploits the sequence abundances associated with pyrosequencing data. We use data sets where samples of known diversity have been amplified and sequenced to quantify the effect of each of the sources of error on OTU inflation and to validate these algorithms. Results AmpliconNoise outperforms alternative algorithms substantially reducing per base error rates for both the GS FLX and latest Titanium protocol. All three sources of error lead to inflation of diversity estimates. In particular, chimera formation has a hitherto unrealised importance which varies according to amplification protocol. We show that AmpliconNoise allows accurate estimates of OTU number. Just as importantly AmpliconNoise generates the right OTUs even at low sequence differences. We demonstrate that Perseus has very high sensitivity, able to find 99% of chimeras, which is critical when these are present at high

Genesis Solar Wind Sample 61422: Experiment in Variation of Sequence of Cleaning Solvent for Removing Carbon-Bearing Contamination

Science.gov (United States)

Allton, J. H.; Kuhlman, K. R.; Allums, K. K.; Gonzalez, C. P.; Jurewicz, A. J. G.; Burnett, D. S.; Woolum, D. S.

2015-01-01

The recovered Genesis collector fragments are heavily contaminated with crash-derived particulate debris. However, megasonic treatment with ultra-pure-water (UPW; resistivity (is) greater than18 meg-ohm-cm) removes essentially all particulate contamination greater than 5 microns in size [e.g.1] and is thus of considerable importance. Optical imaging of Si sample 60336 revealed the presence of a large C-rich particle after UPW treatment that was not present prior to UPW. Such handling contamination is occasionally observed, but such contaminants are normally easily removed by UPW cleaning. The 60336 particle was exceptional in that, surprisingly, it was not removed by additional UPW or by hot xylene or by aqua regia treatment. It was eventually removed by treatment with NH3-H2O2. Our best interpretation of the origin of the 60336 particle was that it was adhesive from the Post-It notes used to stabilize samples for transport from Utah after the hard landing. It is possible that the insoluble nature of the 60336 particle comes from interaction of the Post-It adhesive with UPW. An occasional bit of Post-It adhesive is not a major concern, but C particulate contamination also occurs from the heat shield of the Sample Return Capsule (SRC) and this is mixed with inorganic contamination from the SRC and the Utah landing site. If UPW exposure also produced an insoluble residue from SRC C, this would be a major problem in chemical treatments to produce clean surfaces for analysis. This paper reports experiments to test whether particulate contamination was removed more easily if UPW treatment was not used.

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Sequencing Batch Reactor (SBR) for the removal of Hg{sup 2+} and Cd{sup 2+} from synthetic petrochemical factory wastewater

Energy Technology Data Exchange (ETDEWEB)

Malakahmad, Amirhossein, E-mail: amirhossein@petronas.com.my [Faculty of Energy and Environmental Studies, Islamic Azad University, Science and Research branch, Hesarak, Tehran (Iran, Islamic Republic of); Civil Engineering Department, Universiti Teknologi PETRONAS, Tronoh, Perak (Malaysia); Hasani, Amirhesam [Faculty of Energy and Environmental Studies, Islamic Azad University, Science and Research branch, Hesarak, Tehran (Iran, Islamic Republic of); Eisakhani, Mahdieh [School of Social, Development and the Environment, Universiti Kebangsaan Malaysia, Bangi, Selangor (Malaysia); Isa, Mohamed Hasnain [Civil Engineering Department, Universiti Teknologi PETRONAS, Tronoh, Perak (Malaysia)

