Reactor having coolant recycling pump

International Nuclear Information System (INIS)

Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

1991-01-01

In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

Fast reactors bulk sodium coolant disposal NOAH process application

International Nuclear Information System (INIS)

Magny, E. de; Berte, M.

1997-01-01

Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

Upper internals of PWR with coolant flow separator

International Nuclear Information System (INIS)

Chevereau, G.; Heuze, A.

1989-01-01

The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

2011-01-01

SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

Flow boiling test of GDP replacement coolants

International Nuclear Information System (INIS)

Park, S.H.

1995-01-01

The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

International Nuclear Information System (INIS)

2002-06-01

All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Correlation measurements of sodium flow rate with magnetic sensors

International Nuclear Information System (INIS)

Kebadze, B.V.; Krasnoyarov, N.V.; Adamovskij, L.A.; Golushko, V.V.; Sroelov, V.S.

1978-01-01

The results of bench-mark experiments and those carried out at the BOR-60 reactor to measure the sodium coolant flow rate by a correlation method are presented. The method is based on detecting the eddy type flow hydraulic nonuniformities using magnetic flowmeters. The measurements were fulfilled in a broad range of flow rates (G=10-10 4 m 3 /h, Re=2x10 5 -2x10 7 ). The measured and calculated mutual correlation functions are presented with parallel and perpendicular orientations of the flowmeters magnetic fields. A good accord is stated. Prerequirements to the arrangement of the measuring systems are formulated. As an important advantage of the correlation method a possibility of the flowmeter calibration in situ is hydhlighted

Automatic coolant flow control device for a nuclear reactor assembly

Science.gov (United States)

Hutter, Ernest

1986-01-01

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

A system for cooling electronic elements with an EHD coolant flow

International Nuclear Information System (INIS)

Tanski, M; Kocik, M; Barbucha, R; Garasz, K; Mizeraczyk, J; Kraśniewski, J; Oleksy, M; Hapka, A; Janke, W

2014-01-01

A system for cooling electronic components where the liquid coolant flow is forced with ion-drag type EHD micropumps was tested. For tests we used isopropyl alcohol as the coolant and CSD02060 diodes in TO-220 packages as cooled electronic elements. We have studied thermal characteristics of diodes cooled with EHD flow in the function of a coolant flow rate. The transient thermal impedance of the CSD02060 diode cooled with 1.5 ml/min EHD flow was 7.8°C/W. Similar transient thermal impedance can be achieved by applying to the diode a large RAD-A6405A/150 heat sink. We found out that EHD pumps can be successfully applied for cooling electronic elements.

Effect of ribbed and smooth coolant cross-flow channel on film cooling

International Nuclear Information System (INIS)

Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

2017-01-01

Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

Effect of ribbed and smooth coolant cross-flow channel on film cooling

Energy Technology Data Exchange (ETDEWEB)

Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

2017-05-15

Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

Labelling Of Coolant Flow Anomaly Using Fractal Structure

International Nuclear Information System (INIS)

Djainal, Djen Djen

1996-01-01

This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

Numerical experimentation on convective coolant flow in Ghana ...

African Journals Online (AJOL)

Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

International Nuclear Information System (INIS)

Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

2001-01-01

In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

Brenciaglia, G.; Gurevich, Y.; Liu, G.

1996-01-01

The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

The Experimental Characterization of the Magnetic Field Effect on a Liquid Sodium Flow

International Nuclear Information System (INIS)

Kim, Hee Reyoung; Kim, Jong Man; Cha, Jae Eun; Choi, Jong Hyun; Nam, Ho Yoon

2006-01-01

A liquid sodium coolant is used for a LMR such as KALIMER and a magnetic field is generated in the electromagnetic pump or flowmeter. The magnetic field has an effect on the electrically conducting metal flow by a generation of an electromagnetic pressure drop. Therefore, in the present study, a theoretical calculation is carried out for the effect of an external magnetic field and the magnetic field is measured over the electromagnet system manufactured for the magnetohydrodynamic experiments

The Theoretical Investigation of the Magnetic Field Effect on a Liquid Sodium Flow

International Nuclear Information System (INIS)

Kim, Hee Reyoung; Kim, Jong Man; Cha, Jae Eun; Choi, Jong Hyun; Nam, Ho Yoon

2005-01-01

The liquid sodium coolant is used for LMR such as KALIMER and magnetic field is generated in the electromagnetic pump or flowmeter. The magnetic field takes an effect on the electrically conducting metal flow by the generation of the electromagnetic pressure drop. Therefore, in the present study, the theoretical calculation is carried out for an effect from the external magnetic field and the magnetic field is firstly measured over the electromagnet system manufactured for the magnetohydrodynamic experiments

A Review of Fragmentation Models Relative to Molten UO2 Breakup when Quenched in Sodium Coolant

International Nuclear Information System (INIS)

Cronenberg, A.W.; Grolmes, M.A.

1976-01-01

An important aspect of the fuel-coolant interaction problem relative to liquid metal fast breeder reactor (LMFBR) safety analysis is the fragmentation of molten oxide fuel during contact with liquid sodium coolant. A proper description of the kinetics of such an event requires an understanding of the breakup process and an estimate of the size and dispersion of such finely divided fuel in coolant. In recent years, considerable interest has centered on the problem of determining the nature of such fragmentation. In this paper, both analytic and experimental studies pertaining to such breakup are reviewed in light of recent developments in the understanding of heat transfer and solidification phenomena during quenching of UO 2 in sodium. A more extensive review of this subject can be found in Ref. 1. In conclusion: As discussed, a number of models have been proposed in an attempt to understand the nature of the UO 2 fragmentation process. The four principle mechanisms considered likely to cause such fragmentation (impact forces, boiling, violent gas release, and shell solidification) have been developed to the point where comparative analysis is possible. In addition, recent developments in the understanding of the physics of oxide fuel behavior in sodium coolant (boiling regime criteria, vapor nucleation theories, and prediction of solidification kinetics enable us to asses whether or not the various model assumptions are realistic. In view of this knowledge the following conclusions are made. For the case of hydrodynamic influence on fragmentation, it can be said that although the disruptive forces of impact and viscous drag may contribute to breakup, their effects are not controlling with respect to high temperature materials, including UO 2 -sodium. With respect to the vapor bubble growth and collapse mechanism it was shown that for sodium quenching, where coolant contact may, be expected (as opposed to water), the thermodynamic work potential of the bubble is

Fast Flux Test Facility primary sodium valves

International Nuclear Information System (INIS)

Rabe, G.B.; Ezra, B.C.

1977-01-01

The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

Quantitative determination of a hydrogen impurity in a sodium coolant by hydride thermal dissociation

Science.gov (United States)

Ivanovskiy, M. N.; Pavlova, G. D.; Shmatko, B. A.; Milovidova, A. V.; Konovalov, E. YE.; Arnoldov, M. N.; Pleshivtsev, A. D.

1988-01-01

A molten sodium coolant containing hydrogen was heated in a vacuum at 450 C, and the gases generated pumped through a liquid nitrogen trap, and the H2 was then oxidized on a copper oxide substrate heated to 400 C. The accuracy of the method is 1.5 percent and the sensitivity is 1x10 to the -5 wt percent hydrogen.

Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

International Nuclear Information System (INIS)

Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

1976-04-01

The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

International Nuclear Information System (INIS)

Curtis, Franklin G.; Ekici, Kivanc; Freels, James D.

2011-01-01

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

New method for NPP sodium coolant pipeline austenization

International Nuclear Information System (INIS)

Malashonok, V.A.; Rotshtejn, A.V.; Gotshalk, A.L.; Miryushchenko, E.F.

1980-01-01

Heat treatment technology is considered for pipelines intended for the NPP cooling systems employing sodium coolant. Various techniques are discussed which are used for protecting the pipeline internal surfaces against oxidation in the process of heat treatment. It is noted that the austenite formation of welded joints of steel 12Kh18N9 and steel Kh16N11M3 at temperatures of 1050 and 1100 deg C releases welding-induced stresses and reduces a possibility of local damages. Evacuation down to 1 mm Hg appears to be the most rational protective technique. The considered procedure of the pipeline heat treatment has been utilized for mounting the equipment of the BN-600 reactor at the Beloyarskaya NPP. The economic gain resulting from the use of the procedure, owing to decrease in argon consumption and reduction of labour input, makes up 150 000 roubles

CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

Energy Technology Data Exchange (ETDEWEB)

Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

2015-09-02

The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

Metallic sodium as a coolant of high speed nuclear reactors, (2)

International Nuclear Information System (INIS)

Atsumo, Hideo

1975-01-01

Tables are given on all the sodium loops in Japan and most of the sodium loops all over the world. Name and purpose of the loops, time of establishment, highest temperature, amount of sodium, flow rate, the materials used for the construction of the loops, and the diameter of the main pipings are given. Also, the problems related with these loops are discussed. For example, the high temperature sodium facility at HEDL-WADCO was made for the FFTF component test and instrument test, and uses 50,000 gallons of metallic sodium. The highest temperature is 590 0 C. The sodium flows at the rate of 60 g/m. The body is made of Type 304 stainless steel. Main data of existing sodium-cooled reactors in the world are also tabulated. The data include thermal output, electric output, the structure of the reactor cores, the dimensions of the cores, fuel used, the highest temperature in the reactors, the temperature of sodium at the inlet and outlet, the rate of multiplication, the amount of sodium used, number of control rods, number of heat exchangers, and the pressure of steam. The Monju type nuclear reactor in Japan uses 1,800 ton of sodium. (Fukutomi, T.)

Evaluation of conservatism in analysis of fuel-coolant interaction

International Nuclear Information System (INIS)

Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Haas, P.M.; Allen, C.L.

Using the ANL parametric model developed by Cho e.a. the following mechanisms and parameters involved in fuel-coolant interaction were examined: coherence of fuel-sodium mixing; two-phase heat transfer; sodium-to-fuel mass ratio; fuel particle size; heat transfer to plenum and core cladding; constraint geometry. Both overpower and loss-of-flow transients were studied. Main attention is given to the maximum mechanical work to be expected. As a general conclusion, it can be stated that more realistic models will result in a reduction of the estimated mechanical work

Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

2017-02-15

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

Comparison of thermohydraulic characteristics in the use of various coolants

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

2000-11-01

Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet

Determination of primary flow by correlation of temperatures of the coolant

International Nuclear Information System (INIS)

Villanueva, Jose

2003-01-01

Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

Energy Technology Data Exchange (ETDEWEB)

Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

2012-07-01

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

International Nuclear Information System (INIS)

Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

2016-01-01

Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

International Nuclear Information System (INIS)

Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

2013-01-01

The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

FBR structural material test facility in flowing sodium environment

International Nuclear Information System (INIS)

Shanmugasundaram, M.; Kumar, Hemant; Ravi, S.

2016-01-01

In Fast Breeder Reactor (FBR), components such as Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer arm and primary sodium pumps etc., are experiencing friction and wear between the moving parts in contact with liquid sodium at high temperature. Hence, it is essential to evaluate the friction and wear behaviour to validate the design of components. In addition, the above core structural reactor components such as core cover plate, control plugs etc., undergoes thermal striping which is random thermal cycling induced by flow stream resulting from the mixing of non isothermal jets near that component. This leads to development of surface cracks and assist in crack growth which in turn may lead to failure of the structural component. Further, high temperature components are often subjected to low cycle fatigue due to temperature gradient induced cyclic thermal stresses caused by start-ups, shutdowns and transients. Also steady state operation at elevated temperature introduces creep and the combination of creep and fatigue leads to creep-fatigue interactions. Therefore, resistance to low cycle fatigue, creep and creep-fatigue are important considerations in the design of FBR components. Liquid sodium is used as coolant and hence the study of the above properties in dynamic sodium are equally important. In view of the above, facility for materials testing in sodium (INSOT) has been constructed and in operation for conducting the experiments such as tribology, thermal stripping, low cycle fatigue, creep and creep-fatigue interaction etc. The salient features of the operation and maintenance of creep and fatigue loops of INSOT facility are discussed in detail. (author)

Cavitation erosion in sodium flow, sodium cavitation tunnel testing

International Nuclear Information System (INIS)

Courbiere, Pierre.

1981-04-01

The high-volume sodium flows present in fast neutron reactors are liable to induce cavitation phenomena in various portion of the sodium lines and pumps. The absence of sufficient data in this area led the C.E.A. to undertake an erosion research program in cavitating sodium flow. This paper discusses the considerations leading to the definition and execution of sodium cavitation erosion tests, and reviews the tests run with 400 0 C sodium on various steel grades: 316, 316 L, 316 Ti (Z8CNDT17-12), Poral (Z3CND18-12), 304 L and LN2 - clad 316 L (Ni coating-clad 316 L). Acoustic detection and signal processing methods were used with an instrument package designed and implemented at the Cadarache Nuclear Research Center

Evaluation of effective coolant flow rate in advanced design of the small scale VHTR core

International Nuclear Information System (INIS)

Fumizawa, Motoo; Suzuki, Kunihiko; Murakami, Tomoyuki.

1988-02-01

This report describes the evaluation of effective coolant flow rate in the advanced design of the small scale VHTR core. The analytical design study was carried out after the 2nd stage of detailed design in order to reduce the cost of construction. The summary of the analytical results are as follows: (1) Crossflow loss coefficient of flange type fuel block having 0.1 mm of sealing gap is about 100 times higher than that of dowel type block adopted in the 2nd stage of detailed design. (2) In case that coolant channel outer diameter is 52 mm and hydraulic diameter is 6 mm, the effective coolant flow rates using flange and dowel type fuel blocks are 80 % and 70 % respectively. Because the crossflow loss coefficients of dowel type are lower than that of flange type. (3) The effective coolant flow rate, when crossflow loss coefficients are distributed along with the axial direction, agrees well with that using mean value of crossflow loss coefficient i.e. 5 x 10 11 m -4 . (author)

Design of the solid target structure and the study on the coolant flow distribution in the solid target using the 2-dimensional flow analysis

International Nuclear Information System (INIS)

Haga, Katsuhiro; Terada, Atsuhiko; Ishikura, Shuichi; Teshigawara, Makoto; Kinoshita, Hidetaka; Kobayashi, Kaoru; Kaminaga, Masaki; Hino, Ryutaro; Susuki, Akira

1999-11-01

A solid target cooled by heavy water is presently under development under the Neutron Science Research Project of the Japan Atomic Energy Research Institute (JAERI). Target plates of several millimeters thickness made of heavy metal are used as the spallation target material and they are put face to face in a row with one to two millimeters gaps in between though which heavy water flows, as the coolant. Based on the design criteria regarding the target plate cooling, the volume percentage of the coolant, and the thermal stress produced in the target plates, we conducted thermal and hydraulic analysis with a one dimensional target plate model. We choosed tungsten as the target material, and decided on various target plate thicknesses. We then calculated the temperature and the thermal stress in the target plates using a two dimensional model, and confirmed the validity of the target plate thicknesses. Based on these analytical results, we proposed a target structure in which forty target plates are divided into six groups and each group is cooled using a single pass of coolant. In order to investigate the relationship between the distribution of the coolant flow, the pressure drop, and the coolant velocity, we conducted a hydraulic analysis using the general purpose hydraulic analysis code. As a result, we realized that an uniform coolant flow distribution can be achieved under a wide range of flow velocity conditions in the target plate cooling channels from 1 m/s to 10 m/s. The pressure drop along the coolant path was 0.09 MPa and 0.17 MPa when the coolant flow velocity was 5 m/s and 7 m/s respectively, which is required to cool the 1.5 MW and 2.5 MW solid targets. (author)

Evaluation of organic coolants for the transportation of LMFBR spent fuel rods

International Nuclear Information System (INIS)

Arnold, C. Jr.

1978-05-01

The physical and chemical processes that are likely to occur when sodium coated LMFBR spent fuel rods are submerged in various aromatic organic coolants was defined by means of immersion experiments carried out with sodium coated 304 stainless steel coupons. Upon immersion of sodium coated coupons at 220 0 C in hydrocarbon type coolants such as Therminol 88, a mixture of terphenyls, not only was the metallic sodium retained on the coupon, but a carbonaceous coating formed on the surface of the sodium. In contrast, coolants that contained aromatic ether bonds, such as Dowtherm A, reacted with sodium at 220 0 C to form phenolate and other salts, which precipitated from the coolant in the form of a dark sludge. With Dowtherm A, removal of metallic sodium from the coupon was essentially complete in a matter of hours at temperatures of 160--220 0 C. Data on the rate and efficiency of sodium removal upon immersion in Dowtherm A at elevated temperatures were obtained. In addition the kinetics and chemistry of the sodium/Dowtherm A reaction were defined. Because sodium sludges are potentially incompatible with the containing structural materials and the fuel elements, it is recommended that sodium be removed prior to immersion in the coolant via reaction with benzoic acid; this method should be adaptable to the facilities at reactor sites. In aging studies Dowtherm A was found to be thermally stable up to 400 0 C and radiatively stable at ambient conditions. The combined effect of heat and radiation was not defined

Analysis of coolant flow in central tube of WWER-440 fuel assemblies

International Nuclear Information System (INIS)

Zsiros, G.; Toth, S.; Attila Aszodi, A.

2011-01-01

Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

System and method for determining coolant level and flow velocity in a nuclear reactor

Science.gov (United States)

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

Directory of Open Access Journals (Sweden)

Guangchao Li

2012-12-01

Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

The problems of using a high-temperature sodium coolant in nuclear power plants for the production of hydrogen and other innovative applications

Science.gov (United States)

Sorokin, A. P.; Alexeev, V. V.; Kuzina, Ju. A.; Konovalov, M. A.

2017-11-01

The intensity of the hydrogen sources arriving from the third contour of installation in second in comparison with the hydrogen sources on NPP BN-600 increases by two - three order at using of high-temperature nuclear power plants with the sodium coolant (HT-NPP) for drawing of hydrogen and other innovative applications (gasification and a liquefaction of coal, profound oil refining, transformation of biomass to liquid fuel, in the chemical industry, metallurgy, the food-processing industry etc.). For these conditions basic new technological solutions are offered. The main condition of their implementation is raise of hydrogen concentration in the sodium coolant on two - three order in comparison with the modern NPP, in a combination to hydrogen removal from sodium and its pumping out through membranes from vanadium or niobium. The researches with use diffusive model have shown possibility to expel a casium inflow in sodium through a leakproof shell of fuel rods if vary such parameters as a material of fuel rods shell, its thickness and maintenance time at design of fuel rods for high-temperature NPP. However maintenance of high-temperature NPP in the presence of casium in sodium is inevitable at loss of leakproof of a fuel rods shell. In these conditions for minimisation of casium diffusion in structural materials it is necessary to provide deep clearing of sodium from cesium.

An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

International Nuclear Information System (INIS)

Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

2014-01-01

The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

Health physics aspects of processing EBR-I coolant

International Nuclear Information System (INIS)

Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

1998-01-01

The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed

Sodium aerosol formation in an argon flow over hot sodium

International Nuclear Information System (INIS)

Clement, C.F.; Dolias, M.J.; UKAEA Atomic Energy Research Establishment, Harwell. Thermal Hydraulics Div.)

1987-01-01

Vapour evaporation, which partly forms aerosol, occurs when a cold gas flows over a hot liquid. A previous well-mixed model is extended to predict the final vapour plus aerosol content of such a flow in terms of its initial and final temperatures. The predictions are compared to results of the Copacabana II experiment in which argon passed over a sodium pool. Agreement is obtained for the final sodium density at moderate flow rates, and physical reasons are given as to why deviations occur at low and high flow rates. (author)

Core performance of equilibrium fast reactors for different coolant materials and fuel types

International Nuclear Information System (INIS)

Mizutani, Akihiko; Sekimoto, Hiroshi

1998-01-01

Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

International Nuclear Information System (INIS)

Nishimura, Satoshi; Ueda, Nobuyuki

2007-01-01

Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

International Nuclear Information System (INIS)

Chenu, A.

2011-10-01

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

Sodium fires and its extinguishment

International Nuclear Information System (INIS)

Mikhedov, V.G.

1979-01-01

The fire safety problems of NPP with sodium coolants in USSR are presented. The design of sodium reactors is made with premises with sodium coolants being hermetic and filled with nitrogen. Some engineering solutions of fire safety including design, elaboration and choice of construction and protection materials are presented. Some theoretical aspects of sodium burning are presented as well as methods of sodium fire extinguishing methods including the use of powder

Ultrasonic flow-meter test in sodium

International Nuclear Information System (INIS)

Ishii, Y.; Uno, O.; Kamei, M.

1978-01-01

As a part of the R and D programme for the prototype fast breeder reactor MONJU, an ultrasonic flow-meter (USFM) test is being carried out in sodium in the O-Arai Engineering Center of PNC. Prior to the present test, an in-water test was done at the manufacturer's as a preliminary investigation. The results reported here are the results up to the present. Calibration tests using the actual fluid were conducted on a 12-inch ultrasonic flow-meter with guide rods fabricated for sodium flow measurement. The test conditions in sodium were a temperature of 200 approximately 400 0 C and flow-rates of 0 approximately 6m/s. The main results are: (1) The linearity of output signal was good and accuracy was within 1%; (2) The alternating type of the USFM was much better than the fixed type in temperature change; (3) 2MHz of transducer frequency was better than 3MHz in sodium; (4) The S/N ratio of the ultrasonic signal and the length/diameter effect in a wide range in sodium surpassed the in-water test. (author)

Full sized tests on a french coolant pump under two-phase flow

International Nuclear Information System (INIS)

Huchard, J.C.; Bore, C.; Dueymes, E.

1997-01-01

The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

USSR position paper for the specialists' meeting on thermal stratification in sodium

International Nuclear Information System (INIS)

Rinejskij, A.A.; Shcherbakov, S.I.

