Experience with cleaning of sodium-wetted components and decontamination at Nuclear Research Centre Karlsruhe

Energy Technology Data Exchange (ETDEWEB)

Menzenhauer, P; Borgstedt, U; Stamm, H H; Dippel, Th; Kunze, S; Hentschel, D [Kernforschungszentrum Karlsruhe, Karlsruhe (Germany)

1978-08-01

Within the framework of the Fast Breeder Project various institutes of the KarIsruhe Nuclear Research Center operate sodium loops for corrosion studies and component tests under sodium. The operation of these loops has led to extensive experience in cleaning sodium-wetted components. This experience relates to the alcohol method, the removal of sodium by melting, storage in air, and to cleaning by means of steam. Deposition samples from radioactive sodium loops were used for decontamination experiments employing various decontaminating agents. The department concerned with the treatment of radioactive waste studied the use of molten salts and paste type cleansers for components unsuitable for mechanical decontamination, primarily with the objective to reduce the amount of radioactive waste. (author)

Experience with cleaning of sodium-wetted components and decontamination at Nuclear Research Centre Karlsruhe

International Nuclear Information System (INIS)

Menzenhauer, P.; Borgstedt, U.; Stamm, H.H.; Dippel, Th.; Kunze, S.; Hentschel, D.

1978-01-01

Within the framework of the Fast Breeder Project various institutes of the KarIsruhe Nuclear Research Center operate sodium loops for corrosion studies and component tests under sodium. The operation of these loops has led to extensive experience in cleaning sodium-wetted components. This experience relates to the alcohol method, the removal of sodium by melting, storage in air, and to cleaning by means of steam. Deposition samples from radioactive sodium loops were used for decontamination experiments employing various decontaminating agents. The department concerned with the treatment of radioactive waste studied the use of molten salts and paste type cleansers for components unsuitable for mechanical decontamination, primarily with the objective to reduce the amount of radioactive waste. (author)

Corrosion and cleaning aspects of sodium side crevices in components of LMFBR's

International Nuclear Information System (INIS)

Chirer, E.G.

1978-01-01

Although the presence of crevices is excluded in critical areas of sodium components by design, their occurrence in other areas cannot be eliminated completely. During the lifetime of a component high concentrations of sodium compounds, such as caustics, may be formed in crevices. These compounds can remain within the crevices for some time. In this respect the following situations are recognized: - Reaction products from initial contaminants such as oxide scales. The component with crevices containing oxide scale either from the manufacturing process or insufficient cleaning after water pressure testing is exposed to sodium during actual operation. - Reaction products formed during or after cleaning. Sodium in the crevices of a drained component reacts with water vapour or water during cleaning or during subsequent storing or handling under non-perfect conditions. Before refilling with sodium the component is heated to preheat temperature. Same situation as above, however the component is exposed to sodium at operating temperature. These cycles can be repeated several times. - Products from a small sodium-water reaction. Caustic products from a small sodium-water reaction may be present in crevices or dead ends of a component which is exposed to high temperature during sodium operation or during vacuum distillation. The aims of the investigations are: determination of the corrosive aspects of high concentration of caustic reaction products of sodium in crevices on the structural materials of the component; comparison of the effectiveness of different cleaning procedures in respect to removal of sodium from crevices, e.g. water, steam, alcohol cleaning, vacuum distillation. Concerning the first item, in particular the possibility of the occurrence of intercrystalline corrosion and stress corrosion cracking is investigated. Materials investigated are the Cr-Mo steels 2 1/4Cr1Mo stabilized with Nb, 9Cr1Mo, 12Cr1Mo and the austenitic stainless steal AISI 304. The

Sodium fire tests for investigating the sodium leak in Monju

International Nuclear Information System (INIS)

Seino, Hiroshi; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi

1996-01-01

As a part of the work for investigating the sodium leak accident which occurred in Monju on December 8, 1995, three tests, (1) sodium leak test, (2) sodium fire test-I, and (3) sodium fire test-II, were carried out at OEC/PNC. Main objectives of these tests are to confirm leak and burning behavior of sodium from the damaged thermometer, and effects of the sodium fire on integrity of the surrounding structure, etc. The main conclusions obtained from the tests are shown as below. 1) Average sodium leak rate obtained from the sodium leak test was about 50 g/sec. This was equivalent to the value estimated from level change in the sodium overflow tank in the Monju accident. 2) Observation from video cameras in the sodium fire tests revealed that in early stages of sodium leak, sodium dropped down out of the flexible tube of thermometer in drips. This dripping and burning were expanded in range as sodium splashed on the duct. 3) Though, in the sodium fire test-I, there was a decrease of about 1 mm at a thickness of the burning pan in the vicinity in just under in the leak point, there were completely no crack and failure. In the meantime, in the sodium fire test-II the six open holes were found in the floor liner. By this liner failure, the reaction between sodium and concrete might take place. At present, while the detailed evaluation on the sodium fire test-II has been mainly carried out, the investigation for clarifying the cause of the liner failure has been also carried out. (author)

Disassembly and removal of sodium instrumentation test loop

International Nuclear Information System (INIS)

Ishikawa, Okinobu; Onojima, Takamitu; Nagai, Keiichi

2000-07-01

In 1999, the Sodium Instrumentation Test Loop was disassembled and removed. This report describes the tasks and experiences obtained in removing sodium from a storage tank, disassembling, and cleansing components and related activities. Overall the disassembly, handling and cleansing tasks proceeded as planned and the activities were carried out efficiently and safely. Documentation of the process is meant to establish not only a procedure, but also a guideline for future similar tasks. (author)

Test Your Sodium Smarts

Science.gov (United States)

... You may be surprised to learn how much sodium is in many foods. Sodium, including sodium chloride ... foods with little or no salt. Test your sodium smarts by answering these 10 questions about which ...