2011-07-15

Highlights: {yields} We assessed SBR performances to treat synthetic wastewater containing Hg{sup 2+} and Cd{sup 2+}. {yields} SBR was able to remove 76-90% of Hg{sup 2+} and 96-98% of Cd{sup 2+}. {yields} COD removal efficiency and MLVSS was affected by Hg{sup 2+} and Cd{sup 2+} concentrations. {yields} Removal was not only biological process but also by biosorption process of sludge. - Abstract: Petrochemical factories which manufacture vinyl chloride monomer and poly vinyl chloride (PVC) are among the largest industries which produce wastewater contains mercury and cadmium. The objective of this research is to evaluate the performance of a lab-scale Sequencing Batch Reactor (SBR) to treat a synthetic petrochemical wastewater containing mercury and cadmium. After acclimatization of the system which lasted 60 days, the SBR was introduced to mercury and cadmium in low concentrations which then was increased gradually to 9.03 {+-} 0.02 mg/L Hg and 15.52 {+-} 0.02 mg/L Cd until day 110. The SBR performance was assessed by measuring Chemical Oxygen Demand, Total and Volatile Suspended Solids as well as Sludge Volume Index. At maximum concentrations of the heavy metals, the SBR was able to remove 76-90% of Hg{sup 2+} and 96-98% of Cd{sup 2+}. The COD removal efficiency and MLVSS (microorganism population) in the SBR was affected by mercury and cadmium concentrations in influent. Different species of microorganisms such as Rhodospirilium-like bacteria, Gomphonema-like algae, and sulfate reducing-like bacteria were identified in the system. While COD removal efficiency and MLVSS concentration declined during addition of heavy metals, the appreciable performance of SBR in removal of Hg{sup 2+} and Cd{sup 2+} implies that the removal in SBR was not only a biological process, but also by the biosorption process of the sludge.

The mechanism of enhanced wastewater nitrogen removal by photo-sequencing batch reactors based on comprehensive analysis of system dynamics within a cycle.

Science.gov (United States)

Ye, Jianfeng; Liang, Junyu; Wang, Liang; Markou, Giorgos

2018-07-01

To understand the mechanism of enhanced nitrogen removal by photo-sequencing batch reactors (photo-SBRs), which incorporated microalgal photosynthetic oxygenation into the aerobic phases of a conventional cycle, this study performed comprehensive analysis of one-cycle dynamics. Under a low aeration intensity (about 0.02 vvm), a photo-SBR, illuminated with light at 92.27 μ·mol·m -2 ·s -1 , could remove 99.45% COD, 99.93% NH 4 + -N, 90.39% TN, and 95.17% TP, while the control SBR could only remove 98.36% COD, 83.51% NH 4 + -N, 78.96% TN, and 97.75% TP, for a synthetic domestic sewage. The specific oxygen production rate (SOPR) of microalgae in the photo-SBR could reach 6.63 fmol O 2 ·cell -1 ·h -1 . One-cycle dynamics shows that the enhanced nitrogen removal by photo-SBRs is related to photosynthetic oxygenation, resulting in strengthened nitrification, instead of direct nutrient uptake by microalgae. A too high light or aeration intensity could deteriorate anoxic conditions and thus adversely affect the removal of TN and TP in photo-SBRs. Copyright © 2018 Elsevier Ltd. All rights reserved.

LMFBR post accident heat removal testing needs and conceptual design of a test facility

International Nuclear Information System (INIS)

Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

1977-03-01

A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

Study on heat removal capability concrete cask system with horizontal orientation

International Nuclear Information System (INIS)

Nabemoto, Toyonobu; Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatsugu

2002-01-01

In Japan, nuclear fuel cycle, has been promoted, so the recycle fuels formed at nuclear power stations are planned to be processed at reprocessing facilities in future. However, as forming quantities of the recycle fuels are more than reprocessing quantities of the facilities, it is needed to practice a facility (interim storage facility (ISF)) to temporarily store them among the recycle fuels will be reprocessed. The Ishikawajima-Harima Heavy Industries, Co., Ltd. has investigated on vault system and concrete cask system for dry storage system with excellent economical efficiency among various systems on ISFs. As the latter method has a number of actual results in U.S.A., its practice is progressed after some improvements suitable for Japan. When progressing this practice on the latter method on fiscal year 1999, at first, a concrete cask with actual size was experimentally produced, to confirm its productivity. On fiscal year 2000, aiming to establish heat removal evaluation at storage, a thermal load test simulated at the storage was carried out by using this trial product. Here was reported results obtained at a test simulated at repacking carried out on fiscal year 2001. (G.K.)