1983-01-01

The problem of adequate coolant mixing and flow pattern optimisation becomes more significant while considering pool type fast reactors. Within a vessel of this reactor type there are considerable bulks of coolant at different temperatures. Because of the complex design requirements certain difficulties exist in the problem of ensuring sufficient mixing of the coolant at the outlet of the core, radial blanket, in-pile fuel storage and in-pile neutron shielding. n the framework of the R and D activities carried out to justify fast reactor design solutions a proper attention is given to investigations of the coolant flow pattern within plena and in those sections of pool type reactor, where the coolant now takes place. In this connection It should be noted that, up to now, we did not obtain any clear evidence of the presence of a pronounced thermal stratification of sodium within reactor mixing chambers under normal operating conditions. The conclusion may be drawn on the basis of our experiments and design experience that there are certain gradients of temperature and Inefficient mixing of coolant flows with different temperatures within the chambers (both 'hot' and 'cold' ones). All these factors to a certain degree complicate a reactor operation. However, from our point of view, the main point in the problem of arranging an 'optimum' coolant flow pattern within the chambers is to insure the stability of coolant motion in the chambers (Including flows with different temperatures) by means ol special design solutions based on results of experimental and calculation studies. It is considered that the instability of coolant flow could be the most important factor influencing temperature distribution within structures and components. That is why coolant flow in all reactor chambers and cavities should be stable, including natural convection flows

Assessment of flow induced vibration in a sodium-sodium heat exchanger

Energy Technology Data Exchange (ETDEWEB)

Prakash, V. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)], E-mail: prakash@igcar.gov.in; Thirumalai, M.; Prabhakar, R.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

2009-01-15

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60 deg. sector model. This paper discusses the flow induced vibration measurements carried out in 60 deg. sector model of IHX, the modeling criteria, the results and conclusion.

Dryout heat flux in a debris bed with forced coolant flow from below

International Nuclear Information System (INIS)

Bang, Kwang-Hyun; Kim, Jong-Myung

2004-01-01

The objective of the present study is to experimentally investigate the enhancement of dryout heat flux in debris beds with coolant flow from below. The experimental facility consists mainly of an induction heater (40 kW, 35 kHz), a double-wall quartz-tube test section containing steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of particle bed was achieved by induction heating. This paper reports the experimental data for 5 mm particle bed and 300 mm bed height. The dryout heat rate data were obtained of both top-flooding case and forced coolant injection from below with the injection mass flux up to 1.5 kg/m 2 s. For the top-flooded case, the volumetric dryout heat rate was about 4 MW/m 3 and it increased as the rate of coolant injection from below was increased. At the coolant injection mass flux of 1.5 kg/m 2 s, the volumetric dryout heat rate was about 10 MW/m 3 , the enhancement factor was more than two. (author)

Comparative analysis of coolants for FBR of future nuclear power

International Nuclear Information System (INIS)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

2001-01-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

Comparative analysis of coolants for FBR of future nuclear power

Energy Technology Data Exchange (ETDEWEB)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

2001-07-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

Multirods burst tests under loss-of-coolant conditions

International Nuclear Information System (INIS)

Kawasaki, S.; Uetsuka, H.; Furuta, T.

1983-01-01

In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

Chemistry of liquid metal coolants and sensors

International Nuclear Information System (INIS)

Gnanasekaran, T.

2015-01-01

Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

SIMMER-III applications to fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1998-01-01

The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

Development of Sodium Two Phase Flow Model for Kalimer Core Analysis

International Nuclear Information System (INIS)

Chang, W.P.; Hahn, Dohee

2002-01-01

An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid Metal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examined the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations. (authors)

Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

Science.gov (United States)

Xie, Huaqing; Li, Yang; Yu, Wei

2010-05-01

We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2O 3, ZnO, TiO 2, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2O 3, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

International Nuclear Information System (INIS)

Xie Huaqing; Li Yang; Yu Wei

2010-01-01

We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2 O 3 , ZnO, TiO 2 , and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2 O 3 , and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

Energy Technology Data Exchange (ETDEWEB)

Xie Huaqing, E-mail: hqxie@eed.sspu.c [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China); Li Yang; Yu Wei [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China)

2010-05-31

We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al{sub 2}O{sub 3}, ZnO, TiO{sub 2}, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al{sub 2}O{sub 3}, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

Use of flow models to analyse loss of coolant accidents

International Nuclear Information System (INIS)

Pinet, Bernard

1978-01-01

This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

A numerical approach to the simulation of one-phase and two phase reactor coolant flow around nuclear fuel spacers

International Nuclear Information System (INIS)

Stosic, Z.V.; Stevanovic, V.D.

2001-01-01

A methodology for the simulation and analysis of one-phase and two-phase coolant flows around one or a row of spacers is presented. It is based on the multidimensional two-fluid mass, momentum and energy balance equations and application of adequate turbulence models. Necessary closure laws for interfacial transfer processes are presented. The stated general approach enables simulation and analyses of reactor coolant flow around spacers on different scale levels of the rod bundle geometry: detailed modelling of coolant flow around spacers and investigation of the influence of spacer's geometry on the coolant thermal-hydraulics, as well as prediction of global thermal-hydraulic parameters within the whole rod bundle with the investigation of the influence of rows of spacers on the bulk thermal-hydraulic processes. Sample problems are included illustrating these different modelling approaches. (author)

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Science.gov (United States)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

Sodium flow distribution test of the air cooler tubes

International Nuclear Information System (INIS)

Uchida, Hiroyuki; Ohta, Hidehisa; Shimazu, Hisashi

1980-01-01

In the heat transfer tubes of the air cooler which is installed in the auxiliary core cooling system of the fast breeder prototype plant reactor ''Monju'', sodium freezing may be caused by undercooling the sodium induced by an extremely unbalanced sodium flow in the tubes. Thus, the sodium flow distribution test of the air cooler tubes was performed to examine the flow distribution of the tubes and to estimate the possibility of sodium freezing in the tubes. This test was performed by using a one fourth air cooler model installed in the water flow test facility. As the test results show, the flow distribution from the inlet header to each tube is almost equal at any operating condition, that is, the velocity deviation from normalized mean velocity is less than 6% and sodium freezing does not occur up to 250% air velocity deviation at stand-by condition. It was clear that the proposed air cooler design for the ''Monju'' will have a good sodium flow distribution at any operating condition. (author)

Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

Energy Technology Data Exchange (ETDEWEB)

Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

Numerical Simulation of Non-Rotating and Rotating Coolant Channel Flow Fields. Part 1

Science.gov (United States)

Rigby, David L.

2000-01-01

Future generations of ultra high bypass-ratio jet engines will require far higher pressure ratios and operating temperatures than those of current engines. For the foreseeable future, engine materials will not be able to withstand the high temperatures without some form of cooling. In particular the turbine blades, which are under high thermal as well as mechanical loads, must be cooled. Cooling of turbine blades is achieved by bleeding air from the compressor stage of the engine through complicated internal passages in the turbine blades (internal cooling, including jet-impingement cooling) and by bleeding small amounts of air into the boundary layer of the external flow through small discrete holes on the surface of the blade (film cooling and transpiration cooling). The cooling must be done using a minimum amount of air or any increases in efficiency gained through higher operating temperature will be lost due to added load on the compressor stage. Turbine cooling schemes have traditionally been based on extensive empirical data bases, quasi-one-dimensional computational fluid dynamics (CFD) analysis, and trial and error. With improved capabilities of CFD, these traditional methods can be augmented by full three-dimensional simulations of the coolant flow to predict in detail the heat transfer and metal temperatures. Several aspects of turbine coolant flows make such application of CFD difficult, thus a highly effective CFD methodology must be used. First, high resolution of the flow field is required to attain the needed accuracy for heat transfer predictions, making highly efficient flow solvers essential for such computations. Second, the geometries of the flow passages are complicated but must be modeled accurately in order to capture all important details of the flow. This makes grid generation and grid quality important issues. Finally, since coolant flows are turbulent and separated the effects of turbulence must be modeled with a low Reynolds number

The state-of-the-art report for flow and pressure measurement techniques in the piping system

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho; Lee, Jeong Han; Kim, Bong Soo; Kim, Jong Man; Choi, Jong Hyun

2002-12-01

The sodium is widely used as one of the coolants in the liquid metal reactor system since it has important safety features such as a long thermal response time, a large margin to coolant boiling, and operating in near atmospheric pressure, etc. The state-of-the-art on the flow and pressure measurement techniques in the piping system worldwide is investigated and reviewed to utilize it as a basis for developing a new technique applying for the high temperature sodium flow environment.

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

Optimization of mass flow rate in RGTT200K coolant purification for Carbon Monoxide conversion process

International Nuclear Information System (INIS)

Sumijanto; Sriyono

2016-01-01

Carbon monoxide is a species that is difficult to be separated from the reactor coolant helium because it has a relatively small molecular size. So it needs a process of conversion from carbon monoxide to carbondioxide. The rate of conversion of carbon monoxide in the purification system is influenced by several parameters including concentration, temperature and mass flow rate. In this research, optimization of the mass flow rate in coolant purification of RGTT200K for carbon monoxide conversion process was done. Optimization is carried out by using software Super Pro Designer. The rate of reduction of reactant species, the growth rate between the species and the species products in the conversion reactions equilibrium were analyzed to derive the mass flow rate optimization of purification for carbon monoxide conversion process. The purpose of this study is to find the mass flow rate of purification for the preparation of the basic design of the RGTT200K coolant helium purification system. The analysis showed that the helium mass flow rate of 0.6 kg/second resulted in an un optimal conversion process. The optimal conversion process was reached at a mass flow rate of 1.2 kg/second. A flow rate of 3.6 kg/second – 12 kg/second resulted in an ineffective process. For supporting the basic design of the RGTT200K helium purification system, the mass flow rate for carbon monoxide conversion process is suggested to be 1.2 kg/second. (author)

Evaluation of the corrosion, reactivity and chemistry control aspects for the selection of an alternative coolant in the secondary circuit of sodium fast reactors

International Nuclear Information System (INIS)

Brissonneau, L.; Simon, N.; Balbaud-Celerier, F.; Courouau, J.L.; Martinelli, L.; Grabon, V.; Capitaine, A.; Conocar, O.; Blat, M.

2009-01-01

Full text of publication follows: Sodium Fast Reactors are promising fourth generation reactors as they can contribute to reduce resource demand in uranium and considerably reduce waste level due to their fast spectrum. However, progress can be obtained for these reactors on the investment cost and on safety improvement. To achieve these goals, one of the innovative solutions consists in eliminating the reaction of sodium with water in the steam generators, by replacing the sodium in the secondary circuit by another coolant. A work group composed of experts from CEA, Areva NP and EdF was in charge to evaluate several alternative coolants as Heavy Liquid Metals (HLM), nitrate salts and hydroxide mixtures, through a multi-criteria analysis. Three important criteria for the selection of one coolant are its 'Interactions with the structures', and its 'chemistry control', and 'Reactivity with fluids' which are strongly correlated. The assessment, mainly based on the state-of-art from published literature on these points, is detailed in this paper. The mechanisms of corrosion of steels by the HLM depend on the oxygen content. For Pb-Bi, it has been modelled for oxidation and release domains. The corrosion of steels by nitrate salts presents similarity with the oxidation induced by HLM. The highly corrosive hydroxide mixture requires the use of nickel base alloys, for which oxidation and mass transfer are nevertheless significant. The HLM requires a fine regulation of oxygen content, through measurements and control systems, both to prevent lead oxide precipitation at high level and release corrosion at low level. Nitrate salts decompose into nitrites at sufficiently high temperature, which might induce pressure build-up in the circuit. The hydroxides must be kept under reducing atmosphere to lower the corrosion rate. Though these coolants are relatively inert to air and water, one of the main drawbacks of HLM and nitrate salts are their reactivity with sodium. Bismuth

Analysis of Sodium-23 Data Cross-Sections for Coolant on Generation IV Reactor - SFR

International Nuclear Information System (INIS)

Suwoto; Zuhair

2009-01-01

The integral tests of sodium-23 neutron cross-sections for coolant contained in JENDL-3.3, ENDF/B-VII.0, BROND-2.2 and JEFF-3.1 files have been performed. Cross-sections analysis of sodium-23 such as total cross-sections, elastic scattering, in-elastic scattering and radiative capture cross-sections for several temperature i.e. 300K, 800K and 1500K. The sodium-23 total cross-sections analysis based on MAEKER, R.E. experimental result through Broomstick experiment calculation. Differences between among other evaluated nuclear data file for elastic scattering, in-elastic scattering and radiative capture cross-sections were done analyzed and compared to ENDF/B-VII.0 as standard reference. Analysis of total cross-sections sodium-23 through broomstick calculation show JENDL-3.3 file give the best result on C/E statistical average value is 1.1043 compared to another nuclear data files. Differences sodium-23 total cross-sections on JEFF-3.1 file for all temperature work specially for energy 40keV and 1MeV-2MeV is about 0.2%. Meanwhile, relative small differences on in-elastic total scattering cross-sections are shown for all temperatures are about ±0.1% in JENDL-3.3. On the other hand, BROND-2.2 file give ±6% higher on sodium-23 in-elastic total scattering cross sections for energy range 450keV-550keV. Clearly significant differences on sodium-23 radiative capture cross sections for BROND-2.2 file especially in energy 109.659keV is somewhat higher than 446%, otherwise JENDL-3.3 and JEFF-3.1 give 16% higher than ENDF/B-VII.0 file. Overall result show that JENDL-3.3, ENDFB-VII.0, BROND-2.2 and JEFF-3.1 have little bit differences in total, elastic scattering, in-elastic scattering total cross sections, except BROND-2.2 file due to radiative capture cross-sections with larger discrepancies. (author)

Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

International Nuclear Information System (INIS)

Pradal, L.; Berte, M.; Chiarelli, C.

1985-01-01

The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

A thermal analysis computer programme package for the estimation of KANUPP coolant channel flows and outlet header temperature distribution

International Nuclear Information System (INIS)

Siddiqui, M.S.

1992-06-01

COFTAN is a computer code for actual estimation of flows and temperatures in the coolant channels of a pressure tube heavy water reactor. The code is being used for Candu type reactor with coolant flowing 208 channels. The simulation model first performs the detailed calculation of flux and power distribution based on two groups diffusion theory treatment on a three dimensional mesh and then channel powers, resulting from the summation of eleven bundle powers in each of the 208 channels, are employed to make actual estimation of coolant flows using channel powers and channel outlet temperature monitored by digital computers. The code by using the design flows in individual channels and applying a correction factor based on control room monitored flows in eight selected channels, can also provide a reserve computational tool of estimating individual channel outlet temperatures, thus providing an alternate arrangements for checking Rads performance. 42 figs. (Orig./A.B.)

Development of two-phase Flow Model, 'SOBOIL', for Sodium

International Nuclear Information System (INIS)

Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

2000-03-01

The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification

Influence of flowing sodium on the creep rupture behavior of type 304 SS at 550 degrees C

Energy Technology Data Exchange (ETDEWEB)

Huthmann, H; Grosser, E D [INTERATOM, Bergisch Gladbach (Germany); Tas, H [SCK/CEN, Mol (Belgium); Borgstedt, H U [GfK, Karlsruhe (Germany)

1977-07-01

Substantial components of Liquid Metal Fast Breeder Reactors operate in a temperature range, where the knowledge of time dependent properties is necessary for design purposes. Current high temperature design rules for this temperature range are based on material properties data, which have been obtained in air tests. During the exposure to the flowing sodium environment at elevated temperatures the structural materials are subjected to compositional and microstructural changes, which could alter the mechanical behavior as compared to that in air tests. Within the SNR 300 project a cooperative program between INTERATOM (Federal Republic of Germany), the Nuclear Research Center, GfK, of Karlsruhe and the Studiecentrum voor Kernenergie, SCK/CEN, in Belgium Is conducted to determine quantitatively the influence of sodium environment on the creep-rupture behavior of Type 304 stainless steel. This program is aimed to identify design properties reflecting the possible influence of the coolant. In the present paper, preliminary data are presented on the effect of dynamic un-isothermal sodium on the uniaxial creep-rupture behavior of Type 304 stainless steel at 550 deg. C.

Coolant leakage detection device

International Nuclear Information System (INIS)

Ito, Takao.

1983-01-01

Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

Effect of coolant flow rate on the power at onset of nucleate boiling in a swimming pool type research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Ahmad, N.; Ahmad, S.

1998-01-01

The effect of flow rate of coolant on power of Onset Nucleate Boiling (ONB) in a reference core of a swimming pool type research reactor has been studied using a as standard computer code PARET. It has been found that the decrease in the coolant flow rate results in a corresponding decrease in power at ONB. (author)

A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

International Nuclear Information System (INIS)

Markfort, D.; Kaiser, A.; Dohmen, A.

1975-01-01

In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

International Nuclear Information System (INIS)

Hunsbedt, A.; Boardman, C.E.

1993-01-01

A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

International Nuclear Information System (INIS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

2007-01-01

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

International Nuclear Information System (INIS)

Royl, P.

1985-01-01

The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

Sodium evaporation into a forced argon flow

International Nuclear Information System (INIS)

Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

1975-01-01

Evaporation from a rectangular sodium free surface into an argon flow was measured. Tests were carried out with varying sodium temperature, argon velocity and argon temperature respectively under conditions of fog formation being possible. In order to clarify the enhancement of evaporation by fog formation, convection heat transfer from a plate of the same geometry into an air flow was also measured. The evaporation rate and Sherwood number were compared with those predicted by both the heat transfer experiment and the theory proposed by Hill and Szekely, and also a comparison was run with the previously reported experimental results of sodium evaporation. As a result it was shown that the sodium evaporation rate in this experiment is at least four times as large as that predicted by the heat transfer experiment and varies almost linearly with the heat transfer rate and the sodium vapour pressure. (auth.)

Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Knight, D.D.

1981-01-01

The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

Numerical Analysis of Flow Distribution in a Sodium Chamber of a Finned-tube Sodium-to-Air Heat Exchanger

Energy Technology Data Exchange (ETDEWEB)

Jo, Youngchul; Son, Seokkwon; Kim, Hyungmo; Eoh, Jaehyuk; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

DHR systems consist of two diverse heat removal loops such as passive and active DHR systems, and the heat load imposed on the primary sodium pool is safely rejected into the environment through different kinds of sodium-to-air heat exchangers, e.g. M-shape and helical-coil type air-coolers. The former is called as an FHX(Forced-draft sodium-to-air Heat Exchanger) and the latter is simply called as an AHX(natural-draft sodium-to-Air Heat Exchanger). In a general sodium-to-air heat exchanger design, convection resistance in a shell-side air flow path becomes dominant factor affecting the mechanism of conjugate heat transfer from the sodium flow inside the tube to the air path across the sodium tube wall. Hence verification of the flow and heat transfer characteristics is one of the most important tasks to demonstrate decay heat removal performance. To confirm a kind of ultimate heat sink heat exchanger, a medium-scale Sodium thermal-hydraulic Experiment Loop for Finned-tube sodium-to-Air Heat exchanger (here after called the SELFA) has been designed and is recently being constructed at KAERI site. The introduction of the flow baffle inside the upper sodium chamber of the model FHX unit in the SELFA facility is briefly proposed and discussed as well. The present study aims at introducing a flow baffle design inside the upper sodium chamber to make more equalized flowrates flowing into each heat transfer tube of the model FHX unit. In the cases without the flow baffle geometry, it was observed lager discrepancies in flowrates at the heat transfer tubes. However it was also found that those kinds of discrepancies could be definitely decreased at around 1/10 by employing a flow baffle.

Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

Energy Technology Data Exchange (ETDEWEB)

Supa-Amornkul, S

2006-07-01

In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73{sup o} bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m{sup 3}/s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code

Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

International Nuclear Information System (INIS)

Supa-Amornkul, S.

2006-01-01

In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73 o bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m 3 /s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code - Fluent 6.1- had

Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

Directory of Open Access Journals (Sweden)

Hutli Ezddin

2015-01-01

Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

International Nuclear Information System (INIS)

Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

2013-06-01

The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

Sudden contact of a hot liquid with a volatile coolant: instability of the created vapour film

International Nuclear Information System (INIS)

Pion, Agnes

1983-01-01

As the sudden contact of a hot body with a coolant which may evaporate, results, after some delay, in an explosive evaporation, this research thesis proposes an interpretation based on the study of the destabilization of the vapour film which forms at the surface of the hot body. The author reports the modelling of the evolution of the average thickness of the film before the explosion. The possible chemical reactions at the surface of the hot body are taken into account. A base flow is obtained which allows the calculation of the evolution of Rayleigh-Taylor instabilities which may occur at the gas-coolant interface. This study is applied to the interaction between liquid sodium and water [fr

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

Study of coolant flow distribution within the PWR type reactor vessel

International Nuclear Information System (INIS)

Eberle, L.M.M.

1983-01-01

The thermohydraulic design of a pressurized water reactor requires the determination of the coolant flow distributions within the reactor vessel, particulary at the core inlet. In this work it is proposed the study of this flow, using potencial flow theory governed by Laplace's equation, nabla 2 φ = O. The solution of the potential field is obtained by the finite element method, which simplifies considerably the treatment of complex geometrical configurations. The equation is solved by the finite element computer code ANSYS, developed and licensed for structural and thermal analysis by using the analogy between steady state heat transfer equation without heat generation, nabla 2 T=O, and Laplace's equation of the velocity potential. The proposed method has been applied to a commercial reactor, and the results are consistent with the available experimental data. (author) [pt

Assessment of RANS at low Prandtl number and simulation of sodium boiling flows with a CMFD code

Energy Technology Data Exchange (ETDEWEB)

Mimouni, S., E-mail: stephane.mimouni@edf.fr; Guingo, M.; Lavieville, J.

2017-02-15

Highlights: • Modelling of boiling sodium flows in a multiphase flow solver. • Rod heated with a constant heat flux in a pipe liquid metal flow. • Sodium boiling flow around a rod heated with a constant heat. • Computations in progress in an assembly constituted of 19 pins equipped with a wrapped wire. - Abstract: In France, Sodium-cooled Fast Reactors (SFR) have recently received a renewed interest. In 2006, the decision was taken by the French Government to initiate research in order to build a first Generation IV prototype (called ASTRID) by 2020. The improvement in the safety of SFR is one of the key points in their conception. Accidental sequences may lead to a significant increase of reactivity. This is for instance the case when the sodium coolant is boiling within the fissile zone. As a consequence, incipient boiling superheat of sodium is an important parameter, as it can influence boiling process which may appear during some postulated accidents as the unexpected loss of flow (ULOF). The problem is that despite the reduction in core power, when boiling conditions are reached, the flow decreases progressively and vapour expands into the heating zone. A crucial investigating way is to optimize the design of the fissile assemblies of the core in order to lead to stable boiling during a ULOF accident, without voiding of the fissile zone. Moreover, in order to evaluate nuclear plant design and safety, a CFD tool has been developed at EDF in the framework of the nuclear industry. Advanced models dedicated to boiling flows have been implemented and validated against experimental data for ten years now including a wall law for boiling flows, wall transfer for nucleate boiling, turbulence and polydispersion model. This paper aims at evaluating the generalization of these models to SFR. At least two main issues are encountered. Firstly, at low Prandtl numbers such as those of liquid metal, classical approaches derived for unity or close to unity fail to

Control of sodium fires and sodium-water reactions in breeder reactors

International Nuclear Information System (INIS)

Foerster, K.; Ruloff, G.; Voss, J.

1985-01-01

The excellent neutronic and thermodynamic properties of sodium as a fast-reactor coolant are somewhat counterbalanced by its high oxygen affinity. Because incidents like sodium fires and sodium-water reactions cannot be absolutely excluded, their effects and preventive measures have to be investigated. Characteristics and counter-measures are discussed. (orig.) [de

Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

International Nuclear Information System (INIS)

Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

2016-09-01

The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

International Nuclear Information System (INIS)

In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

2008-01-01

Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

Measurement and analysis of vibrational behavior of an SNR-fuel element in sodium flow

International Nuclear Information System (INIS)

Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

1975-01-01

Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 deg C system temperature in the AKB sodium loop at Interatom, Bensberg. It is known that the coolant flow through a subassembly can induce flutter or vibrations of structural parts such as single pins, the wrapper and the total pin bundle all of which have been of interest during this test. To detect these vibrations of different structural parts simultaneously with a minimum of instrumentation only 3 weldable high temperature strain gauges were employed. These strain gauges were especially prepared and bent in such a way as to form a bridge between the inner wrapper and a fuel pin top and spot-welded to both the wrapper and the fuel pin. Although this arrangement seems to be a rather unusual one, the simultaneous-measurement of bundle, wrapper and pin vibrations was possible and periodic flow fluctuations were also detected. The presented results are only relative due to calibration difficulties with these deformed strain gauges which were first used during this test. It is, however, believed that this arrangement, in connection with the proposed anlytical approach, leads to a simple and technical representation of the vibrational behavior of core elements during sodium tests. Detailed information needed for check and calibration of computer codes are however displayed by the respective power spectral density functions

The relationship between dynamic and average flow rates of the coolant in the channels of complex shape

Science.gov (United States)

Fedoseev, V. N.; Pisarevsky, M. I.; Balberkina, Y. N.

2018-01-01

This paper presents interconnection of dynamic and average flow rates of the coolant in a channel of complex geometry that is a basis for a generalization model of experimental data on heat transfer in various porous structures. Formulas for calculation of heat transfer of fuel rods in transversal fluid flow are acquired with the use of the abovementioned model. It is shown that the model describes a marginal case of separated flows in twisting channels where coolant constantly changes its flow direction and mixes in the communicating channels with large intensity. Dynamic speed is suggested to be identified by power for pumping. The coefficient of proportionality in general case depends on the geometry of the channel and the Reynolds number (Re). A calculation formula of the coefficient of proportionality for the narrow line rod packages is provided. The paper presents a comparison of experimental data and calculated values, which shows usability of the suggested models and calculation formulas.

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

International Nuclear Information System (INIS)

Edelmann, M.

1977-01-01

Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

Sodium flow rate measurement method of annular linear induction pump

International Nuclear Information System (INIS)

Araseki, Hideo

2011-01-01

This report describes a method for measuring sodium flow rate of annular linear induction pumps arranged in parallel and its verification result obtained through an experiment and a numerical analysis. In the method, the leaked magnetic field is measured with measuring coils at the stator end on the outlet side and is correlated with the sodium flow rate. The experimental data and the numerical result indicate that the leaked magnetic field at the stator edge keeps almost constant when the sodium flow rate changes and that the leaked magnetic field change arising from the flow rate change is small compared with the overall leaked magnetic field. It is shown that the correlation between the leaked magnetic field and the sodium flow rate is almost linear due to this feature of the leaked magnetic field, which indicates the applicability of the method to small-scale annular linear induction pumps. (author)

Liquid metal reactor development -Studies on safety measure of LMR coolant

International Nuclear Information System (INIS)

Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

1995-07-01

A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

Liquid metal reactor development -Studies on safety measure of LMR coolant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

Sodium evaporation into a forced argon flow, (1)

International Nuclear Information System (INIS)

Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

1976-01-01

Measurements were made on the rate of evaporation from a rectangular-shaped free surface of liquid sodium into argon flow. Tests were carried out at various levels of sodium temperature, of argon velocity and of argon temperature, under conditions where fog formation could be expected. To gain information on the enhancement of evaporation occasioned by fog formation, a supplementary experiment was performed on convection heat transfer into flowing air from a heated plate of the same geometry as the free surface of the sodium in the preceding measurements. The values obtained for the rate of evaporation and Sherwood number were compared with those predicted by the heat transfer experiment and by the theory by Hill and Szekely. The overall results revealed that the rate of sodium evaporation can amount to as much as three times that predicted by the heat transfer experiment, and that it varies roughly linearly with the heat transfer rate and with the sodium vapor pressure prevailing at the free surface. (auth.)

Sodium fires and its extinguishment; Gorenie, sredstva i sposoby tusheniya natriya

Energy Technology Data Exchange (ETDEWEB)

Mikhedov, V G

1979-03-01

The fire safety problems of NPP with sodium coolants in USSR are presented. The design of sodium reactors is made with premises with sodium coolants being hermetic and filled with nitrogen. Some engineering solutions of fire safety including design, elaboration and choice of construction and protection materials are presented. Some theoretical aspects of sodium burning are presented as well as methods of sodium fire extinguishing methods including the use of powder.

Design and fabrication of magnetic coolant filter

Science.gov (United States)

Prashanth, B. N.

2017-07-01

Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

A dynamic model of the reactor coolant system flow for KMRR plant simulation

International Nuclear Information System (INIS)

Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

1990-01-01

To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

Regulation of liquid metal coolant flow rate in experimental loops

International Nuclear Information System (INIS)

Kozlov, F.A.; Laptev, G.I.

1987-01-01

The possibility to use the VRT-2, RPA-T and R 133 analog temperature regulators for the automated regulation of liquid metal flow rate in the experimental loops for investigations on sodium and sodium-potassium alloy technology is considered. The RPA-T device is shown to be the most convenient one; it is characterized by the following parameters: measuring modulus transfer coefficient is 500; the range of regulating modulus proportionality factor variation - 0.3 - 50; the range of the regulating modulus intergrating time constant variation - 5 - 500 s

Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

Directory of Open Access Journals (Sweden)

Sergey Perevoznikov

2016-10-01

Full Text Available In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas–liquid flow model (sodium–hydrogen–sodium hydroxide. Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

On the method of heat exchange calculation for critical and postcritical regimes in sodium-water steam generators

International Nuclear Information System (INIS)

Khudasko, V.V.; Kardash, D.Yu.; Grachev, N.S.

1986-01-01

Technique for calculating heat exchange in sodium-water steam generators with provisions for steam-water flow non-equilibrium character and moisture additional evaporation in pipes is suggested. Zone of heat exchange crisis representing the zone of transition from developed boiling to postcritical zone is considered. Comparison of estimated and experimental data performed for the following ranges of steame generator parameters: pressure p=7.8-14.0 MPa, coolant flow rate ρw=350-1000 kg/(m 2 xs), inlet sodium temperature T=590-825 K shows their good agreement

Numerical simulation on coolant flow and heat transfer in core

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

Using the coolant temperature noise for measuring the flow rate in the RBMK technological channels

International Nuclear Information System (INIS)

Selivanov, V.M.; Karlov, N.P.; Martynov, A.D.; Prostyakov, V.V.; Lysikov, B.V.; Kuznetsov, B.A.; Pallagi, D.; Khorani, Sh.; Khargitai, T.; Tezher, Sh.

1983-01-01

The problems are considered connected with the possibility of using thermometric correlation method to measure the coolant flow rate in the RBMK reactor technological channels. The main attention is paid to the study of the physical nature of the coolant temperature pulsations and to estimation of the effect of parameters of the primary thermaelectrical converter (TEC) on the results of measurements. In the process of reactor inspections made using the thermometric correlation flowmeter of a special design, the temperature noise distribution in the points of flow rate measurement is studied, the noise intensity and physical nature are determined, as well as the effect of different TEC parameters (TEC inertia and base distance between them) on the measurement accuracy. On the basis of the analysis of the effect on the results of the TEC thermal inertia measured value divergence, tausub(α) and transport time, tau sub(T), a conclusion is made on the necessity of choosing the base distance between TEC with tausub(T)>tausub(d)

Low cycle fatigue testing in flowing sodium at elevated temperatures

International Nuclear Information System (INIS)

Flagella, P.N.; Kahrs, J.R.

1976-01-01

The paper describes equipment developed to obtain low cycle strain-controlled fatigue data in flowing sodium at elevated temperatures. Operation and interaction of the major components of the system are discussed, including the calibration technique using remote strain measurement and control. Confirmation of in-air results using the special technique is demonstrated, with data presented for Type 316 stainless steel tested in high purity flowing sodium at 593 0 C. The fatigue life of the material in sodium is essentially the same as that obtained in air for delta epsilon/sub t/= 1 percent. On the other hand, sodium pre-exposure at 650 0 C for 5000 hours increased the fatigue life in-sodium by a factor of two, and sodium pre-exposure at 718 0 C for 5000 hours increased the fatigue life in-sodium by a factor of three

Sodium flow rate measurement method of annular linear induction pumps

International Nuclear Information System (INIS)

Araseki, Hideo; Kirillov, Igor R.; Preslitsky, Gennady V.

2012-01-01

Highlights: ► We found a new method of flow rate monitoring of electromagnetic pump. ► The method is very simple and does not require a large space. ► The method was verified with an experiment and a numerical analysis. ► The experimental data and the numerical results are in good agreement. - Abstract: The present paper proposes a method for measuring sodium flow rate of annular linear induction pumps. The feature of the method lies in measuring the leaked magnetic field with measuring coils near the stator end on the outlet side and in correlating it with the sodium flow rate. This method is verified through an experiment and a numerical analysis. The data obtained in the experiment reveals that the correlation between the leaked magnetic field and the sodium flow rate is almost linear. The result of the numerical analysis agrees with the experimental data. The present method will be particularly effective to sodium flow rate monitoring of each one of plural annular linear induction pumps arranged in parallel in a vessel which forms a large-scale pump unit.

Development of two-phase Flow Model, 'SOBOIL', for Sodium

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

2000-03-01

The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification.

The effect of coolant quantity on local fuel–coolant interactions in a molten pool

International Nuclear Information System (INIS)

Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

2015-01-01

Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

Experimental study on combustion characteristics of sodium fire in a columnar flow

International Nuclear Information System (INIS)

Zhang Zhigang; Peng Kangwei; Guo Ming; Huo Yan

2014-01-01

In the operation of the sodium-cooled fast reactor, the accident caused by the leakage and combustion of liquid sodium is common and frequent in sodium-related facilities. This paper is based on an experimental study of sodium fire in a columnar flow, which was carried out to focus on the burning characteristics by analyzing the temperature fields in the burner. The injection of 200°C liquid sodium with the flux of 0.5 m 3 /h was poured into a 7.9 m 3 volume stainless steel cylindrical burner to shape a sodium fire, and the data of temperature fields in the burner have been collected by dozens of thermocouples which are laid in the combustion space and sodium collection plate. These results show that the sodium fire in a columnar flow is composed of the foregoing centered columnar fire, the subsequent spray fire caused by atomization and the pool fire on the collection plate. The temperature close to the burning sodium flow maximally reaches up to 950°C. The radial temperatures apart from the sodium flow are relatively low and generally about 200°C, and maximally just 300°C even when close to the sodium collection plate. The maximum temperature of the burning sodium dropping on the collection plate rises in the center of plate, about 528°C. This study is helpful to evaluate the combustion characteristics, formation process and composing forms of the sodium fire in the sodium-related facilities. (author)

Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

International Nuclear Information System (INIS)

Kozma, R.; van Dam, H.; Hoogenboom, J.E.

1992-01-01

The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels

BACCHUS-3D/SP. A computer programme for the three-dimensional description of sodium single-phase flow in bundle geometry

International Nuclear Information System (INIS)

Bottoni, M.; Dorr, B.; Homann, C.; Struwe, D.

1983-07-01

The computer programme BACCHUS implemented at KfK includes a steady-state version, a two-dimensional and a three-dimensional transient single-phase flow version describing the thermal-hydraulic behaviour of the coolant (sodium or water) in bundle geometry under nominal or accident conditions. All versions are coupled with a pin model describing the temperature distribution in fuel (or electrical heaters) and cladding. The report describes the programme from the viewpoints of the geometrical model, the mathematical foundations and the numerical treatment of the basic equations. Although emphasis is put on the three-dimensional version, the two-dimensional and the steady state versions are also documented in self-consistent sections. (orig.) [de

Analytical study on coolant temperature of several leak flows in the experimental VHTr core

International Nuclear Information System (INIS)

Fumizawa, Motoh; Arai, Taketoshi; Miyamoto, Yoshiaki

1982-08-01

This report describes heat transfer analysis of several leak flows which bypass main coolant flow path in the experimental VHTR core. The analysis contains the leak flow at permanent reflectors, replaceable reflectors and gaps between fuel columns. The summary of the results are as follows: (1) the temperature of the leak flow gas increases up to the surface temperature of permanent reflectors, (2) the gas temperature at replaceable reflectors increases at least 40 0 C in case of the worst analytical condition, (3) the gas temperature increases remarkably with decreasing equivalent diameter which is changed by the angle of bevel edge of the reflector, (4) while the gas temperature is low at the upper part of the fuel element, the temperature increases rapidly when it flow down along the gap of the fuel columns. (author)

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

Transformation of sodium from the Rapsodie fast breeder reactor into sodium hydroxide

International Nuclear Information System (INIS)

Roger, J.; Latge, C.; Rodriguez, G.

1994-01-01

One of the major problems raised by decommissioning a fast breeder reactor (FBR) concerns the disposal of the sodium coolant. The Desora operation was undertaken to eliminate the Rapsodie primary sodium as part of the partial decommissioning program, and to develop an operational sodium treatment unit for other needs. The process involves reacting small quantities of sodium in water inside a closed vessel, producing aqueous sodium hydroxide and hydrogen gas. It is described in this work. (O.L.). 4 figs

Fuel-coolant interaction-phenomena under prompt burst conditions

International Nuclear Information System (INIS)

Jacobs, H.; Young, M.F.; Reil, K.O.

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

International Nuclear Information System (INIS)

Ogawa, Masuro

1989-07-01

In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

CFD simulation of a coolant flow and a heat transfer in a pebble bed reactor - HTR2008-58334

International Nuclear Information System (INIS)

In, W. K.; Lee, W. J.; Hassan, Y. A.

2008-01-01

This CFD study is to simulate a coolant(gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO, mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles. (authors)

Study on parameters of self-oscillations of the coolant flow rate in an evaporating channel of a boiling-type reactor

International Nuclear Information System (INIS)

Proshutinskij, A.P.; Lobachev, A.G.

1979-01-01

The experimental data on the oscillation frequencies and amplitudes of the coolant flow rate at the limit of the thermohydraulic stability of the boiling type reactor evaporating channel are presented. The experiments have been carried out on the channel simulators of three modifications -smooth-tube, with intensifiers of a transverse crimp type and of an inner spiral ribbing type. The range of the investigated regime parameters is as follows: the pressure - 2.5-14MPa; the heat flux density is 0.015-0.8MV/m 2 , mass velocity is 252-2520 kg/(m 2 xs), the temperature at the channel entrance is from 50 deg C up to (tsub(s) -5)deg C. The experimental data analysis is carried out on the assumption that the period of parameter oscillations in the steam generating channel equals the time of the coolant transfer through the channel. The formular is obtained which provides 25% accuracy of the oscillation frequency calculation in the range of underheating parameter variation B=0.5-3.0. As a result the following conclusions have been made: the oscillation frequency of the coolant flow rate is connected with the time of its transfer through the channel and does not practically depend on the type of the heat exchange intensifiers and the degree of the flux throttling at the channel entrance; the self-oscillation amplitude of the coolant flow rate depends on the regime and structural parameters as well

CANDU with supercritical water coolant: conceptual design features

International Nuclear Information System (INIS)

Spinks, N.

1997-01-01

An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

Distribution and behavior of tritium in the Coolant-Salt Technology Facility

International Nuclear Information System (INIS)

Mays, G.T.; Smith, A.N.; Engel, J.R.

1977-04-01

A 1000-MW(e) Molten-Salt Breeder Reactor (MSBR) is expected to produce 2420 Ci/day of tritium. As much as 60 percent of the tritium produced may be transported to the reactor steam system (assuming no retention by the secondary coolant salt), where it would be released to the environment. Such a release rate would be unacceptable. Experiments were conducted in an engineering-scale facility--the Coolant-Salt Technology Facility (CSTF)--to examine the potential of sodium fluoroborate, the proposed coolant salt for an MSBR, for sequestering tritium. The salt was believed to contain chemical species capable of trapping tritium. A series of 5 experiments--3 transient and 2 steady-state experiments--was conducted from July of 1975 through June of 1976 where tritium was added to the CSTF. The CSTF circulated sodium fluoroborate at temperatures and pressures typical of MSBR operating conditions. Results from the experiments indicated that over 90 percent of tritium added at steady-state conditions was trapped by sodium fluoroborate and appeared in the off-gas system in a chemically combined (water-soluble) form and that a total of approximately 98 percent of the tritium added at steady-state conditions was removed through the off-gas system overall

Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

Energy Technology Data Exchange (ETDEWEB)

Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

2014-10-15

Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

Sodium voiding analysis in Kalimer

International Nuclear Information System (INIS)

Chang, Won-Pyo; Jeong, Kwan-Seong; Hahn, Dohee

2001-01-01

A sodium boiling model has been developed for calculations of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. The sodium boiling in liquid metal reactors using sodium as coolant should be modeled because of phenomenon difference observed from that in light water reactor systems. The developed model is a multiple -bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbles that fill the whole cross section of the coolant channel except for liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble. The present study is focused on not only demonstration of the sodium voiding behavior predicted by the developed model, but also confirmation on qualitative acceptance for the model. In results, the model catches important phenomena for sodium boiling, while further effort should be made for the complete analysis. (author)

Experimental studies of flow induced vibrations of the fuel assembly for the PEC reactor

International Nuclear Information System (INIS)

Pitimada, D.; Presaghi, M.; Tampone, O.; Cesari, F.

1977-01-01

The vibration behaviour of an assembly of seven mock-up fuel bundles of PEC reactor has been investigated. The assembly was excited by a parallel flow of water simulating sodium. The motion of the group (or of a single bundle in the group) has been measured in transverse sections detecting two orthogonal components of displacement. During the experiences the following parameters were varied: bundle foot and pads restraints, flow rate condition, coolant flow outlet conditions at the head of fuel bundles. Experimental data were processed in order to obtain: trajectories of three points of fuel bundle axis, power density spectra of measured vibration amplitudes, correlations between coolant flow rate and vibration amplitude R.M.S. (author)

Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

International Nuclear Information System (INIS)

Zeuch, W.R.; A-Moneim, M.T.

1979-01-01

A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL

Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

International Nuclear Information System (INIS)

Chvetsov, I.; Kouznetsov, I.; Volkov, A.

2000-01-01

GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

Energy Technology Data Exchange (ETDEWEB)

Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

2016-02-15

Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

Energy Technology Data Exchange (ETDEWEB)

Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

2002-07-01

Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

International Nuclear Information System (INIS)

Lister, D.

2002-01-01

Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

Particle image velocimetry measurement of complex flow structures in the diffuser and spherical casing of a reactor coolant pump

Directory of Open Access Journals (Sweden)

Yongchao Zhang

2018-04-01

Full Text Available Understanding of turbulent flow in the reactor coolant pump (RCP is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms. Keywords: Diffuser, Flow Structures, Particle Image Velocimetry, Reactor Coolant Pump, Spherical Casing, Velocity Distribution

The Performance test of Mechanical Sodium Pump with Water Environment

International Nuclear Information System (INIS)

Cho, Chungho; Kim, Jong-Man; Ko, Yung Joo; Jeong, Ji-Young; Kim, Jong-Bum; Ko, Bock Seong; Park, Sang Jun; Lee, Yoon Sang

2015-01-01

As contrasted with PWR(Pressurized light Water Reactor) using water as a coolant, sodium is used as a coolant in SFR because of its low melting temperature, high thermal conductivity, the high boiling temperature allowing the reactors to operate at ambient pressure, and low neutron absorption cross section which is required to achieve a high neutron flux. But, sodium is violently reactive with water or oxygen like the other alkali metal. So Very strict requirements are demanded to design and fabricate of sodium experimental facilities. Furthermore, performance testing in high temperature sodium environments is more expensive and time consuming and need an extra precautions because operating and maintaining of sodium experimental facilities are very difficult. The present paper describes performance test results of mechanical sodium pump with water which has been performed with some design changes using water test facility in SAM JIN Industrial Co. To compare the hydraulic characteristic of model pump with water and sodium, the performance test of model pump were performed using vender's experimental facility for mechanical sodium pump. To accommodate non-uniform thermal expansion and to secure the operability and the safety, the gap size of some parts of original model pump was modified. Performance tests of modified mechanical sodium pump with water were successfully performed. Water is therefore often selected as a surrogate test fluid because it is not only cheap, easily available and easy to handle but also its important hydraulic properties (density and kinematic viscosity) are very similar to that of the sodium. Normal practice to thoroughly test a design or component before applied or installed in reactor is important to ensure the safety and operability in the sodium-cooled fast reactor (SFR). So, in order to estimate the hydraulic behavior of the PHTS pump of DSFR (600 MWe Demonstraion SFR), the performance tests of the model pump such as performance

Real-time algorithm for the measurement of liquid metal coolant flow velocity with correlated thermal signals

International Nuclear Information System (INIS)

Moazzeni, Taleb; Jiang, Yingtao; Ma, Jian; Li, Ning

2009-01-01

One flow meter was developed especially for the environment of high irradiation, pressure, and temperature. The transit time of natural random temperature fluctuation in process, for example nuclear reactor, can be obtained based on the cross-correlation method, which has already been shown that it is capable in situations where no other flow meter can be used. Thereby, the flow rate can be derived in pipe flow if the area of cross-section is known. In practice, the evaluation of the integrals over the measurement time in cross-correlation method will lead errors caused by peak detection from flat cross correlation coefficient distribution or additional peaks. One Auto-Adaptive Impulse Response Function estimation is introduced and significantly narrower peak will be obtained. Fiber optic sensors are advantageous for temperature measurements in the reactor pressure vessels. However, the corrosive coolant (as liquid lead/lead alloy or molten salt coolant) is a barrier of the optic sensor in such application. Thermocouple with grounded stainless steel shielding material would have same life time with structure material in reactor, although thermocouple has relatively slow response. The degradation due to corrosion/erosion will not introduce measurement error or necessary calibration, because only the correlation between signals is taken into consideration during measurements. Experiments conducted in a testing hydraulic facility approved the considerable improvement of accuracy by this new algorithm using thermocouple temperature sensors. (author)

Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

2011-01-01

Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

International Nuclear Information System (INIS)

Dinh Van Thin; Tran Thi Nhan

2015-01-01

Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

International Nuclear Information System (INIS)

Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

2010-02-01

The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

International Nuclear Information System (INIS)

Malkis, V.A.; Lokshin, V.A.