Development of a steady-state calculation model for the KALIMER PDRC(Passive Decay Heat Removal Circuit)

International Nuclear Information System (INIS)

Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Kwon, Young Min; Eoh, Jae Hyuk; Lee, Yong Bum

2003-06-01

A sodium circuit has usually featured for a Liquid Metal Reactor(LMR) using sodium as coolant to remove the decay heat ultimately under accidental conditions because of its high reliability. Most of the system codes used for a Light Water Reactor(LWR) analysis is capable of calculating natural circulation within such circuit, but the code currently used for the LMR analysis does not feature stand alone capability to simulate the natural circulation flow inside the circuit due to its application limitation. To this end, the present study has been carried out because the natural circulation analysis for such the circuit is realistically raised for the design with a new concept. The steady state modeling is presented in this paper, development of a transient model is also followed to close the study. The incompressibility assumption of sodium which allow the circuit to be modeled with a single flow, makes the model greatly simplified. Models such as a heat exchanger developed in the study can be effectively applied to other system analysis codes which require such component models

Effects of remediation train sequence on decontamination of heavy metal-contaminated soil containing mercury.

Science.gov (United States)

Hseu, Zeng-Yei; Huang, Yu-Tuan; Hsi, Hsing-Cheng

2014-09-01

When a contaminated site contains pollutants including both nonvolatile metals and Hg, one single remediation technology may not satisfactorily remove all contaminants. Therefore, in this study, chemical extraction and thermal treatment were combined as a remediation train to remove heavy metals, including Hg, from contaminated soil. A 0.2 M solution of ethylenediamine tetraacetic acid (EDTA) was shown to be the most effective reagent for extraction of considerable amounts of Cu, Pb, and Zn (> 50%). Hg removal was ineffective using 0.2 M EDTA, but thermogravimetric analysis suggested that heating to 550 degrees C with a heating rate of 5 degrees C/min for a duration of 1 hr appeared to be an effective approach for Hg removal. With the employment of thermal treatment, up to 99% of Hg could be removed. However executing thermal treatment prior to chemical extraction reduced the effectiveness of the subsequent EDTA extraction because nonvolatile heavy metals were immobilized in soil aggregates after the 550 degrees C treatment. The remediation train of chemical extraction followed by thermal treatment appears to remediate soils that have been contaminated by many nonvolatile heavy metals and Hg. Implications: A remediation train conjoining two or more techniques has been initialized to remove multiple metals. Better understandings of the impacts of treatment sequences, namely, which technique should be employed first on the soil properties and the decontamination efficiency, are in high demand. This study provides a strategy to remove multiple heavy metals including Hg from a contaminated soil. The interactions between thermal treatment and chemical extraction on repartitioning of heavy metals was revealed. The obtained results could offer an integrating strategy to remediate the soil contaminated with both heavy metals and volatile contaminants.

Chemical heat pump and chemical energy storage system

Science.gov (United States)

Clark, Edward C.; Huxtable, Douglas D.

1985-08-06

A chemical heat pump and storage system employs sulfuric acid and water. In one form, the system includes a generator and condenser, an evaporator and absorber, aqueous acid solution storage and water storage. During a charging cycle, heat is provided to the generator from a heat source to concentrate the acid solution while heat is removed from the condenser to condense the water vapor produced in the generator. Water is then stored in the storage tank. Heat is thus stored in the form of chemical energy in the concentrated acid. The heat removed from the water vapor can be supplied to a heat load of proper temperature or can be rejected. During a discharge cycle, water in the evaporator is supplied with heat to generate water vapor, which is transmitted to the absorber where it is condensed and absorbed into the concentrated acid. Both heats of dilution and condensation of water are removed from the thus diluted acid. During the discharge cycle the system functions as a heat pump in which heat is added to the system at a low temperature and removed from the system at a high temperature. The diluted acid is stored in an acid storage tank or is routed directly to the generator for reconcentration. The generator, condenser, evaporator, and absorber all are operated under pressure conditions specified by the desired temperature levels for a given application. The storage tanks, however, can be maintained at or near ambient pressure conditions. In another form, the heat pump system is employed to provide usable heat from waste process heat by upgrading the temperature of the waste heat.