1983-01-01

Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Jacobs, H.; Young, M.F.; Reil, K.O.

1979-01-01

The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

Renal blood flow, early distal sodium, and plasma renin concentrations during osmotic diuresis

DEFF Research Database (Denmark)

Leyssac, P P; Holstein-Rathlou, N H; Skøtt, O

2000-01-01

.6 mmHg. Urine flow increased 10-fold, and sodium excretion increased by 177%. Plasma renin concentration (PRC) increased by 58%. Renal blood flow and glomerular filtration rate decreased, however end-proximal flow remained unchanged. After a similar volume of hypotonic glucose (152 mM), ED......(NaCl) increased by 3.6 mM, (P renal hemodynamics, urine flow, sodium excretion rate, or PRC. Infusion of 300 micromol NaCl in a smaller volume caused ED(NaCl) to increase by 6.4 mM without significant changes in PRC. Urine flow and sodium excretion increased significantly...

Development of a model to predict flow oscillations in low-flow sodium boiling

International Nuclear Information System (INIS)

Levin, A.E.; Griffith, P.

1980-04-01

Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed

Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

2002-01-01

In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

Study on the quench behavior of molten fuel material jet into coolant

International Nuclear Information System (INIS)

Abe, Yutaka; Kizu, Tetsuya; Arai, Takahiro; Nariai, Hideki; Chitose, Keiko; Koyama, Kazuya

2004-01-01

In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. In the present experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The distributions of the fragmented droplet diameter from the molten material jet are evaluated by correcting the solidified particles. The experimental results of the mean fragmented droplet diameter are compared with the existing theories. Consequently, the fragmented droplet diameter is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass ratio of the molten particle to the total injected mass by combining an appropriate heat transfer model. The heat transfer model used in the present study is composed of the fragmentation model based on the Kelvin-Helmholtz instability. The mass ratio of the molten fragment to total mass of the melted mixed oxide fuel in sodium coolant estimated in the present study is very small. The result means that most of the molten mixed oxide fuel material injected into the sodium coolant can be cooled down under the solidified temperature, that is so called quenched, if the amount of the coolant is sufficient. (author)

Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.

1986-01-01

In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

International Nuclear Information System (INIS)

Siamidis, John; Mason, Lee

2006-01-01

This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS O pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids

Transient two-phase performance of LOFT reactor coolant pumps

International Nuclear Information System (INIS)

Chen, T.H.; Modro, S.M.

1983-01-01

Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

CFD analyses of coolant channel flowfields

Science.gov (United States)

Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

1993-01-01

The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2015-01-01

Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

Science.gov (United States)

Salamon, V.; Senthil kumar, D.; Thirumalini, S.

2017-08-01

The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

Temperature noise analysis and sodium boiling detection in the fuel failure mockup

International Nuclear Information System (INIS)

Sides, W.H. Jr.; Fry, D.N.; Leavell, W.H.; Mathis, M.V.; Saxe, R.F.

1976-01-01

Sodium temperature noise was measured at the exit of simulated, fast-reactor fuel subassemblies in the Fuel Failure Mockup (FFM) to determine the feasibility of using temperature noise monitors to detect flow blockages in fast reactors. Also, acoustic noise was measured to determine whether sodium boiling in the FFM could be detected acoustically and whether noncondensable gas entrained in the sodium coolant would affect the sensitivity of the acoustic noise detection system. Information from these studies would be applied to the design of safety systems for operating liquid-metal fast breeder reactors (LMFBRs). It was determined that the statistical properties of temperature noise are dependent on the shape of temperature profiles across the subassemblies, and that a blockage upstream of a thermocouple that increases the gradient of the profile near the blockage will also increase the temperature noise at the thermocouple. Amplitude probability analysis of temperature noise shows a skewed amplitude density function about the mean temperature that varies with the location of the thermocouple with respect to the blockage location. It was concluded that sodium boiling in the FFM could be detected acoustically. However, entrained noncondensable gas in the sodium coolant at void fractions greater than 0.4 percent attenuated the acoustic signals sufficiently that boiling was not detected. At a void fraction of 0.1 percent, boiling was indicated only by the two acoustic detectors closest to the boiling site

Determination of primary flow by correlation of temperatures of the coolant; Medicion de caudal primario por correlacion de temperaturas del refrigerante

Energy Technology Data Exchange (ETDEWEB)

Villanueva, Jose [Comision Nacional de Energia Atomica, Ezeiza (Argentina). Centro Atomico Ezeiza

2003-07-01

Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

Flow in sodium loop surge tank

International Nuclear Information System (INIS)

Matal, O.; Martoch, J.

1977-01-01

The alternate liquid flow, the condition of vortex formation, gas entrainment in the discharge and the liquid level characteristics are studied using the models of the vertical and horizontal surge tanks of a sodium circuit with pump and heat exchangers. The conditions for vortex formation are more favourable in the vertical cylindrical tank than in the horizontal tank. The size of the vortex produced in the tank is affected by the initial speed circulation, due as a rule to an unsuitable inlet design. The proposed design considers an inlet below the sodium level using capped perforated pipes. Vortex formation, gas transport to the discharge pipe and turbulences of the liquid in the tank may be prevented by dividing the tank to the discharge and the inlet areas using perforated partitions, and by inserting the discharge cylinder above the discharge pipe outflow. The liquid level in the tank may be calmed by screens or by perforated plates. The adaptation of the surge tank of the sodium circuit will probably eliminate vortex formation and the entrainment of cover gas into the discharge piping and the sodium circuit under nominal conditions. (J.B.)

Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

Science.gov (United States)

Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

2017-11-01

The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

APPLICATION FEATURES OF SPATIAL CONDUCTOMETRY SENSORS IN MODELLING OF COOLANT FLOW MIXING IN NUCLEAR POWER UNIT EQUIPMENT

Directory of Open Access Journals (Sweden)

A. A. Barinov

2016-01-01

Full Text Available Coolant flow mixing processes with different temperatures and concentrations of diluted additives widely known in nuclear power units operation. In some cases these processes make essential impact on the resource and behavior of the nuclear unit during transient and emergency situations. The aim of the study was creation of measurement system and test facility to carry out basic tests and to embed spatial conductometry method in investigation practice of turbulent coolant flows. In the course of investigation measurement system with sensors and experimental facility was designed, several first tests were carried out. A special attention was dedicated to calibration and clarification of conductometry sensor application methodologies in studies of turbulent flow characteristics. Investigations involved method of electrically contrast tracer jet with concurrent flow in closed channel of round crosssection. The measurements include both averaged and unsteady realizations of measurement signal. Experimental data processing shows good agreement with other tests acquired from another measurement systems based on different physical principles. Calibration functions were acquired, methodical basis of spatial conductometry measurement system application was created. Gathered experience of spatial sensor application made it possible to formulate the principles of further investigation that involve large-scale models of nuclear unit equipment. Spatial wire-mesh sensors proved to be a perspective type of eddy resolving measurement devices.

Sodium fires and nuclear power station safety

International Nuclear Information System (INIS)

Ivanenko, V.N.; Zubin, A.; Drobyshev, A.V.

1986-01-01

The danger of sodium aerosol release at a design basis accident (DBA) of a sodium-cooled fast reactor that involves coolant leakage and burning, is being analyzed. It has been shown that radioactive and toxic releases at DBA do not exceed permissible values. Some means of sodium fire fighting are described. (author)

Korea advanced liquid metal reactor development - Development of measuring techniques of the sodium two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Cha, Jae Eun [Pohang University of Science and Technology, Pohang (Korea)

2000-04-01

The technology which models and measures the behavior of bubble in liquid sodium is very important to insure the safety of the liquid metal reactor. In this research, we designed/ manufactured each part and loop of experimental facility for sodium two phase flow, and applied a few possible methods, measured characteristic of two phase flow such as bubbly flow. A air-water loop similar to sodium loop on each measuring condition was designed/manufactured. This air-water loop was utilized to acquire many informations which were necessary in designing the two phase flow of sodium and manufacturing experimental facility. Before the manufacture of a electromagnetic flow meter for sodium, the experiment using each electromagnetic flow mete was developed and the air-water loop was performed to understand flow characteristics. Experiments for observing the signal characteristics of flow were performed by flowing two phase mixture into the electromagnetic flow mete. From these experiments, the electromagnetic flow meter was designed and constructed by virtual electrode, its signal processing circuit and micro electro magnet. It was developed to be applicable to low conductivity fluid very successfully. By this experiment with the electromagnetic flow meter, we observed that the flow signal was very different according to void fraction in two phase flow and that probability density function which was made by statistical signal treatment is also different according to flow patterns. From this result, we confirmed that the electromagnetic flow meter could be used to understand the parameters of two phase flow of sodium. By this study, the experimental facility for two phase flow of sodium was constricted. Also the new electromagnetic flow meter was designed/manufactured, and experimental apparatus for two phase flow of air-water. Finally, this study will be a basic tool for measurement of two phase flow of sodium. As the fundamental technique for the applications of sodium at

Coolant voiding analysis following SGTR for an HLMC reactor

International Nuclear Information System (INIS)

Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

2000-01-01

Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

Experimental evaluation of permanent magnet probe flowmeter measuring high temperature liquid sodium flow in the ITSL

Energy Technology Data Exchange (ETDEWEB)

Jeong, Uiju; Kim, Yun Ho [Nuclear engineering Department, Hanyang University, 17 Haengdang-dong, Seongdong-gu, Seoul 133-791 (Korea, Republic of); Kim, Jong-Man; Kim, Tae-Joon [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Sung Joong, E-mail: sungjkim@mit.edu [Nuclear engineering Department, Hanyang University, 17 Haengdang-dong, Seongdong-gu, Seoul 133-791 (Korea, Republic of)

2013-12-15

Highlights: • An Instrument Test Sodium Loop (ITSL) has been built and tested in various conditions at KAERI. • Free fall of liquid sodium was conducted experimentally and numerically. • A Permanent Magnet Probe Flowmeter (PMPF) was experimented in the ITSL. • Excellent linearity of the PMPF was achieved under high temperature condition. - Abstract: The Instrument Test Sodium Loop (ITSL) installed at Korea Atomic Energy Research Institute (KAERI) is a medium-size experimental facility dedicated to obtaining relevant experimental data of liquid sodium flow characteristics under various thermal hydraulic conditions and sodium purification. The ITSL has been utilized to perform thermal flow measurement of the liquid sodium and to calibrate a Permanent Magnet Probe Flowmeter (PMPF). The primary objective of this study is to obtain liquid sodium flow rate given a wide temperature range using the PMPF. Non-stationary method was adopted for the calibration of the probe given the liquid sodium temperature range of 150–415 °C. A relationship between the measured voltage signal and flow rate was obtained successfully. It is observed that the calibration experiments result in excellent linear relationships between measured voltage and volumetric flow rate at various temperature conditions. Also a computational analysis using FlowMaster, is employed to facilitate the calibration process by predicting the liquid sodium flow rate. Finally the effect of the fluid temperature on thermal flow measurements is discussed in light of the obtained experimental data.

Analysis of clad motion during a loss of flow (LOF) accident in a fast sodium cooled reactor

International Nuclear Information System (INIS)

Henkel, P.

1985-10-01

A new model describing clad motion during a Loss of Flow (LOF) accident in a Liquid Metal Cooled Fast (Breeder) Reactor (LMFBR) is presented. Its special features are Clad motion is treated within a fuel pin bundle. The bundle geometry is represented by an equivalent annular geometry which serves as the descriptional basis for the clad motion analysis; Several flow regimes are considered. These include a wave or film flow along the fuel pin surfaces as well as a drop flow within the coolant channels. A new entrainment criterion is successfully applied to describe the entrainment of molten cladding and the coolant flow is modelled as a two-dimensional, monstationary flow. Therefore, radial cross flows in a pin bundle can be calculated. Especially, thermal incoherency effects can be treated consistently. The analysis of clad motion in the two experiments STAR1 and STAR2 using the subsequently presented SANDCMOT model gives good agreement with the experimental data. (orig.) [de

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

Enhancing resistance to burnout via coolant chemistry

Energy Technology Data Exchange (ETDEWEB)

Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

2003-07-01

Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

Study of core characteristics on fuel and coolant type. Results of F/S phase-I

International Nuclear Information System (INIS)

Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

2001-03-01

The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

International Nuclear Information System (INIS)

Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

2007-01-01

One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

2013-06-01

Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

International Nuclear Information System (INIS)

Yoo, J. W.; Kim, S. J.; Kim, Y. I.

2000-01-01

A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

Development of industrial utilization of metallic sodium

International Nuclear Information System (INIS)

Yuhara, Shunichi

1995-01-01

Sodium exists in large quantity, being ranked to 6th in the existence proportion of elements, and takes 2.83% of the matters composing earth crust. Sodium is an alkali metal which is light weight, chemically very active and a strong reducing substance. It is excellent in the compatibility with iron and steel materials, and it possesses good heat conduction and flow characteristics and stable nuclear characteristics. Since the industrial production of sodium became practical, its utilization was developed as the reducing agent and catalyst in chemical industry, the core coolant and heat transport medium for nuclear reactors, the material composing the secondary batteries for storing electric power, and the auxiliaries for metal refining and so on. The industrial production of metallic sodium is carried out by the electrolysis of melted salt, namely Downs process. The production of metallic sodium in Japan is 3000-6000 t yearly, and its import is 300-350 t. Its main use is for organic chemical industry including dye production. The grades of metallic sodium products and their uses are shown. The utilization of sodium for large fast reactors, the utilization of NaK as the heat transport and cooling medium for space use nuclear reactors and deep sea fast reactor system, and the utilization of sodium as the catalyst in dye production, for silicon carbide fiber production and for agricultural and medical chemical production are reported. (K.I.)

Toxicology of plutonium-sodium

International Nuclear Information System (INIS)

Hackett, P.L.

1982-01-01

Scenarios for liquid-metal fast breeder reactor (LMFBR) accidents predict the loss of sodium coolant, with subsequent core melt-down and release of mixed sodium-fuel aerosols [Na-(PuU)O 2 ] into the environment. Studies in other laboratories demonstrated that mixed aerosols of Na 2 O-PuO 2 were more readily transported from the lung than PuO 2 aerosols. We therefore devised a continuous aerosol-generating system for animal exposures in which laser-generated fuel aerosols were swept through sodium vapor to form sodium-fuel aerosols. These fuel and sodium-fuel aerosols were compared with regard to their physicochemical properties and their biological behavior following inhalation studies in rats and dogs

A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident

International Nuclear Information System (INIS)

Gaballah, I.

1978-09-01

A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident. A theory was developed for the calculation of a dispersed two phase flow with heat addition in a channel with general area change. The theory was used to study different thermodynamic and gasdynamic processes, which may occur during the emergency cooling after a LOCA of a pressurized water reactor. The basic equations were formulated and solved numerically. The heat transfer mechanism was examined. Calculations have indicated that the radiative heat flux component is small compared to the convective component. A drop size spectrum was used in the calculations. Its effect on the heat transfer was investigated. It was found that the calculation with a mean drop diameter gives good results. Significant thermal non-equilibrium has been evaluated. The effect of different operating parameters on the degree of thermal non-equilibrium was studied. The flow and heat transfer in a channel with cross-sectional area change were calculated. It was shown that the channel deformation affects the state properties and the heat transfer along the channel very strongly. (orig.) 891 GL [de

Reactor coolant pump seal leakage monitoring

International Nuclear Information System (INIS)

Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

1986-01-01

Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

Physical properties of liquid sodium

International Nuclear Information System (INIS)

Alberdi Primicia, J.; Martinez Piquer, T.A.

1977-01-01

The molten sodium has been the more accepted coolant for the first generation of FBR, by this reason the knowledge of its technology is needed for the development of the next LMFBR. A series of necessary data for designing sodium liquid systems are given. Tables and graphics about the most important physical sodium properties between 1200-1400 degC are gathered. The results have been obtained from equations that relate the properties with temperature using a Fortran IV program. (author) [es

Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

Directory of Open Access Journals (Sweden)

Navid Bozorgan

2015-07-01

Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

The dismantling of fast reactors: sodium processing

International Nuclear Information System (INIS)

Rodriguez, G.; Berte, M.; Serpante, J.P.

1999-01-01

Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

1998-03-01

The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

International Nuclear Information System (INIS)

Hahn, G.

1974-01-01

The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

A study on safety measure of LMR coolant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tai; Choi, Y D; Choi, J H; Kim, T J; Jeong, K C; Kwon, S W; Kim, B H; Jeong, J Y; Park, J H; Kim, K R; Jo, B R

1997-08-01

A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

A study on safety measure of LMR coolant

International Nuclear Information System (INIS)

Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

1997-08-01

A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H 2 gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs

CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

International Nuclear Information System (INIS)

Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

2015-01-01

Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

A basic experimental study on combustion of suspended sodium droplet. 2

International Nuclear Information System (INIS)

Sato, Kenji

1999-10-01

For appropriate understanding and/or prediction of the combustion behavior of sodium, working as liquid coolant in fast breeder reactors, in case of leakage accident, phenomenological analyses of the behavior must be also important along with conventional engineering approach. Following our previous study in the last year, the major objective of this experimental research is to elucidate the effects of the initial temperature and diameter of droplet, and of the air flow velocity on ignition process of a sodium droplet, by exposing a suspended droplet to the air flow at room-temperature. In the experiments, a high-temperature droplet suspended from the end of a fine stainless steel nozzle of the liquid sodium supply system was exposed to an upward air flow, and the ignition and succeeding combustion phenomena were observed by using high-speed color video recording system. In the preliminary study, the effects of lighting and image data processing on obtaining pictures suitable to analyses were investigated with the apparatus used in the previous study. After the experimental apparatus was modified partially in order to expose the unreacted droplet to the air flow more quickly, main experiments were performed in synthetic dry air or oxygen-nitrogen mixture of 21% oxygen. Good quality pictures of the phenomena achieved under good conditions were recorded even for a few cases. The details of the ignition process of a sodium droplet, including the aspects of the surface and light emission, were examined, and the effects of the air flow velocity were discussed. Since number of performed experimental runs was small, the effects of the initial droplet temperature were not examined. (author)

Contact condensation effects in the main coolant pipe

International Nuclear Information System (INIS)

Haefner, W.; Fischer, K.

1990-01-01

Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

Problems of technology and corrosion in sodium coolant and cover gas

International Nuclear Information System (INIS)

Kuenstler, K.; Ullmann, H.

1977-07-01

The meeting encloses the following themes: (i) Reactions in the system sodium-steel-cover gas (ii) Corrosion behaviour of structural and cladding materials (iii) Determination of impurities in sodium and cover gas (iv) Technology of sodium and cover gas (v) Testing equipments (vi) Safety problems

Study on chemical reactivity control of liquid sodium. Research program

International Nuclear Information System (INIS)

Saito, Jun-ichi; Ara, Kuniaki; Sugiyama, Ken-ichiro; Kitagawa, Hiroshi; Oka, Nobuki; Yoshioka, Naoki

2007-01-01

Liquid sodium has the excellent properties as coolant of the fast breeder reactor (FBR). On the other hand, it reacts high with water and oxygen. So an innovative technology to suppress the reactivity is desired. The purpose of this study is to control the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles (we call them Nano-particles) into liquid sodium. We focus on the atomic interaction between Nano-particles and sodium atoms. And we try to apply it to suppress the chemical reactivity of liquid sodium. Liquid sodium dispersing Nano-particles is named 'Nano-fluid'. Research programs of this study are the Nano-particles production, the evaluation of reactivity suppression of liquid sodium and the feasibility study to FBR plant. In this paper, the research programs and status are described. The important factors for particle production were understood. In order to evaluate the chemical reactivity of Nano-fluid the research programs were planned. The feasibility of the application of Nano-fluid to the coolant of FBR plant was evaluated preliminarily from the viewpoint of design and operation. (author)

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

Effect of removal of a central thimble on coolant flow distribution in a research reactor fuel element

International Nuclear Information System (INIS)

Green, W.J.

1977-01-01

Using two twice full-size models of a HIFAR research reactor fuel element, experiments have been performed to determine how the flow distribution of coolant gas through the element in a transfer flask is affected by removal of the central instrumentation thimble. With the thimble present, experimental flow results agree with theoretical predictions. Over the range of total flowrates considered, mass flow apportioning among the five annular channels was independent of annular channel Reynolds number (in the range 3500 to 10,500) and ranged between 13% and 27% of the total flowrate. For the case with the thimble removed, interesting experimental flow characteristics were obtained which could not have been predicted. Flow apportioning among the annular channels was found to be uniquely dependent upon total flowrate and ranged between 3% and 8% for the experimental conditions investigated (annular channel Reynolds numbers in the range 800 to 4000). (Author)

Experiment studies of fuel rod vibration in coolant flow for substantiation of vibration stability of fuel rods with no fretting-wear

International Nuclear Information System (INIS)

Egorov, Yu. V.; Afanasiev, A. V.; Makarov, V. V.; Matvienko, I. V.

2013-01-01

For substantiation of vibration stability it is necessary to determine the ultimate permissible vibration levels which do not cause fretting, to compare them with the level of fuel rod vibration caused by coolant flow. Another approach is feasible if there is experience of successful operation of FA-prototypes. In this case in order to justify vibration stability it may be sufficient to demonstrate that the new element does not cause increased vibration of the fuel rod. It can be done by comparing the levels of hydro-dynamic fuel rod vibration and FA new designs. Program of vibration tests of TVS-2M model included studies of forced oscillations of 12 fuel rods in the coolant flow in the spans containing intensifiers, in the reference span without intensifiers, in the lower spans with assembled ADF and after its disassembly. The experimental results for TVS-2M show that in the spans with intensifier «Sector run» the level of movements is 6% higher on the average than in the span without intensifiers, in the spans with intensifier «Eddy» it is 2% higher. The level of fuel rod vibration movements in the spans with set ADF is 2 % higher on the average than without ADF. During the studies of TVS-KVADRAT fuel rod vibration, the following tasks were solved: determination of acceleration of the middle of fuel rod spans at vibration excited due to hydrodynamics; determination of influence of coolant thermal- hydraulic parameters (temperature, flowrate, dynamic pressure) on fuel rod vibration response; determination of influence of span lengths on the vibration level. Conclusions: 1) The vibration tests of the full-scale model of TVS-2M in the coolant flow showed that the new elements of TVS-2M design (intensifiers of heat exchange and ADF) are not the source of fuel rod increased vibration. Considering successful operation of similar fuel rod spans in the existing TVS-2M design, vibration stability of TVS-2M fuel rods with new elements is ensured on the mechanism of

Carbon in sodium: a status review of the USA R and D work

International Nuclear Information System (INIS)

McCown, J.J.; Bagnall, C.

1979-01-01

A review is presented concerning R and D work on carbon in sodium with reference to LMFBR primary coolant circuits. The chemistry of carbon in sodium, analysis and monitoring of carbon in sodium, carbon meters, and problems of carbon in sodium are described. 31 references

Experimental and numerical investigation of the coolant mixing during fast deboration transients

International Nuclear Information System (INIS)

Hoehne, T.; Rohde, U.; Weiss, F.P.

1999-01-01

For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

Energy Technology Data Exchange (ETDEWEB)

Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

Triboengineering problems of lead coolant in innovative fast reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

2013-01-01

Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

Experiments on simulation of coolant mixing in fuel assembly head and core exit channel of WWER-440 reactor

International Nuclear Information System (INIS)

Kobzar, L.L; Oleksyuk, D.A.

2006-01-01

RRC 'Kurchatov Institute' has performed coolant mixing investigation in a head of a full-size simulator of WWER-440 fuel assembly. The experiments were focused on obtaining the data important for investigating the trends in temperature difference between the value registered by a ICIS thermocouple and the value of average temperature. The completed experiments ensure representative of configuration simulation by reproducing every construction peculiar feature of flow part of fuel assembly in the domain between the lower spacing grid and thermocouple location, and also by slightly modified fuel assembly regular elements (or analogues thereof). For the purpose of effectiveness of coolant mixing assessment within the head cross section of FA simulator, we measured coolant temperature distribution both in the place where coolant flow leaves the rod bundle simulator (in 39 data points along the cross section) and in the cross section location of regular ICIS thermocouple simulator (30 data points). The testing was conducted with pressure of (90 - 95) bar, mass coolant flow rates up to 2000 kg/(m 2 .s), temperature of coolant heating in 'hot' parts of the bundle up to 35.. and differences between coolant temperature extremes measured in rod bundle simulator outlet up to 20... Temperature fields were registered in 63 conditions that differ in coolant flow and inlet coolant temperature, electrical heating rate of FA simulator, and radial coolant distribution. In certain registered conditions we simulated coolant leakage to the space between the fuel assemblies. The received test data may be important both for investigation of dependencies between the coolant temperature in regular thermocouple location or average outlet temperature in assembly head, and for validation of CFD codes or subchannel codes (Authors)

Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

International Nuclear Information System (INIS)

Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

1984-07-01

The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

Methodology for Extraction of Remaining Sodium of Used Sodium Containers

International Nuclear Information System (INIS)

Jung, Minhwan; Kim, Jongman; Cho, Youngil; Jeong, Jiyoung

2014-01-01

Sodium used as a coolant in the SFR (Sodium-cooled Fast Reactor) reacts easily with most elements due to its high reactivity. If sodium at high temperature leaks outside of a system boundary and makes contact with oxygen, it starts to burn and toxic aerosols are produced. In addition, it generates flammable hydrogen gas through a reaction with water. Hydrogen gas can be explosive within the range of 4.75 vol%. Therefore, the sodium should be handled carefully in accordance with standard procedures even though there is a small amount of target sodium remainings inside the containers and drums used for experiment. After the experiment, all sodium experimental apparatuses should be dismantled carefully through a series of draining, residual sodium extraction, and cleaning if they are no longer reused. In this work, a system for the extraction of the remaining sodium of used sodium drums has been developed and an operation procedure for the system has been established. In this work, a methodology for the extraction of remaining sodium out of the used sodium container has been developed as one of the sodium facility maintenance works. The sodium extraction system for remaining sodium of the used drums was designed and tested successfully. This work will contribute to an establishment of sodium handling technology for PGSFR. (Prototype Gen-IV Sodium-cooled Fast Reactor)

Under-Sodium-Viewing as one technique for periodic inspections in sodium-cooled fast reactors-- possibilities and limits

International Nuclear Information System (INIS)

Weiss, H.

1979-07-01

Periodic inspections are gaining increasingly technical importance for fast sodium cooled reactors. Among others the reactor tank and its internals have to be inspected, whereby licensing experts partly are requesting the standards of Light Water Reactors. This leads to difficulties in sodium cooled reactors because of the non-transparent coolant sodium and their compact structure. In order to avoid the complete dumping of the sodium, the under sodium viewing shall be applied besides other inspection methods. Since this is a new method, which is still in its development phase, this report presents and discusses the technical and physical basis and outlines possibilities and limits [de

Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

Energy Technology Data Exchange (ETDEWEB)

Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

2006-07-01

This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

International Nuclear Information System (INIS)

Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

2006-01-01

This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

International Nuclear Information System (INIS)

Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

2014-01-01

The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

Recent developments in the analysis of coolant sodium for fast breeder reactors

International Nuclear Information System (INIS)

Keough, R.F.; McCown, J.J.

1976-01-01

The measurement of impurities in sodium is an important part of the sodium technology for the fast breeder reactor program. A knowledge of impurity levels in sodium provides an important source of information on such things as corrosion rates, air and water leak location and detection, and the effectiveness of the sodium purification systems. A discussion is presented of some of the analytical techniques for sodium characterization in reactor heat transfer systems. Emphasis is placed on some recently developed alternatives to laboratory analysis

Compatibility of structural materials with fusion reactor coolant and breeder fluids

International Nuclear Information System (INIS)

DeVan, J.H.

1979-01-01

Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

2016-11-15

Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Brachet, Alain.

1981-01-01

This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

Flow induced vibration of the large-sized sodium valve for MONJU

Energy Technology Data Exchange (ETDEWEB)

Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel Development Corporation, Nariata-cho, O-arai Machi, Ibaraki-ken (Japan)

1977-12-01

Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

Flow induced vibration of the large-sized sodium valve for MONJU

International Nuclear Information System (INIS)

Sato, K.

1977-01-01

Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Hobbins, R.R.

1981-01-01

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

Science.gov (United States)

Gaugler, R. E.

1994-01-01

As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

Minimizing secondary coolant blowdown in HANARO

International Nuclear Information System (INIS)

Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

2000-01-01

There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse

2012-01-01

the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

International Nuclear Information System (INIS)

Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

1975-01-01

Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Energy Technology Data Exchange (ETDEWEB)

Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

2017-08-15

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

HTGR-GT primary coolant transient resulting from postulated turbine deblading

International Nuclear Information System (INIS)

Cadwallader, G.J.; Deremer, R.K.

1980-11-01

The turbomachine is located within the primary coolant system of a nuclear closed cycle gas turbine plant (HTGR-GT). The deblading of the turbine can cause a rapid pressure equilibration transient that generates significant loads on other components in the system. Prediction of and design for this transient are important aspects of assuring the safety of the HTGR-GT. This paper describes the adaptation and use of the RATSAM program to analyze the rapid fluid transient throughout the primary coolant system during a spectrum of turbine deblading events. Included are discussions of (1) specific modifications and improvements to the basic RATSAM program, which is also briefly described; (2) typical results showing the expansion wave moving upstream from the debladed turbine through the primary coolant system; and (3) the effect on the transient results of different plenum volumes, flow resistances, times to deblade, and geometries that can choke the flow

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

Science.gov (United States)

Rezania, A.; Rosendahl, L. A.

2012-06-01

The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

International Nuclear Information System (INIS)

1994-01-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-07-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

Friction and wear in sodium

International Nuclear Information System (INIS)

Hoffman, N.J.; Droher, J.J.

1973-01-01

In the design of a safe and reliable sodium-cooled reactor one of the more important problem areas is that of friction and wear of components immersed in liquid sodium or exposed to sodium vapor. Sodium coolant at elevated temperatures may severely affect most oxide-bearing surface layers which provide corrosion resistance and, to some extent, lubrication and surface hardness. Consequently, accelerated deterioration may be experienced on engaged-motion contact surfaces, which could result in unexpected reactor shutdown from component malfunction or failure due to galling and seizure. An overall view of the friction and wear phenomena encountered during oscillatory rubbing of surfaces in high-temperature, liquid-sodium environments is presented. Specific data generated at the Liquid Metal Engineering Center (LMEC) on this subject is also presented. (U.S.)

Coolant void effect investigation - case of a na-cooled fast reactor

International Nuclear Information System (INIS)

Glinatsis, G.; Gugiu, D.

2013-01-01

In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

International Nuclear Information System (INIS)

Carter, G.W.

1979-01-01

Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

Characteristics of liquid and boiling sodium flows in heating pin bundles

International Nuclear Information System (INIS)

Menant, Bernard

1976-01-01

This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged. (author) [fr

Development of a model to predict flow oscillations in low-flow sodium boiling. [Loss-of-Piping Integrity accidents

Energy Technology Data Exchange (ETDEWEB)

Levin, A.E.; Griffith, P.

1980-04-01

Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed.

Numerical investigation on turbulence mixing characteristics under thermal striping flows. Investigations on fluid temperature fluctuation phenomena in air and sodium

Energy Technology Data Exchange (ETDEWEB)

Murakami, Satoshi [Customer System Co. Ltd., Tokai, Ibaraki (Japan); Muramatsu, Toshiharu

1999-05-01

A three-dimensional thermal striping analysis was carried out using a direct numerical simulation code DINUS-3, for a coaxial jet configuration using air and sodium as a working fluid, within the framework of the EJCC thermo-hydraulic division. From the analysis, the following results have been obtained: (1) Calculated potential core length in air and sodium turbulence flows agreed with a theoretical value (5d - 7d ; d : diameter of jet nozzle) in the two-dimensional free jet theory. (2) Hydraulic characteristics in sodium flows as the potential core length can be estimated by the use of that of air flow characteristics. (3) Shorter thermally potential core length defined by spatial temperature distribution was evaluated in sodium flows, compared with that in air flows. This is due to the higher thermal conductivity of sodium. (4) Thermal characteristics in sodium flows as the thermally potential core length can not be evaluated, based on that air thermal characteristics. (author)

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

Conceptural design of multipurpose sodium test loop

International Nuclear Information System (INIS)

Kim, W.C.; Lee, Y.W.; Nam, H.Y.; Chun, S.Y.; Kim, J.; Yuh, M.W.

1982-01-01

This report describes the conceptural design of the multipurpose sodium test loop (MSTL). This MSTL consists mainly of impurity control and measurement system, corrosion and masstransfer system and heat transfer system. Problems associated with liquid sodium coolant will be studied and operating experiences will be obtained by the use of this facility. This technology will be used to evaluate safety and reliability of large sodium facility in the future. The total cost excluding the cost of building construction is estimated to 175 thousand dollars. (Author)

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

International Nuclear Information System (INIS)

Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira; Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

2015-01-01

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Graduate School of Engineering, The University of Tokyo, Ibaraki (Japan); Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki [Japan Atomic Energy Agency, Ibaraki (Japan)

2015-10-15

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

In-core failure of the instrumented BWR rod by locally induced high coolant temperature

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1985-12-01

In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

Energy Technology Data Exchange (ETDEWEB)

Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

2012-10-15

The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

International Nuclear Information System (INIS)

Genter, G.

1981-03-01

The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

International Nuclear Information System (INIS)

Thiele, R.; Anglart, H.

2014-01-01

This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

Advanced sodium fast reactor accident source terms :

Energy Technology Data Exchange (ETDEWEB)

Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

2010-09-01

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

International Nuclear Information System (INIS)

Frajndlich, R.; Sousa, J.A. de.

1985-01-01

A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author) [pt

Development of honeycomb type orifices for flow zoning in PFBR

Energy Technology Data Exchange (ETDEWEB)

Pandey, G.K., E-mail: gkpandey@igcar.gov.in; Ramdasu, D.; Padmakumar, G.; Prakash, V.; Rajan, K.K.

2013-09-15

Highlights: • Cavitation free flow zoning devices are developed for reactor core in PFBR. • These devices are experimentally investigated for their hydraulic characteristics. • Pressure drop and cavitation are two main characteristics to be investigated. • Various configurations of devices utilized in different zones are discussed. • Loss coefficient for each configuration is compared and reported. -- Abstract: The prototype fast breeder reactor (PFBR) is in its advanced phase of construction at Kalpakkam, India. It is a sodium cooled, pool type reactor with two loop concept where each loop have one primary sodium pump (PSP), one secondary sodium pump (SSP) and two intermediate heat exchangers (IHX). PFBR core subassemblies (SA) are supported vertically inside the sleeves provided in the grid plate (GP). The GP acts as a coolant header through which flow is distributed among the SA to remove fission heat. Since the power profile of the reactor core is not uniform, it is necessary to distribute the coolant flow (called flow zoning) to each subassembly according to their power levels to get maximum mean outlet temperature of sodium at core outlet. To achieve this, PFBR core is divided into 15 zones such as fuel, blanket, reflector, storage, etc. according to their respective power levels. The flow zoning in the different SAs of the reactor core is achieved by installing permanent pressure dropping devices in the foot of the subassembly. Orifices having honey-comb type geometry were developed to meet the flow zoning requirements of fuel zone. These orifices being of very complex geometry requires precision methods of manufacturing to achieve the desired shape under specified tolerances. Investment casting method was optimized to manufacture this orifice plate successfully. Hydraulics of these orifices is important in achieving the required pressure drop without cavitation. The pressure drop across these orifice geometries depends mainly on geometrical

Coolant and ambient temperature control for chillerless liquid cooled data centers

Science.gov (United States)

Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

2016-02-02

Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

Directory of Open Access Journals (Sweden)

Seon Oh Yu

2017-08-01

Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

Natural circulation in reactor coolant system

International Nuclear Information System (INIS)

Han, J.T.

1987-01-01

Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

Coolant clean-up system in the primary coolant circuit for nuclear reactor

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

Combined convective heat transfer of liquid sodium flowing across tube banks

International Nuclear Information System (INIS)

Ma, Ying; Sugiyama, Ken-ichiro; Ishiguro, Ryoji

1989-01-01

In order to clarify the heat transfer characteristics of combined convection of liquid sodium, a numerical analysis is performed for liquid sodium which flows through a single horizontal row of tubes in the direction of gravity. The correlation of heat transfer characteristics between liquid sodium and ordinary fluids is also discussed. The heat transfer characteristics at large Reynolds numbers are improved when the Richardson number is increased, and the improvement rate is enlarged with increase in p/d value, since convection effect is relatively large. However heat transfer coefficients do not differ from those of forced convection at small Reynolds numbers even when the Richardson number reaches a high value because of conduction effect. A good consistence of heat transfer characteristics of combined convection between liquid sodium and air is obtained at the same Peclet number and Richardson number. This means that the fundamental heat transfer characteristics of combined convection of liquid sodium can be investigated with ordinary fluids. (author)

Continuous method for refining sodium. [for use in LMFBR type reactors

Energy Technology Data Exchange (ETDEWEB)

Batoux, B; Laurent-Atthalin, A; Salmon, M

1973-11-16

The invention relates to a refining method according to which commercial sodium provides a high purity sodium with, in particular, a very small calcium content. The method consists in continuously feeding a predetermined amount of sodium peroxide into a sodium stream, mixing and causing said sodium peroxide to reach with sodium at an appropriate temperature, and, finally, separating the reaction products from sodium by decanting and filtering same. The thus obtained high purity sodium meets the requirements of atomic industries in particular, in view of its possible use as coolant in nuclear reactors of the ''breeder'' type.

LOFT advanced densitometer for nuclear loss-of-coolant experiments

International Nuclear Information System (INIS)

Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

1979-01-01

A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

Experimental investigation of the unsteady one-phase flow through perforated plates

International Nuclear Information System (INIS)

Casadei, F.

1982-07-01

The flow of the coolant through the perforated dip-plate during a hypothetical core-disruptive accident in a sodium-cooled fast breeder reactor was simulated in a one-dimensional model. Several experiments with water as fluid and with various perforation ratios of the dip-plate and different initial heights of the fluid head over the dip plate were run. The pressure drop through the dip-plate and the forces acting on the dip-plate and on the upper plug of the reactor vessel in a wide range of the Reynolds and Strouhal numbers were measured. The flow pattern downstreams the perforated plate was filmed with high-speed cameras. The resistance coefficients for the unsteady flow of the coolant through the perforated plate were obtained as a function of the acceleration. The forces acting on the upper plug and their time integral were compared with those acting on the dip-plate. Finally, using the high-speed film pictures the formation of fluid jets downstream the dip-plate was investigated. (orig.) [de

Removal of sodium from the component of the sodium purification loop

International Nuclear Information System (INIS)

Kim, Byung Ho; Jeong, Kyung Chai; Jeong, Ji Young; Kim, Jong Man; Choi, Byung Hae; Nam, Ho Yun

2005-01-01

The purpose of a cleaning process is to remove the residual sodium adhering to the component walls once it has been properly drained. It is necessary to clean and decontaminate a component, especially the large components of the primary coolant system; such as the intermediate heat exchangers and the primary pump. Improper and inadequate cleaning has in a number of cases resulted in problems in the storage, handling, and reuse of components. Several types of failures due to improper cleaning procedures have been defined in the past. Inadequate and incomplete removal of sodium results in residues which may contain metallic sodium and alkaline compounds such as sodium hydroxide, sodium oxide, sodium carbonate, and various types of alcoholates. Reinsertion of components containing these compounds into a high-temperature sodium system can result in either the intergranular penetration characteristic of a high-oxygen sodium or an accelerated corrosion due to oxygen. The methods used for cleaning sodium equipment depend on the condition and types of equipment to be cleaned and whether the equipment is to be reused. Cleaning methods are needed that will avoid a deleterious local overheating, material surface degradation or deposits, chemical, physical, or mechanical damage, and external effects. This paper discusses a steam-nitrogen gas cleaning method for the routine applications that permits the reuse of the cold trap in sodium

User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

Energy Technology Data Exchange (ETDEWEB)

Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

1979-10-01

The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

Coolant Chemistry Control: Oxygen Mass Transport in Lead Bismuth Eutectic

International Nuclear Information System (INIS)

Weisenburger, A.; Mueller, G.; Bruzzese, C.; Glass, A.

2015-01-01

In lead-bismuth cooled transmutation systems, oxygen, dissolved in the coolant at defined quantities, is required for stable long-term operation by assuring the formation of protective oxide scales on structural steel surfaces. Extracted oxygen must be permanently delivered to the system and distributed in the entire core. Therefore, coolant chemistry control involves detailed knowledge on oxygen mass transport. Beside the different flow regimes a core might have stagnant areas at which oxygen delivery can only be realised by diffusion. The difference between oxygen transport in flow paths and in stagnant zones is one of the targets of such experiments. To investigate oxygen mass transport in flowing and stagnant conditions, a dedicated facility was designed based on computational fluid dynamics (CFD). CFD also was applied to define the position of oxygen sensors and ultrasonic Doppler velocimetry transducers for flow measurements. This contribution will present the test facility, design relevant CFD calculations and results of first tests performed. (authors)

JOYO coolant sodium and cover gas purity control database (MK-II core)

International Nuclear Information System (INIS)

Ito, Kazuhiro; Nemoto, Masaaki

2000-03-01

The experimental fast reactor 'JOYO' served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon dioxide, methane and helium in argon gas with the reactor condition. (author)

Design of creep machine and creep specimen chamber for carrying out creep tests in flowing liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Jayakumar, T.

2014-02-15

Highlights: • Design of a lever type creep machine for carrying out creep test in flowing sodium. • Leveling of lever during creep was achieved by automated movement of fulcrum. • Design of creep chamber for providing constant sodium flow rate across creep specimen. • Minimum use of bellow in chamber for sodium containment and mechanical isolation. • Mini-lever mechanism to counter balance load reduction on specimen due to bellow stiffness. - Abstract: A creep testing system has been designed, fabricated, installed and validated for carrying out creep tests in flowing liquid sodium. The testing system consists of two sections namely creep testing machine and an environmental chamber. The testing system has the ability of (i) applying tensile load to the test specimen through a lever, (ii) monitoring continuously the creep elongation and (iii) allowing sodium to flow around the creep specimen at constant velocity. The annular space between the creep specimen and the environmental chamber has been suitably designed to maintain constant sodium flow velocity. Primary and secondary bellows are employed in the environmental chamber to (i) mechanically isolate the creep specimen, (ii) prevent the flowing sodium in contact with air and (iii) maintain an argon gas cover to the leaking sodium if any from primary bellow, with a provision to an alarm get activated by a spark plug. The lever-horizontality during creep test has been maintained by automatically lifting up the fulcrum instead of lowering down the pull rod as conventionally used. A mini lever mechanism has been incorporated in the load train to counter balance the load reduction on specimen from the changing stiffness of the bellows. The validation of the testing system has been established by carrying out creep tests on 316L(N) stainless steel at 873 K over a wide stress range and comparing the results with those obtained in air by employing the developed and conventional creep testing machines.