De novo transcriptome sequencing of Isaria cateniannulata and comparative analysis of gene expression in response to heat and cold stresses.

Directory of Open Access Journals (Sweden)

Dingfeng Wang

Full Text Available Isaria cateniannulata is a very important and virulent entomopathogenic fungus that infects many insect pest species. Although I. cateniannulata is commonly exposed to extreme environmental temperature conditions, little is known about its molecular response mechanism to temperature stress. Here, we sequenced and de novo assembled the transcriptome of I. cateniannulata in response to high and low temperature stresses using Illumina RNA-Seq technology. Our assembly encompassed 17,514 unigenes (mean length = 1,197 bp, in which 11,445 unigenes (65.34% showed significant similarities to known sequences in NCBI non-redundant protein sequences (Nr database. Using digital gene expression analysis, 4,483 differentially expressed genes (DEGs were identified after heat treatment, including 2,905 up-regulated genes and 1,578 down-regulated genes. Under cold stress, 1,927 DEGs were identified, including 1,245 up-regulated genes and 682 down-regulated genes. The expression patterns of 18 randomly selected candidate DEGs resulting from quantitative real-time PCR (qRT-PCR were consistent with their transcriptome analysis results. Although DEGs were involved in many pathways, we focused on the genes that were involved in endocytosis: In heat stress, the pathway of clathrin-dependent endocytosis (CDE was active; however at low temperature stresses, the pathway of clathrin-independent endocytosis (CIE was active. Besides, four categories of DEGs acting as temperature sensors were observed, including cell-wall-major-components-metabolism-related (CWMCMR genes, heat shock protein (Hsp genes, intracellular-compatible-solutes-metabolism-related (ICSMR genes and glutathione S-transferase (GST. These results enhance our understanding of the molecular mechanisms of I. cateniannulata in response to temperature stresses and provide a valuable resource for the future investigations.

Heat transfer system

Science.gov (United States)

Not Available

1980-03-07

A heat transfer system for a nuclear reactor is described. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-01-15

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Time Separation Between Events in a Sequence: a Regional Property?

Science.gov (United States)

Muirwood, R.; Fitzenz, D. D.

2013-12-01

Earthquake sequences are loosely defined as events occurring too closely in time and space to appear unrelated. Depending on the declustering method, several, all, or no event(s) after the first large event might be recognized as independent mainshocks. It can therefore be argued that a probabilistic seismic hazard assessment (PSHA, traditionally dealing with mainshocks only) might already include the ground shaking effects of such sequences. Alternatively all but the largest event could be classified as an ';aftershock' and removed from the earthquake catalog. While in PSHA the question is only whether to keep or remove the events from the catalog, for Risk Management purposes, the community response to the earthquakes, as well as insurance risk transfer mechanisms, can be profoundly affected by the actual timing of events in such a sequence. In particular the repetition of damaging earthquakes over a period of weeks to months can lead to businesses closing and families evacuating from the region (as happened in Christchurch, New Zealand in 2011). Buildings that are damaged in the first earthquake may go on to be damaged again, even while they are being repaired. Insurance also functions around a set of critical timeframes - including the definition of a single 'event loss' for reinsurance recoveries within the 192 hour ';hours clause', the 6-18 month pace at which insurance claims are settled, and the annual renewal of insurance and reinsurance contracts. We show how temporal aspects of earthquake sequences need to be taken into account within models for Risk Management, and what time separation between events are most sensitive, both in terms of the modeled disruptions to lifelines and business activity as well as in the losses to different parties (such as insureds, insurers and reinsurers). We also explore the time separation between all events and between loss causing events for a collection of sequences from across the world and we point to the need to

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Six-phase soil heating accelerates VOC extraction from clay soil

International Nuclear Information System (INIS)

Gauglitz, P.A.; Roberts, J.S.; Bergsman, T.M.; Caley, S.M.; Heath, W.O.; Miller, M.C.; Moss, R.W.; Schalla, R.; Jarosch, T.R.; Eddy-Dilek, C.A.