Numerical investigation on critical heat flux and coolant volume required for transpiration cooling with phase change

International Nuclear Information System (INIS)

He, Fei; Wang, Jianhua

2014-01-01

Highlights: • Five states during the transpiration cooling are discussed. • A suit of applicable program is developed. • The variations of the thickness of two-phase region and the pressure are analyzed. • The relationship between heat flux and coolant mass flow rate is presented. • An approach is given to define the desired case of transpiration cooling. - Abstract: The mechanism of transpiration cooling with liquid phase change is numerically investigated to protect the thermal structure exposed to extremely high heat flux. According to the results of theoretical analysis, there is a lower critical and an upper critical external heat flux corresponding a certain coolant mass flow rate, between the two critical values, the phase change of liquid coolant occurs within porous structure. A strongly applicable self-edit program is developed to solve the states of fluid flow and heat transfer probably occurring during the phase change procedure. The distributions of temperature and saturation in these states are presented. The variations of the thickness of two-phase region and the pressure including capillary are analyzed, and capillary pressure is found to be the main factor causing pressure change. From the relationships between the external heat flux and coolant mass flow rate obtained at different cooling cases, an approach is given to estimate the maximal heat flux afforded and the minimal coolant consumption required by the desired case of transpiration cooling. Thus the pressure and coolant consumption required in a certain thermal circumstance can be determined, which are important in the practical application of transpiration cooling

Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

International Nuclear Information System (INIS)

Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

1979-01-01

A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

Sodium flow distribution in test fuel assembly P-23B

International Nuclear Information System (INIS)

Taylor, J.P.S.

1978-08-01

Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

Chaotic Dynamos Generated by a Turbulent Flow of Liquid Sodium

International Nuclear Information System (INIS)

Ravelet, F.; Monchaux, R.; Aumaitre, S.; Chiffaudel, A.; Daviaud, F.; Dubrulle, B.; Berhanu, M.; Fauve, S.; Mordant, N.; Petrelis, F.; Bourgoin, M.; Odier, Ph.; Plihon, N.; Pinton, J.-F.; Volk, R.

2008-01-01

We report the observation of several dynamical regimes of the magnetic field generated by a turbulent flow of liquid sodium (VKS experiment). Stationary dynamos, transitions to relaxation cycles or to intermittent bursts, and random field reversals occur in a fairly small range of parameters. Large scale dynamics of the magnetic field result from the interactions of a few modes. The low dimensional nature of these dynamics is not smeared out by the very strong turbulent fluctuations of the flow

Theoretical studying the stability of steady-state regime of a channel with a coolant condensation

International Nuclear Information System (INIS)

Savikhin, O.G.

1987-01-01

Based on the boiling channel stability theory, the channel steady-state stability with the coolant condensation is studied. Condensable coolants are used in the NPP steam-separator superheaters as well as in cryogenic technique. Under certain conditions the coolant flow rate and temperature fluctuations may be excited in the parallel channel system with coolant condensation, which produce a sufficient effect on the heat exchange equipment operation reliability. To describe unsteady processes of heat and mass transfer in the channel, a homogeneous two-phase flow one dimensional model is used. The results obtained allow one to make a conclusion concerning the effect of some parameters on condensing channel steady-state regime stability: reduction of inlet and outlet unheated communication length, pressure drop increase at the outlet plate and its reduction at the inlet one lead to the increase of stability margin

Experimental study on combustion and suppression characteristics of sodium fire in a columnar flow using extinguishing powder

International Nuclear Information System (INIS)

Huo Yan; Zhang Zhigang; Li Jinke; Liu Zhongkun; Ma Yaolong

2017-01-01

In the operation of the sodium-cooled fast reactor, the leakage and fire accident of liquid sodium is common and it is frequent in sodium-related facilities. This study focuses on the combustion and suppression characteristics of sodium fire in a columnar flow. Liquid sodium (250°C) is injected into a 7.9 m"3 cylindrical chamber at a flow rate of about 1.0 m"3/h to create a columnar sodium fire, and 18.4 kg class D extinguishing powder is sprayed after the liquid sodium injection. The temperature in the chamber space and sodium collection plate and the heat release rate from sodium fire are measured and analyzed. Based on the temperature data the sodium fire under suppression could be divided into four phases of dropping sharply, continuously remaining lower, rising and declining mildly, and depressing. The sodium fire in the space could be suppressed and cooled down if the extinguishing agent could spray in the early period of the liquid sodium injection. The extinguishing agent could suppress the combustion and spreading of liquid sodium dropping on the collection plate, limit the pool combustion area and postpone the commencement of sodium pool burning in spite of its later re-ignition happening. This study promises to evaluate the combustion and suppression characteristics of sodium fire in the sodium-related facilities. (author)

Sound velocity in the coolant of boiling nuclear reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

Feasibility study on the type of KALIMER coolant circulation pump

International Nuclear Information System (INIS)

Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

1997-07-01

The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

Reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere. Fundamental study on sodium carbonate process in FBR bulk sodium coolant disposal technology

International Nuclear Information System (INIS)

Tadokoro, Yutaka; Yoshida, Eiichi

1999-11-01

A sodium carbonate processing method, which changes sodium to sodium carbonate and/or sodium bicarbonate by humid carbon dioxide, has been examined and about to be applied to large test loops dismantling. However, that the basic data regarding the progress of the reaction is insufficient on the other hand, is a present condition. The present report therefore aims at presenting basic data regarding the reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere, and observing the reaction progress, for the application to large test loops dismantling. The test result is summarized as follows. (1) Although the reaction velocity of sodium varied with sodium specimen sizes and velocity measurement methods, the reaction velocity of sodium hydration was in about 0.16 ∼ 0.34 mmh -1 (0.016 ∼ 0.033g cm -2 h -1 , 6.8x10 -4 ∼ 1.4x10 -3 mol cm -2 h -1 ) and that of sodium carbonation was in about 0.16 ∼ 0.27mmh -1 (0.016 ∼ 0.023g cm -2 h -1 , 6.8x10 -4 ∼ 1.1x10 -3 mol cm -2 h -1 ) (26 ∼ 31degC, RH 100%). (2) The reaction velocity of sodium in carbon dioxide atmosphere was greatly affected by vapor partial pressure (absolutely humidity). And the velocity was estimated in 0.08 ∼ 0.12mmh -1 (0.008 ∼ 0.012g cm -2 h -1 , 3.4x10 -4 ∼ 5.2x10 -4 mol cm -2 h -1 ) in the carbon dioxide atmosphere, whose temperature of 20degC and relative humidity of 80% are assumed real sodium carbonate process condition. (3) By the X-ray diffraction method, NaOH was found in humid air reaction product. Na 2 CO 3 , NaHCO 3 were found in carbon dioxide atmosphere reaction product. It was considered that Sodium changes to NaOH, and subsequently to NaHCO 3 through Na 2 CO 3 . (4) For the application to large test loops dismantling, it is considered possible to change sodium to a target amount of sodium carbonate (or sodium bicarbonate) by setting up gas supply quantity and also processing time appropriately according to the surface area

The effect of variations in carbon activity on the carburization of austenitic steels in sodium

International Nuclear Information System (INIS)

Gwyther, J.R.; Hobdell, M.R.; Hooper, A.J.

1978-07-01

Experience has shown that the liquid sodium coolant of fast breeder reactors is an effective carbon-transport medium; the resulting carburization of thin austenitic stainless steel components (eg IHX and fuel cladding) could adversely affect their mechanical integrity. The degree and nature of steel carburization depend, inter alia, on the carbon activity of the sodium environment. Exploratory tests are described in which specimens of austenitic stainless steel were carburized in sodium, the carbon activity of which was continuously monitored by a BNL electrochemical carbon meter. The sodium carbon activity was initially high, but decreased with time, simulating conditions equivalent to plant start-up or coolant clean-up following accidental oil ingress. The extent and nature of steel carburization was identified by metallography, electron microscopy, X-ray crystallography and chemical analysis. (author)

Production and release of gas and volatile elements from sodium-based targets

CERN Multimedia

Plewinski, F; Wildner, E; Catherall, R

Several large scale facilities being studied for Europe use sodium or a sodium-based alloy either as a target or as a coolant for heavier solid targets subjected to MW proton beams, such as the European Spallation Source (ESS) and $\\beta$-beam projects. ESS will be the neutron source in use from the year 2020 in Europe, providing high intensity neutron fluxes over large energy spectra ( from 10$^{-3}$ eV to 10$^{3}$ eV) to scientists, to explore materials from 10$^{-2}$m to 10$^{-16}$m scale. A sodium-cooled array of tungsten blocks is one of the potential solutions for the target that will convert protons from the 5 MW 2.5 GeV linac into neutrons. Sodium is a tried and tested coolant in fast nuclear reactors with associated technologies and design standards. Its application to a spallation environment however remains to be validated. The ISOLDE facility is well placed to perform detailed measurements of radioisotopes produced in sodium with a proton beam whose energy of 1.4 GeV is very close to the ESS base...

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.

Sodium removal from Hallam Reactor components

International Nuclear Information System (INIS)

Huntsman, L.K.; Meservey, R.H.

1979-08-01

This report discussed the removal of sodium from major components of the Hallam Nuclear Power Facility. This facility contained the experimental ractor used to test the feasibility of sodium coolant. The Idaho Operations Office of the Department of Energy assigned EG and G Idaho, Inc., the task of carrying out this decontamination and decommissioning program at the Idaho National Engineering Laboratory (INEL). Since their shipment to the INEL from Lincoln, Nebraska in 1968, the Hallam Reactor components had been stored in inert nitrogen to prevent the sodium in the components from reacting with moisture in the air. The procedure used to react the sodium in the components and to decontaminate them is discussed. Problems and unusual occurrences in the decontamination and decommissioning process are also reported

A review of flow analysis methods for determination of radionuclides in nuclear wastes and nuclear reactor coolants.

Science.gov (United States)

Trojanowicz, Marek; Kołacińska, Kamila; Grate, Jay W

2018-06-01

The safety and security of nuclear power plant operations depend on the application of the most appropriate techniques and methods of chemical analysis, where modern flow analysis methods prevail. Nevertheless, the current status of the development of these methods is more limited than it might be expected based on their genuine advantages. The main aim of this paper is to review the automated flow analysis procedures developed with various detection methods for the nuclear energy industry. The flow analysis methods for the determination of radionuclides, that have been reported to date, are primarily focused on their environmental applications. The benefits of the application of flow methods in both monitoring of the nuclear wastes and process analysis of the primary circuit coolants of light water nuclear reactors will also be discussed. The application of either continuous flow methods (CFA) or injection methods (FIA, SIA) of the flow analysis with the β-radiometric detection shortens the analysis time and improves the precision of determination due to mechanization of certain time-consuming operations of the sample processing. Compared to the radiometric detection, the mass spectrometry (MS) detection enables one to perform multicomponent analyses as well as the determination of transuranic isotopes with much better limits of detection. Copyright © 2018 Elsevier B.V. All rights reserved.

Experimental investigations and modelling of sodium-concrete interaction

International Nuclear Information System (INIS)

Schultheiss, G.F.; Deeg, H.J.

1990-01-01

The use of sodium as a coolant in liquid metal fast breeder reactors, fusion reactors, and solar plants requires special consideration of its chemical reactivity and related safety problems in the case of sodium leckage. On contact between hot sodium and concrete an interaction takes place resulting in energy release and hydrogen generation, which may contribute to containment loading by pressurization in a hypothetical accident situation. For this reason, sodium-concrete interactions were investigated experimentally and theoretically. The experiments revealed important effects of quartzitic material within the concrete and of the sodium temperature on the interaction mechanisms, the energy release and the consequent hydrogen production. The numerical model shows good agreement with the experimental results. (orig.) [de

Investigations on flow reversal in stratified horizontal flow

International Nuclear Information System (INIS)

Staebler, T.; Meyer, L.; Schulenberg, T.; Laurien, E.

2005-01-01

The phenomena of flow reversal in stratified flows are investigated in a horizontal channel with application to the Emergency Core Cooling System (ECCS) in Pressurized Water Reactors (PWR). In case of a Loss-of-Coolant-Accident (LOCA), coolant can be injected through a secondary pipe within the feeding line of the primary circuit, the so called hot leg, counter-currently to the steam flow. It is essential that the coolant reaches the reactor core to prevent overheating. Due to high temperatures in such accident scenarios, steam is generated in the core, which escapes from the reactor vessel through the hot leg. In case of sufficiently high steam flow rates, only a reduced amount of coolant or even no coolant will be delivered to the reactor core. The WENKA test facility at the Institute for Nuclear and Energy Technologies (IKET) at Forschungszentrum Karlsruhe is capable to investigate the fluid dynamics of two-phase flows in such scenarios. Water and air flow counter-currently in a horizontal channel made of clear acrylic glass to allow full optical access. Flow rates of water and air can be varied independently within a wide range. Once flow reversal sets in, a strong hysteresis effect must be taken into account. This was quantified during the present investigations. Local experimental data are needed to expand appropriate models on flow reversal in horizontal two-phase flow and to include them into numerical codes. Investigations are carried out by means of Particle Image Velocimetry (PIV) to obtain local flow velocities without disturbing the flow. Due to the wavy character of the flow, strong reflections at the interfacial area must be taken into account. Using fluorescent particles and an optical filter allows eliminating the reflections and recording only the signals of the particles. The challenges in conducting local investigations in stratified wavy flows by applying optical measurement techniques are discussed. Results are presented and discussed allowing

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

International Nuclear Information System (INIS)

Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

2012-01-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

Energy Technology Data Exchange (ETDEWEB)

Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

2012-07-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

Corrosion phenomena in sodium-potassium coolant resulting from solute interaction in multicomponent solution

Science.gov (United States)

Krasin, V. P.; Soyustova, S. I.

2018-03-01

The solubility of Fe, Cr, Ni, V, Mn and Mo in sodium-potassium melt has been calculated using the mathematical framework of pseudo-regular solution model. The calculation results are compared with available published experimental data on mass transfer of components of austenitic stainless steel in sodium-potassium loop under non-isothermal conditions. It is shown that the parameters of pair interaction of oxygen with transition metal can be used to predict the corrosion behavior of structural materials in sodium-potassium melt in the presence of oxygen impurity. The results of calculation of threshold concentration of oxygen of ternary oxide formation of sodium with transitional metals (Fe, Cr, Ni, V, Mn, Mo) are given in conditions when pure solid metal comes in contact with sodium-potassium melt.

Sodium confluent rates of flow values, on 0,5 mixer, of a sodium italian SS-050 circuit component

International Nuclear Information System (INIS)

Walsh, L.M.

1987-01-01

Sodium lines on different temperatures, during an emergency drainage on 0,5 mixer was found. To future valuation by DIMEC of tensions that occurs on that component of SS-050, the confluent rates of flow values were calculated. (L.M.J.) [pt

Analysis of PWR auxiliary coolant: determination of chloride in borax/nitrite solution by known addition - known dilution potentiometry

International Nuclear Information System (INIS)

Midgley, D.; Gatford, C.

1989-11-01

Chloride concentrations of 75-250 μg 1 -1 have been determined in simulated PWR auxiliary coolant containing 1000 mg l -1 each of sodium tetraborate and sodium nitrite. The effects of the two main components of the coolant solution on a variety of chloride-selective electrodes have been studied. Sodium tetraborate posed no problem except through its effect on the pH, which is easily adjusted. Such high concentrations of nitrite, however, caused significant deviations in e.m.f. for all the electrodes and marked tarnishing of the electroactive membrane after only one or two measurements. Sulphamic acid was selected as the best means of removing nitrite and silver chloride electrodes were preferred over mercury(I) chloride electrodes because of their greater robustness in the conditions. At these chloride concentrations, the electrodes are operating in their non-Nernstian response regions and direct potentiometry has poor precision, even if standards could be successfully matched to samples containing such high concentrations of background material. Known addition - known dilution potentiometry was adopted, with internal calibration for both slope factor and standard potential. (author)

Decreased cerebral blood flow after administration of sodium bicarbonate in the distressed newborn infant

DEFF Research Database (Denmark)

Lou, H C; Lassen, N A; Fris-Hansen, B

1978-01-01

In the course of our studies on cerebral blood flow in newborn infants, we have observed a striking depressing effect of sodium bicarbonate infusion on cerebral blood flow which in some cases may severely aggravate cerebral ischemia. We measured cerebral blood flow before and after the treatment...... with 1 to 8 meqs of sodium bicarbonate in seven distressed newborn infants. The 133 Xe clearance technique was used. The results showed in six of the seven cases a decrease in cerebral blood flow, which in most cases was reduced to 14 to 22 ml/100 g/min, which is about half the value prior...... to the bicarbonate infusion. In one case an extreme reduction occurred: cerebral blood flow was reduced to 3 ml/100 g/min, well below the level compatible with tissue survival. The results are discussed with regard to the optimal treatment of the acidotic newborn....

Modeling the spatial distribution of the parameters of the coolant in the reactor volume

International Nuclear Information System (INIS)

Nikonov, S.P.

2011-01-01

In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

Coolant controls of a PEM fuel cell system

Science.gov (United States)

Ahn, Jong-Woo; Choe, Song-Yul

When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2011-07-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

2011-01-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

Development of early core anomaly detection system by using in-sodium microphone in JOYO. Fundamental characteristics test of in-sodium microphone in water and examination of improvement of detection accuracy

International Nuclear Information System (INIS)

Komai, Masafumi

2001-07-01

Fast reactor core anomalies can be detected in near real-time with acoustic sensors. An acoustic detection system senses an in-core anomaly immediately from the fast acoustic signals that propagate through the sodium coolant. One example of a detectable anomaly is sodium boiling due to local blockage in a sub-assembly; the slight change in background acoustic signals can be detected. A key advantage of the acoustic detector is that it can be located outside the core. The location of the anomaly in the core can be determined by correlating multiple acoustic signals. This report describes the testing and fundamental characteristics of a microphone suitable for use in the sodium coolant and examines methods to improve the system's S/N ratio. Testing in water confirmed that the in-sodium microphone has good impulse and wide band frequency responses. These tests used impulse and white noise signals that imitate acoustic signals from boiling sodium. Correlation processing of multiple microphone signals to improve S/N ratio is also described. (author)

Niobium corrosion in flowing liquid sodium between 400 and 600 degrees C

International Nuclear Information System (INIS)

Sannier, J.; Champeix, L.; Darras, R.; Graff, W.

1966-10-01

The corrosion of niobium and two of its alloys was studied under temperature, flow rate, and purity conditions of liquid sodium similar to those likely to occur in a fast neutron reactor. The results are discussed with reference to the following parameters: purification method used for the sodium, temperature, metallurgical condition of the structural metal. Generally speaking, an important role is played by the oxygen content of the liquid metal towards the corrosion of the niobium: although the metal behaves very satisfactorily when a hot trap purification is used, it undergoes corrosion in the presence of sodium which has been purified only by a cold trap. (authors) [fr

Heat exchanger with oscillating flow

Science.gov (United States)

Scotti, Stephen J. (Inventor); Blosser, Max L. (Inventor); Camarda, Charles J. (Inventor)

1993-01-01

Various heat exchange apparatuses are described in which an oscillating flow of primary coolant is used to dissipate an incident heat flux. The oscillating flow may be imparted by a reciprocating piston, a double action twin reciprocating piston, fluidic oscillators or electromagnetic pumps. The oscillating fluid flows through at least one conduit in either an open loop or a closed loop. A secondary flow of coolant may be used to flow over the outer walls of at least one conduit to remove heat transferred from the primary coolant to the walls of the conduit.

Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

International Nuclear Information System (INIS)

Hermansky, Jozef; Zavodsky, Martin

2014-01-01

A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

Secondary seal effects in hydrostatic non-contact seals for reactor coolant pump shaft

International Nuclear Information System (INIS)

Fujita, T.; Koga, T.; Tanoue, H.; Hirabayashi, H.

1987-01-01

The paper presents a seal flow analysis in a hydrostatic non-contact seal for a PWR coolant pump shaft. A description is given of the non-contact seal for the reactor coolant pump. Results are presented for a distortion analysis of the seal ring, along with the seal flow characteristics and the contact pressure profiles of the secondary seals. The results of the work confirm previously reported findings that the seal ring distortion is sensitive to the o-ring location (which was placed between the ceramic seal face and the seal ring retainer). The paper concludes that the seal flow characteristics and the tracking performance depend upon the dynamic properties of the secondary seal. (U.K.)

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

Method of injecting iron ion into reactor coolant

International Nuclear Information System (INIS)

Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

1988-01-01

Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

International Nuclear Information System (INIS)

Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

2009-01-01

In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

International Nuclear Information System (INIS)

Bordner, G.L.

1979-10-01

The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

Generation of a Magnetic Field by Dynamo Action in a Turbulent Flow of Liquid Sodium

International Nuclear Information System (INIS)

Monchaux, R.; Chiffaudel, A.; Daviaud, F.; Dubrulle, B.; Gasquet, C.; Marie, L.; Ravelet, F.; Berhanu, M.; Fauve, S.; Mordant, N.; Petrelis, F.; Bourgoin, M.; Moulin, M.; Odier, Ph.; Pinton, J.-F.; Volk, R.

2007-01-01

We report the observation of dynamo action in the von Karman sodium experiment, i.e., the generation of a magnetic field by a strongly turbulent swirling flow of liquid sodium. Both mean and fluctuating parts of the field are studied. The dynamo threshold corresponds to a magnetic Reynolds number R m ∼30. A mean magnetic field of the order of 40 G is observed 30% above threshold at the flow lateral boundary. The rms fluctuations are larger than the corresponding mean value for two of the components. The scaling of the mean square magnetic field is compared to a prediction previously made for high Reynolds number flows

Experiments on the behaviour of thermite melt injected into sodium: Final report on the THINA test results

International Nuclear Information System (INIS)

Huber, F.; Kaiser, A.; Peppler, W.