1994-08-01

Six-Phase Soil Heating (SPSH) was demonstrated as a viable technology for heating low permeability soils containing volatile organic contaminants. Testing was performed as part of the Volatile Organic Compounds in Non-Arid Soils Integrated Demonstration (VOC Non-Arid ID) at the Savannah River Site. The soil at the integrated demonstration site is contaminated with perchloroethylene (PCE) and trichloroethylene (TCE); the highest soil contamination occurs in clay-rich zones that are ineffectively treated by conventional soil vapor extraction due to the very low permeability of the clay. The SPSH demonstration sought to heat the clay zone and enhance the performance of conventional soil vapor extraction. Thermocouples at thirty locations quantified the areal and vertical heating within the treated zone. Soil samples were collected before and after heating to quantify the efficacy of heat-enhanced vapor extraction of PCE and TCE from the clay soil. Samples were taken (essentially every foot) from six wells prior to heating and adjacent to these wells after heating. Results show that contaminant removal from the clay zone was 99.7% (median) within the electrode array. Outside the array where the soil was heated, but to only 50 degrees C, the removal efficiency was 93%, showing that heating accelerated the removal of VOCs from the clay soil. The accelerated remediation resulted from effective heating of the contaminated clay zone by SPSH. The temperature of the clay zone increased to 100 degrees C after 8 days of heating and was maintained near 100 degrees C for 17 days. Electrical heating removed 19,000 gal of water from the soil as steam, with peak removal rate of 1,500 gpd of condensed steam

Removal of Pb(II) from water by the activated carbon modified by nitric acid under microwave heating.

Science.gov (United States)

Yao, Shuheng; Zhang, Jiajun; Shen, Dekui; Xiao, Rui; Gu, Sai; Zhao, Ming; Liang, Junyu

2016-02-01

The rice husk based activated carbon (RH-AC) was treated by nitric acid under microwave heating, in order to improve its capability for the removal of heavy metal ions from water. The optimal conditions for the modification of RH-AC (M-RH-AC) were determined by means of orthogonal array experimental design, giving those as the concentration of nitric acid of 8mol/L, modification time of 15min, modification temperature of 130°C and microwave power of 800W. The characteristics of the M-RH-AC and RH-AC were examined by BET, XRD, Raman spectrum, pH titration, zeta potential, Boehm titration and FTIR analysis. The M-RH-AC has lower pore surface area, smaller crystallite, lower pHIEP and more oxygen-containing functional groups than the RH-AC. Removal capacity of Pb(II) ions by the M-RH-AC and RH-AC from water solution was estimated concerning the influence of contact time, pH value, and initial concentration. The equilibrium time of Pb(II) removal was found to be around 90min after modification process. Two kinetic models are adopted to describe the possible Pb(II) adsorption mechanism, finding that the adsorption rate of Pb(II) ions by the M-RH-AC is larger than that of RH-AC. Copyright © 2015 Elsevier Inc. All rights reserved.

Droplet heat transfer and chemical reactions during direct containment heating

International Nuclear Information System (INIS)

Baker, L. Jr.

1986-01-01

A simplified model of heat transfer and chemical reaction has been adapted to evaluate the expected behavior of droplets containing unreacted Zircaloy and stainless steel moving through the containment atmosphere during postulated accidents involving direct containment heating. The model includes internal and external diffusive resistances to reaction. The results indicate that reactions will be incomplete for many conditions characteristic of direct containment heating sequences

Innovative techniques for removing concrete surfaces

International Nuclear Information System (INIS)

McFarland, J.M.