1994-01-01

During hypothetical accidents of fast breeder reactors the core melts and part of the core material inventory is ejected into the upper coolant plenum. As a consequence, a fuel to coolant thermal interaction occurs between the melt and the sodium. A series of simulating experiments was carried out in KfK/IRS to improve the knowledge about the phenomenology of molten fuel/coolant interactions and to support theoretical work on the safety of fast breeder reactors. In the tests, a thermite melt of up to 3270 K is injected from below into a sodium pool the temperature of which is between 770 and 820 K. The masses of the melt and the sodium are about five and 150 kg, respectively. Thermal interactions have been observed to occur as a sequence of small local pressure events mainly during the melt injection. Large-scale vapour explosions have not been observed. Generally, the conversion ratios of thermal to mechanical energy have been low. (author)

Technical and QA plan: Boiling behavior during flow instability

International Nuclear Information System (INIS)

Coutts, D.A.

1991-01-01

The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may occur. This localized boiling can restrict coolant flow and the ability to transfer heat out of the reactor system. The maximum operating power for the reactor may be limited by how the coolant system reacts to a flow instability. One of the methods to assure safe operation during a reducing flow transient, is to operate at a power level below that necessary to initiate a flow excursion. Several correlations have been used to predict the conditions which will proceed a flow excursion. These correlations rely on the steady state behavior of the coolant and are based on steady-state testing. There are two significant points which this project will try to identify. The first is when vapor first forms on the channel surface. This might be designated as the Nucleate Vapor Transition. (Steady state equivalent is ONB). The second is when the vapor formation rate is large enough to lead to flow instability and thermal excursion. This point might be designated as the Significant Vapor Transition. (Steady state equivalent is OSV). A correlation will be developed to relate established steady state relations with the behavior of transient systems

A miniature inductive temperature sensor to monitor temperature noise in the coolant of an LMFBR

International Nuclear Information System (INIS)

Dean, S.A.; Sandham, C.W.

1980-01-01

A description is given of the design and performance of miniature inductive sensors developed to monitor fast temperature fluctuations in the sodium coolant above the core of a LMFBR. These instruments, designed to be installed within existing thermocouple containment thimbles, also provide a steady-state temperature indication for reactor control purposes. (author)

Secondary coolant purification system

International Nuclear Information System (INIS)

Stiteler, F.Z.; Donohue, J.P.

1978-01-01

The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

Coolant channel module CCM

International Nuclear Information System (INIS)

Hoeld, Alois

2007-01-01

A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

The development of NRTM-turbine flow meter and measurement of the coolant flow rate in-core of 5 MW heating reactor

International Nuclear Information System (INIS)

Zha Meisheng; Wang Xiuqin; Ni Mengchen

1995-01-01

In order to measure the coolant flow rate in-core of 5 MW Heating Reactor the special turbine flowmeter of the type of NRTM has been developed. It consists of a body, a turbine with long screw blade and six pieces of Alnico magnets, and a coil mounted on the body. The advantage of this turbine flowmeter is of low resistance and long working-life. Another advantage is that when the turbine is working or not working its factor of resistance is about the same. It is very important for a natural circulation heating reactor. Because the cable, which is welded to the coil assembly, is long enough to extend out of the reactor vessel to the control room, the signal of flow rate is easy to be disturbed by noise in the case. The traditional method of counting the frequency of the A-C voltage which is induced in the coil has a poor ability for resisting noise. The method of the frequency-spectrum analysis of the frequency of the A-C voltage is used to make sure the accuracy of the measurement of the turbine flow meter. Compared with the method of the count it has a good ability for resisting noise. After three years operation a lot of valuable data were obtained

Coolant flow monitoring in a PWR core using noise analysis

International Nuclear Information System (INIS)

Kostic, Lj.

1992-01-01

Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

RETRAN analysis of inter-system LOCA within the primary coolant pump

International Nuclear Information System (INIS)

Gangadharan, A.; Pratt, G.F.

1992-01-01

One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

HANARO secondary coolant management

International Nuclear Information System (INIS)

Kim, Seon Duk.

1998-02-01

In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

A study on sodium-concrete reaction

Energy Technology Data Exchange (ETDEWEB)

Pae, Jae Huem; Min, Byung Hoon; Lee, Joon Sik; Lee, Choong Hui; Chung, Ki Hong; Keum, Choong Ki [Suwon University, Suwon (Korea, Republic of)

1994-07-15

Sodium is commonly used as a coolant in liquid metal reactor. A large amount of its leakage may be possible in hypothetical accidents, even though the possibility is very low. In case that the leaked hot sodium comes in direct contact with structural concrete of liquid metal reactor, the reactor`s integrity can be challenged by the rupture of structure materials, hydrogen generation and its explosion, and release of radioactive aerosols due to sodium-concrete reaction. The knowledge of sodium-concrete reaction is evaluated to be one of the important and indispensable technologies for the establishment of safety measure in liquid metal reactor. In this study, the experimental facility of sodium-concrete reaction is to be designed, constructed and operated. And the reaction phenomena of sodium-concrete reaction is also to be analyzed through the experimental results. The aim of this study is to establish the measure of safety and protection for sodium-related facilities and to secure one of the fundamental technologies of liquid metal reactor safety. 47 refs., 7 figs., 13 tab.

Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

Directory of Open Access Journals (Sweden)

Istvan Farkas

2016-08-01

Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

Simplified numerical simulation of hot channel in sodium cooled reactor

International Nuclear Information System (INIS)

Fonseca, F. de A.S. da; Silva Filho, E.

1988-12-01

The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

[Influence of non-sodium restricted diet with diuretics on plasma rennin, renal blood flow and in patients with cirrhotic ascites].

Science.gov (United States)

Zhu, Yin-fang; Gu, Xi-bing; Zhu, Hong-ying; Yang, Xiao-juan; Wang, Dong; Yu, Ping

2013-02-01

To explore influence of sodium restricted diet and non-sodium restricted diet on plasma rennin (PRA), angiotensin II (All), ALD, renal blood flow (RBF) and subside of ascites in patients with cirrhotic ascites. Eighty cases of hepatitis B with cirrhotic ascites were randomly divided into sodium restricted diet group and non-sodium restricted diet group. 39 cases were in non-sodium restricted diet group, taking sodium chloride 6500-8000 mg daily; 41 cases were in sodium restricted diet group, taking sodium chloride 5000 mg daily. Both groups received diuretics furosemide and spironolactone. Blood sodium, urine sodium, PRA, AII, ALD, RBF ascites subsiding were compared after treatment. In non-sodium restricted diet group, blood sodium and urine sodium increased 10 days after treatment compared with those before treatment, and compared with those of sodium restricted diet group 10 days after treatment, P Renal damage induced by low blood sodium after treatment was less in non-sodium restricted diet group than that in sodium restricted diet group, P blood sodium, thus increasing excretion of urine sodium and diuretic effect. It can also decrease levels of PRA, AII and ALD, increase renal blood flow and prevent renal damage induced by low blood sodium and facilitate subsiding of ascites.

Investigation of coolant mixing in WWER-440/213 RPV with improved turbulence model

International Nuclear Information System (INIS)

Kiss, B.; Aszodi, A.

2011-01-01

A detailed and complex RPV model of WWER-440/213 type reactor was developed in Budapest University of Technology and Economics Institute of Nuclear Techniques in the previous years. This model contains the main structural elements as inlet and outlet nozzles, guide baffles of hydro-accumulators coolant, alignment drifts, perforated plates, brake- and guide tube chamber and simplified core. With the new vessel model a series of parameter studies were performed considering turbulence models, discretization schemes, and modeling methods with ANSYS CFX. In the course of parameter studies the coolant mixing was investigated in the RPV. The coolant flow was 'traced' with different scalar concentration at the inlet nozzles and its distribution was calculated at the core bottom. The simulation results were compared with PAKS NPP measured mixing factors data (available from FLOMIX project. Based on the comparison the SST turbulence model was chosen for the further simulations, which unifies the advantages of two-equation (kω and kε) models. The most widely used turbulence models are Reynolds-averaged Navier-Stokes models that are based on time-averaging of the equations. Time-averaging filters out all turbulent scales from the simulation, and the effect of turbulence on the mean flow is then re-introduced through appropriate modeling assumptions. Because of this characteristic of SST turbulence model a decision was made in year 2011 to investigate the coolant mixing with improved turbulence model as well. The hybrid SAS-SST turbulence model was chosen, which is capable of resolving large scale turbulent structures without the time and grid-scale resolution restrictions of LES, often allowing the use of existing grids created for Reynolds-averaged Navier-Stokes simulations. As a first step the coolant mixing was investigated in the downcomer only. Eddies are occurred after the loop connection because of the steep flow direction change. This turbulent, vertiginous flow was

Cooling Characteristics of the V-1650-7 Engine. II - Effect of Coolant Conditions on Cylinder Temperatures and Heat Rejection at Several Engine Powers

Science.gov (United States)

Povolny, John H.; Bogdan, Louis J.; Chelko, Louis J.

1947-01-01

An investigation has been conducted on a V-1650-7 engine to determine the cylinder temperatures and the coolant and oil heat rejections over a range of coolant flows (50 to 200 gal/min) and oil inlet temperatures (160 to 2150 F) for two values of coolant outlet temperature (250 deg and 275 F) at each of four power conditions ranging from approximately 1100 to 2000 brake horsepower. Data were obtained for several values of block-outlet pressure at each of the two coolant outlet temperatures. A mixture of 30 percent by volume of ethylene glycol and 70-percent water was used as the coolant. The effect of varying coolant flow, coolant outlet temperature, and coolant outlet pressure over the ranges investigated on cylinder-head temperatures was small (0 deg to 25 F) whereas the effect of increasing the engine power condition from ll00 to 2000 brake horsepower was large (maximum head-temperature increase, 110 F).

Simulation of IVR-ERVC and estimation method of coolant inflow to the cavity

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyunjin; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

In this study, the temperature distribution outside of RV wall and evaporation rate due to heat from core will be investigated. Using the universal analysis program ANSYS Fluent, the natural convection in the cavity for IVR-ERVC conditions were modelled and performed for heat transfer analysis. The aim of this study is to calculate the appropriate coolant flow so that coolant level in the cavity can be maintained at prescribed level and vessel wall temperature distribution, including RV outside wall temperature are also investigated. Reactor vessel and cavity in case of ex-vessel cooling for severe accident condition were modeled with and without insulators. The heat load into reactor vessel from corium inside of reactor lower head were obtained from MELCORE analysis and used as input B.C of CFD analysis. The Temperature gradient of reactor outer surface and evaporation rate of cooling eater was obtained from the analysis. These results can be used for further analysis of reactor vessel creep behavior and the estimate the coolant flow rate into the reactor cavity.. and The result can be used to verify the natural convection phenomena in the cavity and also to set the design parameters of cavity and coolant flow rate. The vessel outer surface temperature gradient can be also used to more accurate investigation of vessel creep behavior during severe accident condition, The result can also be used set up a strategy for severe accident managements.

Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Yamaguchi, Akira

2002-02-01

A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

PUMP: analog-hybrid reactor coolant hydraulic transient model

International Nuclear Information System (INIS)

Grandia, M.R.

1976-03-01

The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2016-01-01

Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

International Nuclear Information System (INIS)

Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

2014-01-01

The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

Influence of flowing sodium on creep deformation and rupture behaviour of 316L(N) austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

Ravi, S., E-mail: sravi@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Laha, K.; Mathew, M.D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Vijayaraghavan, S.; Shanmugavel, M.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2012-08-15

The influence of flowing sodium on creep deformation and rupture behaviour of AISI 316L(N) austenitic stainless steel has been investigated at 873 K over a stress range of 235-305 MPa. The results were compared with those obtained from testing in air environment. The steady state creep rates of the material were not influenced appreciably by the testing environments. The time to onset of tertiary stage of creep deformation was delayed in sodium environment. The creep-rupture lives of the material increased in sodium environment, which became more pronounced at lower applied stresses. The increase in rupture life of the material in flowing sodium was accompanied by an increase in rupture ductility. The creep damage on specimen surface as well as inside the specimen was less in specimen tested in sodium. SEM fractographic investigation revealed predominantly transgranular dimple failure for the specimen tested in sodium, whereas predominantly intergranular creep failure was observed in the air tested specimens. Almost no oxidation was observed in the specimens creep tested in the sodium environment. Absence of oxidation and less creep damage cavitation extended the secondary state in liquid sodium tests and lead to increase in creep rupture life and ductility of the material as compared to in air.

Research of the fluid flow in a radially orientated coolant channel of a turbine blade; Untersuchung der Stroemung in einem radial gerichteten Kuehlkanal eines Turbinenlaufrades

Energy Technology Data Exchange (ETDEWEB)

Hein, O.

1999-07-01

Due to rotation (Coriolis forces) in a coolant channel a secondary flow is superimposed to the basic flow. This leads to a change in the local heat transfer over the surface of the coolant channel as well as a change in the overall value of the heat transfer. Also the pressure loss over the channel length will change by rotation. By means of computational fluid dynamics (Finite Element Method) it was achieved to figure out the interaction between changing fluid flow and heat transfer. To validate the results obtained by a numerical flow simulation, a new measurement technique was developed. A laser-two-focus velocimeter has been combined with a rotation prism which allows continued measurements in a rotating scaled up channel. (orig.) [German] Bedingt durch die Rotationsbewegung eines Kuehlkanals wird die Grundstroemung von einem Sekundaerwirbel ueberlagert (Corioliskraefte). Durch diese Einfluesse aendert sich sowohl der lokale Waermeuebergang ueber der Kanaloberflaeche als auch die globalen Waermeuebertragungsraten ueber dem gesamten Kanal. Ebenfalls aendert sich durch die Rotation der Druckverlust ueber der Kanallaenge. Durch eine numerische Stroemungssimulation (Finite-Element-Methode) war es moeglich, einen detaillierten Zusammenhang zwischen dem veraenderten Stroemungsverhalten und dem Waermeuebertragungsverhalten darzustellen. Um die numerisch gewonnenen Ergebnisse experimentell abzusichern, wurde eine neuartige Messtechnik entwickelt. Ein Laser-2-Fokus-Velozimeter wurde mit einem Bilddrehprisma kombiniert, und dies erlaubte eine kontinuierliche Messung in einem rotierenden vergroesserten Modellkanal. (orig.)

Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

Directory of Open Access Journals (Sweden)

A. Rama Rao

2008-04-01

Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

Ultrasonic imaging in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Lubeigt, E. [CEA Cadarache, DEN/DTN/STCP/LIET, 13108 Saint-Paul-Lez-Durance Cedex (France); Laboratoire de Mecanique et d' Acoustique, CNRS UPR 7051, 13402 Marseille Cedex 20 (France); Mensah, S.; Chaix, J.F.; Rakotonarivo, S. [Laboratoire de Mecanique et d' Acoustique, CNRS UPR 7051, 13402 Marseille Cedex 20 (France); Gobillot, G. [CEA Cadarache, DEN/DTN/STCP/LIET, 13108 Saint-Paul-Lez-Durance Cedex (France)

2015-07-01

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the

Ultrasonic imaging in liquid sodium

International Nuclear Information System (INIS)

Lubeigt, E.; Mensah, S.; Chaix, J.F.; Rakotonarivo, S.; Gobillot, G.

2015-01-01

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the

Development of a transfer model for design of sodium purification systems for Fast Breeder Reactors

International Nuclear Information System (INIS)

Khatcheressian, N.

2013-01-01

Operating a Sodium Fast Reactor (SFR) in reliable and safe conditions requires to master the quality of the sodium fluid coolant, regarding oxygen and hydrogen impurities contents. A cold trap is a purification unit in SFR, designed for maintaining oxygen and hydrogen contents within acceptable limits. The purification of these impurities is based on crystallization of sodium hydride on cold walls and sodium oxide or hydride on wire mesh packing. Indeed, as oxygen and hydrogen solubilities are nearly nil at temperatures close to the sodium fusion point, i.e. 97.8 C, on line sodium purification can be performed by crystallization of sodium oxide and hydride from liquid sodium flows. However, the management of cold trap performances is necessary to prevent from unforeseen maintenance operations, which could induce shut-down of the reactor. It is thus essential to understand how a cold trap fills up with impurities crystallization in order to optimize the design of this system and to overcome any problems during nominal operation. The objective is to develop a design and simulation tool for cold traps able to predict the location and the amount of the impurities deposited. Crystallization model involve phenomena coupling in a porous medium with hydrodynamics, heat and mass transfer, distinguishing nucleation and growth phases for each impurity. It enables to understand how thermo hydraulic conditions and growing impurities interact on each other. This analysis will adapt operating and management conditions in order to optimize purification requirements. (author) [fr

Containment Sodium Chemistry Models in MELCOR.

Energy Technology Data Exchange (ETDEWEB)

Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-04-01

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRC code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.

EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

International Nuclear Information System (INIS)

Saito, Shinzo

1977-09-01

In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

International Nuclear Information System (INIS)

Whaley, R.L.; Sanders, J.P.

1976-09-01

A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

Mechanical energy yields and pressure volume and pressure time curves for whole core fuel-coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Coddington, P [United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1979-10-15

In determining the damage consequences of a whole core Fuel-Coolant Interaction (FCI), one measure of the strength of a FCI that can be used and is independent of the system geometry is the constant volume mixing mechanical yield (often referred to as the Hicks-Menzies yield), which represents a near upper limit to the mechanical work of a FCI. This paper presents a recalculation of the Hicks-Menzies yields for UO{sub 2} and sodium for a range of initial fuel temperatures and fuel to coolant mass ratios, using recently published UO{sub 2} and sodium equation of state data. The work presented here takes a small number of postulated FCIs with as wide range as possible of thermal interaction parameters and determines their pressure-volume P(V) and pressure-time P(t) relations, using geometrical constraints representative of the reactor. Then by examining these P(V) and P(t) curves a representative pressure-relative volume curve or range of possible curves, for use in containment analysis, is recommended

Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

International Nuclear Information System (INIS)

Melnikov, V.; Nigmatulin, B.

1997-01-01

The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

Generation IV SFR Nuclear Reactors: Under Sodium Robotics for ASTRID

International Nuclear Information System (INIS)

Jouan-de-Kervenoael, T.; Rey, F.; Baque, F.

2013-06-01

For non-removable components of the future ASTRID prototype, repair operations will be performed in a gas environment. If the faulty area is located under the sodium free level, the gas-tight system will have to contain the inspection and repair tools and to protect them from the surrounding liquid sodium. Concerning repair tools, the unique laser tool has been selected for future SFRs: the repair scenario for in-sodium structures will first involve removing the sodium (after bulk draining), then machining and finally welding. Concerning conventional tools (brush or gas blower for sodium removal, milling machine for machining and TIG for welding for which its feasibility was demonstrated in the 1990's) are still considered as a back-up solution. The maintenance of future ASTRID nuclear reactor prototype (inspection, repair) will be performed during shut down periods with some robotic carriers which have to be introduced within the main vessel, in primary 200 deg. C sodium coolant with argon gas cover. Inspection campaigns will be 20 days long. These robots (or carriers) will allow bringing inspection and repairing tools up to concerned components and structures. The needed degrees of freedom associated to these operations will be assumed either directly by the carrier itself or by specifics lower end carrier device for accurate local positioning. Several carriers will be designed, well adapted to specific needs: type of maintenance operation and location of inspection and repair sites. Feedback experience was gained during Superphenix SFR operation with the MIR robot which allowed to successfully make the Non Destructive Examination of main vessel welding joints, the carrier being outside bulk sodium. Operating conditions for the ASTRID robots will be harder from those of the MIR robot: temperature ranging from 180 deg. C to 200 deg. C, radiation dose ranging from 105 to 106 Gy, but mainly direct immersion within liquid sodium coolant. At the design phase of

Multidimensional analysis of fluid flow in the loft cold leg blowdown pipe during a loss-of-coolant experiment

International Nuclear Information System (INIS)

Demmie, P.N.; Hofmann, K.R.

1979-03-01

A computer analysis of fluid flow in the Loss-of-Fluid Test (LOFT) cold leg blowdown pipe during a loss-of-coolant experiment (LOCE) was performed using the computer program K-FIX/MOD1. The purpose of this analysis was to evaluate the capability of K-FIX/MOD1 to calculate theoretical fluid quantity distributions in the blowdown pipe during a LOCE for possible application to the analysis of LOFT experimental data, the determination of mass flow, or the development of data reduction models. A rectangular section of a portion of the LOFT blowdown pipe containing measurement Station BL-1 was modeled using time-dependent boundary conditions. Fluid quantities were calculated during a simulation of the first 26 s of LOFT LOCE L1-4. Sensitivity studies were made to determine changes in void fractions and velocities resulting from specific changes in the inflow boundary conditions used for this simulation

Fragmentation of molten copper drop caused by entrapment of liquid sodium

International Nuclear Information System (INIS)

Abe, N.; Sugiyama, K.; Nishimura, S.; Kinoshita, I.

2001-01-01

In core meltdown accidents, it is possible to occur thermal interactions between molten fuel and coolant. Analysis of the steam explosion, which is one of the most severe phenomena in such thermal interactions, is important for the safety evaluation. The steam explosion is a phenomenon that intensive pressure waves are caused by the explosive thermal interaction between high and low temperature liquids, and is considered to be one of the phenomena that can cause a serious failure of the nuclear reactor structures. In a large-scale steam explosion, the fragmentation of hot molten material causes a rapid increase of heat transfer area, and it is achieved to transmit instantaneously a large amount of heat to coolant. Two ideas are chiefly considered as the mechanism of the fragmentation. The one is the hypothesis that hydrodynamic effect causes fragmentation of hot liquid. According to this hypothesis, the high temperature drops flake off from the surface. The other is that fragmentation is caused by the interface instability accompanied by collapse of the steam bubble formed around a hot liquid. In this research, the possibility of the internal fragmentation caused by the coolant jet is focused in. Experiments were conducted on the condition that the surface of melt drops solidify at the moment drops contact the coolant. The possibility of the fragmentation of hot liquid from its surface was eliminated in this condition. To satisfy this condition, molten copper was chosen as hot liquid, and liquid sodium was used as coolant to verify the effect of the driving force of the sodium jet. (author)

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Mahoney, J.J.

1980-10-01

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

Nuclear fuel assembly

International Nuclear Information System (INIS)

Wakamatsu, Mitsuo.