1980-01-01

This report centers on the use of heat to decompose contaminated concrete to facilitate its removal. It discusses the use of electrical resistance heating and induction heating to cause differential expansion between the reinforcing steel and the concrete in order to spall the concrete. It introduces the concept of using induction heating to both decompose and spall steel impregnated concrete, acknowledging the work of Charles H. Henager in this field. The techniques are offered as theoretical and untested possibilities. Their practical application depends upon the effectiveness of alternatives and upon further development of these concepts

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Effectiveness of photocatalytic filter for removing volatile organic compounds in the heating, ventilation, and air conditioning system.

Science.gov (United States)

Yu, Kuo-Pin; Lee, Grace Whei-May; Huang, Wei-Ming; Wu, Chih-Cheng; Lou, Chia-ling; Yang, Shinhao

2006-05-01

Nowadays, the heating, ventilation, and air conditioning (HVAC) system has been an important facility for maintaining indoor air quality. However, the primary function of typical HVAC systems is to control the temperature and humidity of the supply air. Most indoor air pollutants, such as volatile organic compounds (VOCs), cannot be removed by typical HVAC systems. Thus, some air handling units for removing VOCs should be added in typical HVAC systems. Among all of the air cleaning techniques used to remove indoor VOCs, photocatalytic oxidation is an attractive alternative technique for indoor air purification and deodorization. The objective of this research is to investigate the VOC removal efficiency of the photocatalytic filter in a HVAC system. Toluene and formaldehyde were chosen as the target pollutants. The experiments were conducted in a stainless steel chamber equipped with a simplified HVAC system. A mechanical filter coated with Degussa P25 titania photocatalyst and two commercial photocatalytic filters were used as the photocatalytic filters in this simplified HVAC system. The total air change rates were controlled at 0.5, 0.75, 1, 1.25, and 1.5 hr(-1), and the relative humidity (RH) was controlled at 30%, 50%, and 70%. The ultraviolet lamp used was a 4-W, ultraviolet-C (central wavelength at 254 nm) strip light bulb. The first-order decay constant of toluene and formaldehyde found in this study ranged from 0.381 to 1.01 hr(-1) under different total air change rates, from 0.34 to 0.433 hr(-1) under different RH, and from 0.381 to 0.433 hr(-1) for different photocatalytic filters.

Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

International Nuclear Information System (INIS)

1980-01-01

This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

International Nuclear Information System (INIS)

Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung

2008-01-01

A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the service

Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2008-05-15

A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the

Transient Analysis of a Magnetic Heat Pump

Science.gov (United States)

Schroeder, E. A.

1985-01-01

An experimental heat pump that uses a rare earth element as the refrigerant is modeled using NASTRAN. The refrigerant is a ferromagnetic metal whose temperature rises when a magnetic field is applied and falls when the magnetic field is removed. The heat pump is used as a refrigerator to remove heat from a reservoir and discharge it through a heat exchanger. In the NASTRAN model the components modeled are represented by one-dimensional ROD elements. Heat flow in the solids and fluid are analyzed. The problem is mildly nonlinear since the heat capacity of the refrigerant is temperature-dependent. One simulation run consists of a series of transient analyses, each representing one stroke of the heat pump. An auxiliary program was written that uses the results of one NASTRAN analysis to generate data for the next NASTRAN analysis.

Method of extracting heat from dry geothermal reservoirs

Science.gov (United States)

Potter, R.M.; Robinson, E.S.; Smith, M.C.

1974-01-22

Hydraulic fracturing is used to interconnect two or more holes that penetrate a previously dry geothermal reservoir, and to produce within the reservoir a sufficiently large heat-transfer surface so that heat can be extracted from the reservoir at a usefully high rate by a fluid entering it through one hole and leaving it through another. Introduction of a fluid into the reservoir to remove heat from it and establishment of natural (unpumped) convective circulation through the reservoir to accomplish continuous heat removal are important and novel features of the method. (auth)

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

International Nuclear Information System (INIS)

Stratmann, W.; Hages, P.

2004-01-01

Within the scope of nuclear licensing procedures of on-site interim storage facilities for dual purpose casks it is necessary, among other things, to provide proof of sufficient removal of the residual decay heat emitted by the casks. The results of the analyses performed for this purpose define e.g. the boundary conditions for further thermal analyses regarding the permissible cask component temperatures or the maximum permissible temperatures of the fuel cladding tubes of the fuel elements stored in the casks. Up to now, for the centralized interim storage facilities in Germany such analyses were performed on the basis of experimental investigations using scaled-down storage geometries. In the engineering phase of the Lingen on-site interim storage facility, proof was furnished for the first time using the CFD (computational fluid dynamics) program FLUENT. The program FLUENT is an internationally recognized and comprehensively verified program for the calculation of flow and heat transport processes. Starting from a brief discussion of modeling and the different boundary conditions of the computation, this contribution presents various results regarding the temperatures of air, cask surfaces and storage facility components, the mass flows through the storage facility and the heat transfer at the cask surface. The interface point to the cask-specific analyses is defined to be the cask surface

Experimental evaluation of sodium to air heat exchanger performance

International Nuclear Information System (INIS)

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

2013-01-01

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

Graphene-enhanced thermal interface materials for heat removal from photovoltaic solar cells

Science.gov (United States)

Saadah, M.; Gamalath, D.; Hernandez, E.; Balandin, A. A.

2016-09-01

The increase in the temperature of photovoltaic (PV) solar cells affects negatively their power conversion efficiency and decreases their lifetime. The negative effects are particularly pronounced in concentrator solar cells. Therefore, it is crucial to limit the PV cell temperature by effectively removing the excess heat. Conventional thermal phase change materials (PCMs) and thermal interface materials (TIMs) do not possess the thermal conductivity values sufficient for thermal management of the next generation of PV cells. In this paper, we report the results of investigation of the increased efficiency of PV cells with the use of graphene-enhanced TIMs. Graphene reveals the highest values of the intrinsic thermal conductivity. It was also shown that the thermal conductivity of composites can be increased via utilization of graphene fillers. We prepared TIMs with up to 6% of graphene designed specifically for PV cell application. The solar cells were tested using the solar simulation module. It was found that the drop in the output voltage of the solar panel under two-sun concentrated illumination can be reduced from 19% to 6% when grapheneenhanced TIMs are used. The proposed method can recover up to 75% of the power loss in solar cells.

Heat exchanger restart evaluation

International Nuclear Information System (INIS)

Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

1992-01-01

On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized

Heat exchanger restart evaluation

International Nuclear Information System (INIS)

Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

1992-01-01

On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summary herein

Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

International Nuclear Information System (INIS)

Kim, Bo Gyung; Kang Hyun Gook; Yoon, Ho Joon

2016-01-01

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results

Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Abu Dhabi (United Arab Emirates)

2016-05-15

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results.

Influence of the Kinetics of Heat and Mass Transfer in a Binary-Rectification Column on the Realizability Range of its Regimes

Science.gov (United States)

Zaeva, M. A.; Tsirlin, A. M.; Sukin, I. A.

2018-05-01

The range of realizable rates of flows in a binary-rectification column in which heat is supplied into the boiler and is removed from the dephlegmator was investigated. It is shown that this range is determined by two characteristic parameters related to the kinetics of heat and mass transfer in the column and the composition of the mixture subjected to separation. The limiting capabilities of a cascade of two binary-rectification columns for the separation of a ternary mixture in it were considered. The conditions for an optimum sequence of separation of a mixture in this cascade and for a consistent arrangement of its heat and mass exchange surfaces and the relation between the ultimate production rate of the cascade and the total heat losses in it were determined.

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

Cyro Power and Heat Transfer

National Research Council Canada - National Science Library

Chow, L

1998-01-01

.... The heat generated from a 9x9-heater array was removed by liquid nitrogen pool boiling. The orientation and space limitation of the array were varied to explore their effects on the critical heat flux (CHF) value...

Validation of the TASS/SMR-S Code for the PRHRS Condensation Heat Transfer Model

International Nuclear Information System (INIS)