1974-01-01

Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

International Nuclear Information System (INIS)

Monta, Kazuo.

1974-01-01

Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

International Nuclear Information System (INIS)

Andreychek, T.S.

1987-01-01

To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

Local velocity measurements in lead-bismuth and sodium flows using the ultrasound doppler velocimetry

International Nuclear Information System (INIS)

Eckert, S.; Gerbeth, G.

2003-01-01

We will present measurements of the velocity profiles in liquid sodium and eutectic lead-bismuth by means of the Ultrasonic Doppler Velocimetry (UDV). A sodium flow in a rectangular duct exposed to an external, transverse magnetic field has been examined. To demonstrate the capability of UDV the transformation of the well-known turbulent, piston-like profile to an M-shaped velocity profile for growing magnetic field strength was observed. The significance of artifacts such as caused by the existence of reflecting interfaces in the measuring domain will be discussed. In the sodium case, the measurements were performed through the channel wall. An integrated ultrasonic sensor with acoustic wave-guide has been developed to overcome the limitation of ultrasonic transducers to temperatures lower than 200 .deg. C. This sensor can presently be applied at maximum temperatures up to 800 .deg. C. Stable and robust measurements have been performed in various PbBi flows in our laboratory at FZR as well as at the THESYS loop of the KALLA laboratory of the ForschungsZentrum Karlsruhe (FZK). We will also present experimental results obtained in a PbBi bubbly flow at 250...300 .deg. C. Argon bubbles were injected through a single orifice in a cylindrical container filled with stagnant PbBi. Velocity profiles were measured in the bubble plume. Mean values of the liquid as well as the bubble velocity were extracted from the data and will be presented as function of the gas flow rate

Research on coolant radiochemistry

International Nuclear Information System (INIS)

Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

2004-04-01

The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

Calculation and analysis of neutron and radiation characteristics of lead coolants with isotopic tailoring for future nuclear power facilities

Energy Technology Data Exchange (ETDEWEB)

Blokhin, A.I.; Ivanov, A.P.; Korobeinikov, V.V.; Lunev, V.P.; Manokhin, V.N.; Khorasanov, G.L. [SSC RF A. I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation)

2000-03-01

A new type of safe fast reactor with lead coolant was proposed in Russia. The use of coolants with low moderating properties is one of the ways to get a hard neutron spectrum and an increase in the burning of Np-237, Am-243 and other miner actinides(MA) fissionable preferentially in the fast reactor. The stable lead isotope, Pb-208, is proposed as the one of such coolants. The neutron inelastic scattering cross-section of Pb-208 is 3.0-3.5 times less than the one of other lead isotopes. Calculation of the MA transmutation rates in the standard BN-type fast reactor with different coolants is performed by Monte-Carlo method using Code MMKFK. Six various models are simulated for the fast reactor blanket with different kinds of fuel and coolant. The fast reactor with natural-lead coolant practically does not differ from the reactor with sodium coolant relative to MA incineration. The use of Pb-208 as a coolant in the fast reactor results in increasing incineration of MA from 18 to 26% in comparison with a usual fast reactor. Calculation of induced radioactivity was performed using the FISPACT-3 inventory code, also. The results include total induced radioactivity and dose rate for initial material composition and selected long-lived radionuclides. The calculations show that the coolant consisting of lead isotope, Pb-206, or Pb-207, can be considered as the low-activation one because it does not practically contain long-lived toxic radionuclides. (M. Suetake)

Delayed gamma power measurement for sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

2011-01-15

Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

International Nuclear Information System (INIS)

Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

1981-02-01

This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

Determination of blade-to-coolant heat-transfer coefficients on a forced-convection, water-cooled, single-stage turbine

Science.gov (United States)

Freche, John C; Schum, Eugene F

1951-01-01

Blade-to-coolant convective heat-transfer coefficients were obtained on a forced-convection water-cooled single-stage turbine over a large laminar flow range and over a portion of the transition range between laminar and turbulent flow. The convective coefficients were correlated by the general relation for forced-convection heat transfer with laminar flow. Natural-convection heat transfer was negligible for this turbine over the Grashof number range investigated. Comparison of turbine data with stationary tube data for the laminar flow of heated liquids showed good agreement. Calculated average midspan blade temperatures using theoretical gas-to-blade coefficients and blade-to-coolant coefficients from stationary-tube data resulted in close agreement with experimental data.

Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

International Nuclear Information System (INIS)

Sun Yanfei; Zhong Shuoping; Huang Xiaojin

2014-01-01

Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

Coolant system decontamination

International Nuclear Information System (INIS)

Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

1981-01-01

An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

2012-01-01

The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net-power...

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Neutron Physics aspects of using lead as a coolant in Fast Reactors

International Nuclear Information System (INIS)

Kiefhaber, E.

1991-02-01

The use of lead as a coolant for fast reactors is being considered as an attractive alternative in the USSR, especially with respect to its inherent safety features. In order to come to an own assessment at KfK, some investigations have been performed concerning a comparison of the nuclear characteristics of fast reactors with lead and sodium cooling. The studies have shown, that the nuclear and thermal hydraulic design calculations do not face special problems and that the nuclear characteristics of both types of cores do not differ essentially, except for the coolant density or void effect, which is more favourable for smaller sized lead cooled cores. A proper safety assessment of lead cooled cores will however require more detailed safety studies. Crucial points of lead cooling are the strong corrosion of austenitic steels in lead and the unknown behavior of ferritic steels in lead and under irradiation

Development of a Program for Predicting Flow Instability in a Once-through Sodium-Heated Steam Generator (III)

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Yoon, Jung; Kim, Jong Bum; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Two-phase flow systems can be subjected to several types of instability problems. Density-wave oscillation is the most common and important type of instability in boiling channels. Such instability gives difficulties in predictions of system performance and system control, and component failure due to thermal fatigue. A computer program developed for predicting two-phase flow instability in a steam generator heated by liquid sodium was presented in the previous works. Limit cycle was predicted even in a fixed node system. The amplitude of inlet flow rate is larger than that of outlet flow rate. The amplitude of phase change location oscillation within boiling-to-vapor boundary node is larger than that of liquid-to-boiling boundary node. Sodium and steam temperature are invariant at tube exit.

Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

International Nuclear Information System (INIS)

Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

1976-01-01

Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

Upper plenum dump during reflood in PWR loss-of-coolant accident

International Nuclear Information System (INIS)

Sudo, Yukio; Griffith, Peter.

1981-01-01

Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

Directory of Open Access Journals (Sweden)

Zhi Li

2017-09-01

Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

Safety tests carried out at Cadarache. Sodium fires

International Nuclear Information System (INIS)

Fruchard, M.

1976-01-01

Safety test on sodium fires developed at the Cadarache Nuclear Centre by the Department of Nuclear Safety, section of safety experiments on radioactivity transfer are conducted in two main directions: analysis of the behavior and thermodynamic consequences of accidental fires, working on the basis of typical experimental results; research and development of methods and equipment to control and if possible extinguish these fires. The most important part of this programme is concerned with the sodium pool fires which would result from the failure of a secondary coolant circuit pipe [fr

Sodium vapor deposition onto a horizontal flat plate above liquid sodium surface, 2

International Nuclear Information System (INIS)

Kudo, Kazuhiko; Hirata, Masaru.

1977-01-01

The sodium vapor deposition onto a horizontal flat plate above liquid sodium surface was studied. The analysis was performed by assuming that the sodium mist is emitted into the main flow without condensation and then grows up in the main flow and drops on the sodium surface. The effects of growth of sodium mist to the system were investigated. The model of the phenomena is the sodium deposition onto a horizontal flat plate which is placed above the sodium surface with the medium cover gas. One-dimensional analysis can be done. The rate of deposition is greatly reduced when the temperature of the flat plate is lowered. For the analysis of this phenomena, it is assumed that the sodium mist grows by condensation. One of results is that the real state may be the state between the state that the condensation of mist is made in the boundary layer and the state that the mist is condensed in the main flow. Others are that there is no effect of sodium mist condensation on the rate of deposition, and that the rate of the vaporization of sodium is given by the original and the modified model. (Kato, T.)

Flow-throttling orifice nozzle

International Nuclear Information System (INIS)

Sletten, H.L.

1975-01-01

A series-parallel-flow type throttling apparatus to restrict coolant flow to certain fuel assemblies of a nuclear reactor is comprised of an axial extension nozzle of the fuel assembly. The nozzle has a series of concentric tubes with parallel-flow orifice holes in each tube. Flow passes from a high pressure plenum chamber outside the nozzle through the holes in each tube in series to the inside of the innermost tube where the coolant, having dissipated most of its pressure, flows axially to the fuel element. (U.S.)

The role of two-phase coolant in moderating fretting in nuclear steam generators

International Nuclear Information System (INIS)

Dyke, J.M.

2004-01-01

This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. The current paper addresses tube bundle and anti-vibration bars (AVB) geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. In a T(sat) fluid, the tube surface is uniformly coating with growing bubbles whose momentum is perpendicular to the surface at first, then they are swept away by the bulk flow. The combination of this momentum, the phase change and the water film remaining on the surface, counteract the vibration energy of the tube-AVB system, reducing the likelihood of metal-to-metal contact and consequent fretting. To maximize the benefit of the cushioning effect, the following design inputs are needed: 1) the AVB-tube interface should have sufficient clearance for the T(sat) solution to operate, 2) The AVB should be wide enough to generate the necessary cushioning force, and 3) the AVB should be thin enough to be flexible and absorb some of the transferred vibration energy. Furthermore, fretting and crude deposition at the AVB-tube interface can be reduced or eliminated by reducing the number of AVBs, increasing clearances and making the AVBs limber

Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

International Nuclear Information System (INIS)

Soli T. Khericha

2006-01-01

This report presents preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T and FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation

Large eddy simulation of mixing between hot and cold sodium flows - comparison with experiments

Energy Technology Data Exchange (ETDEWEB)

Simoneau, J.P.; Noe, H.; Menant, B.

1995-09-01

The large eddy simulation is becoming a potential powerful tool for the calculation of turbulent flows. In nuclear liquid metal cooled fast reactors, the knowledge of the turbulence characteristics is of great interest for the prediction and the analysis of thermal stripping phenomena. The objective of this paper is to give a contribution in the evaluation of the large eddy simulation technique is an individual case. The problem chosen is the case of the mixing between hot and cold sodium flows. The computations are compared with available sodium tests. This study shows acceptable qualitative results but the simple model used is not able to predict the turbulence characteristics. More complex models including larger domains around the fluctuating zone and fluctuating boundary conditions could be necessary. Validation works are continuing.

Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

Directory of Open Access Journals (Sweden)

S. K. Pavlov

2014-01-01

Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

Coolant clean-up and recycle systems

International Nuclear Information System (INIS)

Ito, Takao.

1979-01-01

Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

Numerical simulation of flow field in the China advanced research reactor flow-guide tank

International Nuclear Information System (INIS)

Xu Changjiang

2002-01-01

The flow-guide tank in China advanced research reactor (CARR) acts as a reactor inlet coolant distributor and play an important role in reducing the flow-induced vibration of the internal components of the reactor core. Numerical simulations of the flow field in the flow-guide tank under different conceptual designing configurations are carried out using the PHOENICS3.2. It is seen that the inlet coolant is well distributed circumferentially into the flow-guide tank with the inlet buffer plate and the flow distributor barrel. The maximum cross-flow velocity within the flow-guide tank is reduced significantly, and the reduction of flow-induced vibration of reactor internals is expected

Atmospheric dispersion of sodium aerosol due to a sodium leak in a fast breeder reactor complex

International Nuclear Information System (INIS)

Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

2008-01-01

Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model dose not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions. (author)

Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

International Nuclear Information System (INIS)

Dibyo, Sukmanto

2000-01-01

Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

Optimization of magnetic amplification by flow constraints in turbulent liquid sodium

International Nuclear Information System (INIS)

Nornberg, M. D.; Taylor, N. Z.; Forest, C. B.; Rahbarnia, K.; Kaplan, E.

2014-01-01

Direct measurements of the vector turbulent emf in a driven two-vortex flow of liquid sodium were performed in the Madison Dynamo Experiment [K. Rahbarnia et al., Astrophys. J. 759, 80 (2012)]. The measured turbulent emf is anti-parallel with the mean current and is almost entirely described by an enhanced resistivity, which increases the threshold for a kinematic dynamo. We have demonstrated that this enhanced resistivity can be mitigated by eliminating the largest-scale eddies through the introduction of baffles. By tailoring the flow to reduce large-scale components and control the helical pitch, we have reduced the power required to drive the impellers, doubled the magnetic flux generated by differential rotation, and increased the decay time of externally applied magnetic fields. Despite these improvements, the flows remain sub-critical to the dynamo instability due to the reemergence of turbulent fluctuations at high flow speeds

Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

2006-01-01

The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

Corrosion behaviors of ceramics against liquid sodium. Sodium corrosion characteristics of sintering additives

International Nuclear Information System (INIS)

Tachi, Yoshiaki; Kano, Shigeki; Hirakawa, Yasushi; Yoshida, Eiichi

1998-01-01

It has been progressed as the Frontier Materials Research to research and develop ceramics to apply for several components of fast breeder reactor using liquid sodium as coolant instead of metallic materials. Grain boundary of ceramics has peculiar properties compared with matrix because most of ceramics are produced by hardening and firing their raw powders. Some previous researchers indicated that ceramics were mainly corroded at grain boundaries by liquid sodium, and ceramics could not be used under corrosive environment. Thus, it is the most important for the usage of ceramics in liquid sodium to improve corrosion resistance of grain boundaries. In order to develop the advanced ceramics having good sodium corrosion resistance among fine ceramics, which have recently been progressed in quality and characteristics remarkably, sodium corrosion behaviors of typical sintering additives such as MgO, Y 2 O 3 and AlN etc. have been examined and evaluated. As a result, the followings have been clarified and some useful knowledge about developing advanced ceramics having good corrosion resistance against liquid sodium has been obtained. (1) Sodium corrosion behavior of MgO depended on Si content. Samples containing large amount of Si were corroded severely by liquid sodium, whereas others with low Si contents showed good corrosion resistance. (2) Both Y 2 O 3 and AlN, which contained little Si, showed good sodium corrosion resistance. (3) MgO, Y 2 O 3 and AlN are thought to be corroded by liquid sodium, if they contain some SiO 2 . Therefore, in order to improve sodium corrosion resistance, it is very important for these ceramics to prevent the contamination of matrix with SiO 2 through purity control of their raw powders. (author)

Device of capturing for radioactive corrosion products

International Nuclear Information System (INIS)

Ohara, Atsushi; Fukushima, Kimichika.

1984-01-01

Purpose: To increase the area of contact between the capturing materials for the radioactive corrosion products contained in the coolants and the coolants by producing stirred turbulent flows in the coolant flow channel of LMFBR type reactors. Constitution: Constituent materials for the nuclear fuel elements or the reactor core structures are activated under the neutron irradiation, corroded and transferred into the coolants. While capturing devices made of pure metal nickel are used for the elimination of the corrosion products, since the coolants form laminar flows due to the viscosity thereof near the surface of the capturing materials, the probability that the corrosion products in the coolants flowing through the middle portion of the channel contact the capturing materials is reduced. In this invention, rotating rolls and flow channels in which the balls are rotated are disposed at the upstream of the capturing device to forcively disturb the flow of the liquid sodium, whereby the radioactive corrosion products can effectively be captured. (Kamimura, M.)

Chimera grids in the simulation of three-dimensional flowfields in turbine-blade-coolant passages

Science.gov (United States)

Stephens, M. A.; Rimlinger, M. J.; Shih, T. I.-P.; Civinskas, K. C.

1993-01-01

When computing flows inside geometrically complex turbine-blade coolant passages, the structure of the grid system used can affect significantly the overall time and cost required to obtain solutions. This paper addresses this issue while evaluating and developing computational tools for the design and analysis of coolant-passages, and is divided into two parts. In the first part, the various types of structured and unstructured grids are compared in relation to their ability to provide solutions in a timely and cost-effective manner. This comparison shows that the overlapping structured grids, known as Chimera grids, can rival and in some instances exceed the cost-effectiveness of unstructured grids in terms of both the man hours needed to generate grids and the amount of computer memory and CPU time needed to obtain solutions. In the second part, a computational tool utilizing Chimera grids was used to compute the flow and heat transfer in two different turbine-blade coolant passages that contain baffles and numerous pin fins. These computations showed the versatility and flexibility offered by Chimera grids.

Coolant leakage detecting device

International Nuclear Information System (INIS)

Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

1995-01-01

The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

Release of fission products from contaminated sodium fires

International Nuclear Information System (INIS)

Jordan, S.

1976-01-01

Leaks in the primary coolant system of a LMFBR and also serious incidents with tank rupture may entail the escape of fission products into the containment of the reactor. For incident analysis it is important to know the retention capability of sodium for the different fission products. The release of cesium and strontium from pools contaminated with 100 to 1000 ppM was investigated by experiments. The cesium content of airborne aerosols depends on oxygen concentration: at 21 percent oxygen concentration the Cs content of sodium-oxide aerosols is 3 times and at 0.5 percent 15 times as high as the initial Cs concentration in the pool. Strontium content of aerosols over burning contaminated sodium pools is 10 3 times smaller than the strontium pool concentration

Process and system for stirring liquid sodium flowing through the primary circuit of a steam generator

International Nuclear Information System (INIS)

Fabregue, J.P.

1982-01-01

The invention concerns the stirring of the liquid sodium of a steam generator comprising a primary circuit composed of an elongated vessel through which the liquid sodium flows, a secondary circuit composed of a number of tubes extending inside the long cyclindrical vessel. The process consists in imparting simultaneously to the liquid sodium, during its passage through the cylindrical vessel, a movement of continuous rotation about the longitudinal axis of the cylindrical vessel and an alternating series of radial movements, centripetal and centrifugal, in relation to the longitudinal axis, so that each unit quantity of the sodium comes into contact with a large number of tubes. The application particularly concerns steam generators for nuclear power stations [fr

Coolant inlet device for nuclear reactors

International Nuclear Information System (INIS)

Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

1969-01-01

Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

Cleaning of Sodium in the Cold Trap

International Nuclear Information System (INIS)

Kim, Byung Ho; Jeong, Ji Young; Kim, Jong Man; Choi, Byung Hae; Nam, Ho Yun

2005-01-01

The purpose of a cleaning process is to remove the residual sodium adhering to the component walls once it has been properly drained. It is necessary to clean and decontaminate a component, especially the large components of the primary coolant system; such as the intermediate heat exchangers and the primary pump. Improper and inadequate cleaning has in a number of cases resulted in problems in the storage, handling, and reuse of components. Several types of failures due to improper cleaning procedures have been defined in the past. Inadequate and incomplete removal of sodium results in residues which may contain metallic sodium and alkaline compounds such as sodium hydroxide, sodium oxide, sodium carbonate, and various types of alcoholates. Reinsertion of components containing these compounds into a high-temperature sodium system can result in either the intergranular penetration characteristic of a high- oxygen sodium or an accelerated corrosion due to oxygen. The methods used for cleaning sodium equipment depend on the condition and types of equipment to be cleaned and whether the equipment is to be reused. Cleaning methods are needed that will avoid a deleterious local overheating, material surface degradation or deposits, chemical, physical, or mechanical damage, and external effects. This paper discusses a steam-nitrogen gas cleaning method for the routine applications that permits the reuse of the cold trap in sodium

THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry

International Nuclear Information System (INIS)

Camous, F.

1983-01-01

The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way

Thermophysical properties of sodium nitrate and sodium chloride solutions and their effects on fluid flow in unsaturated media

International Nuclear Information System (INIS)

Xu, Tianfu; Pruess, Karsten

2001-01-01

Understanding movement of saline sodium nitrate (NaNO 3 ) waste solutions is important for assessing the contaminant migration near leaking waste storage tanks in the unsaturated zone at the Hanford site (Washington, USA). The purpose of this study is to contribute a basic understanding of effects of the thermophysical behavior of NaNO 3 solutions on fluid flow in unsaturated media. We first present mathematical expressions for the dependence of density, viscosity, solubility and vapor pressure of NaNO 3 solutions on both salt concentration and temperature, which were determined by fitting from published measured data. Because the previous studies of thermophysical behavior of sodium chloride (NaCl) solutions can provide a basis for those of NaNO 3 solutions, we also present a comparison of thermophysical properties of both salt solutions. We have implemented the functional thermophysical properties of NaNO 3 solutions into a new TOUGH2 equation-of-state module EWASG-NaNO 3 , which is modified from a previous TOUGH2 equation-of-state module EWASG for NaCl. Using the simulation tool, we have investigated effects of the thermophysical properties on fluid flow in unsaturated media. The effect of density and viscosity of saline solutions has been long recognized. Here we focus our attention on the effect of vapor pressure lowering due to salinity. We present simulations of a one-dimensional problem to study this salinity-driven fluid flow. A number of simulations were performed using different values of thermal conductivity, permeability, and temperature, to illustrate conditions and parameters controlling these processes. Results indicate that heat conduction plays a very important role in this salinity-driven vapor diffusion by maintaining a nearly constant temperature. The smaller the permeability, the more water is transferred into the saline environment. Effects of permeability on water flow are also complicated by effects of capillary pressure and tortuosity. The

Organic coolant for ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

1991-04-01

ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

Primary coolant feed and bleed operating regions for the Midland Plant

International Nuclear Information System (INIS)

Tsai, M.S.

1985-01-01

Operating regions for primary coolant feed and bleed cooling are developed for the Midland Plant using core decay heat, the high-pressure injection (HPI) system capacity, and flow rate relief through the power-operated relief valve (PORV). This mode of cooling is used for accident scenarios in which the normal core cooling means of a nuclear power plant is lost because of loss of water inventory in the steam generators. The HPI flow is based on the capacities of one and two pumps. Saturated steam, saturated water, and subcooled water are considered to be possible states of the fluid being relieved through the PORV. In estimating the PORV relief rate, flow equations are derived from the Electric Power Research Institute test data obtained from the same model and size valve that is used in the Midland Plant. For easy reference by operators, the operating region is displayed on a plane of reactor coolant system pressure and temperature. The technique developed for the Midland Plant provides a convenient method for examining the feed and bleed cooling capability for a nuclear power plant that employs a pressurized water reactor system