UK fast reactor components. Sodium removal decontamination and requalification

International Nuclear Information System (INIS)

Donaldson, D.M.; Bray, J.A.; Newson, I.H.

1978-01-01

Extensive experience gained at the U.K.A.E.A. Dounreay Nuclear Power Development Establishment is being applied to form the basis of the plant to be provided for sodium removal, decontamination, and requalification of components in future commercial fast reactors. In the first part of a three part paper, the factors to be taken into account, showing the UK philosophy and approach to maintenance and repair operations are discussed. In the second part, PFR facilities for sodium removal and decontamination are described and some examples are given of cleaning components such as pumps, charge machine, cold trap baskets, and steam generator units. Similar facilities at DFR are briefly described. In the third part of the paper a short description is given of the Harwell mass transfer loop, currently used to study the deposition of activated stainless steel corrosion products. Decontamination method for pipework specimens cut from the loop are described and results of first screening tests of various chemical decontaminants are presented. (U.K.)

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)

Detection Test for Leakage of CO2 into Sodium Loop

International Nuclear Information System (INIS)

Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong

2015-01-01

This report is about the facility for the detection test for leakage of CO 2 into sodium loop. The facility for the detection test for leakage of CO 2 into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO 2 leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO 2 ) heat exchanger is one of the key components for the supercritical CO 2 Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO 2 heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO 2 exchanger, detection of CO 2 leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO 2 such as sodium carbonate (Na 2 CO 3 ) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na 2 CO 3 in sodium loop has not been developed yet. Therefore, detection of CO 2 and CO from reaction of sodium and CO 2 are proper to detect CO 2 leakage into sodium loop

Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

International Nuclear Information System (INIS)

Althaus; Jansing; Kesseler; Kirchner; Menck

1974-01-01

Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

Sodium-water reaction test facility (SWAT-3)

International Nuclear Information System (INIS)

Shimazu, Hisashi; Ukechi, Kazutoshi; Sasakura, Kazutake; Kusunoki, Junichi

1976-01-01

In the development of the liquid metal cooled fast breeder reactor (LMFBR), the steam generator (SG) is considered one of the most important components. The Power Reactor and Nuclear Fuel Development Corporation (PNC) is now promoting the research and development of the SG system used with the prototype fast breeder reactor ''Monju''. In this research, the phenomena of the sodium-water reaction in the SG are the key which must be investigated for the solution of problems. The test facility (SWAT-3) simulating Monju's SG on the scale of 1/2.5 was designed, fabricated and installed by IHI at Oarai Engineering Center of PNC, its pre-operation being accomplished in February 1975. The purpose of SWAT-3 is summarized as follows: (1) To perform an overall test on the safety of Monju's SG and intermediate heat transport system under the design condition against sodium-water reaction accidents. (2) To investigate the damage of the SG structure caused by the sodium-water reaction, and the possibility of repair and recovery operations. The first test was accomplished successfully on June 9, 1975. As a result of the test, the fundamental function of this test facility was proven to be satisfactory as expected. (auth.)

Testing of Local Velocity Transducer Used at Sodium Thermal Hydraulic Test Facilities

International Nuclear Information System (INIS)

Kim, Tae Joon; Eoh, Jae Hyuk; Hwang, In Koo; Jeong, Ji Young; Kim, Jong Man; Lee, Yong Bum; Kim, Yeong Il

2012-01-01

KAERI (Korea Atomic Energy Research Institute) will perform a test for a thermal hydraulic simulation with STELLA-1 for a Component Performance Test Sodium Loop in the year 2012, and subsequently it will construct for STELLA-2 for a Sodium Thermalhydraulic Experimental Facility in the year 2016. The STELLA-2 consists of a scaled reactor vessel with a core of electric heaters, four IHXs, two PHTS pumps, two DHXs, and two AHXs. In STELLA-2, several kinds of flow measurements exists. In this paper, the local velocity transducer as a prototype tested in IPPE (in Russia), was manufactured as a prototype by a shop in KAERI. This local velocity transducer will be used to measure the flow rate in a pool

Status of sodium removal and component decontamination technology in the SNR programme

Energy Technology Data Exchange (ETDEWEB)

Haubold, W [INTERATOM GmbH, Bergisch Gladbach (Germany); Smit, C Ch [MT-TNO Dept. 50-MW Sodium Component Test Facility, Hengelo (Netherlands); Stade, K Ch [Kernkraftwerk-Betriebsgesellschaft mbH, Eggenstein-Leopoldshafen (Germany)

1978-08-01

This paper summarizes the experience with sodium removal and component decontamination processes within the framework of the SNR project since the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' at Dounreay, April 9-12, 1973. The moist nitrogen process has been successfully applied to remove sodium from all 66 fuel elements of the KNK I core. Progress has been obtained in removing sodium from fuel elements and large components by vacuum distillation. Areas where future development is required are identified. (author)

Status of sodium removal and component decontamination technology in the SNR programme

International Nuclear Information System (INIS)

Haubold, W.; Smit, C.Ch.; Stade, K.Ch.

1978-01-01

This paper summarizes the experience with sodium removal and component decontamination processes within the framework of the SNR project since the IAEA Specialists Meeting on 'Decontamination of Plant Components from Sodium and Radioactivity' at Dounreay, April 9-12, 1973. The moist nitrogen process has been successfully applied to remove sodium from all 66 fuel elements of the KNK I core. Progress has been obtained in removing sodium from fuel elements and large components by vacuum distillation. Areas where future development is required are identified. (author)

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

Summary of HEDL sodium fire tests

International Nuclear Information System (INIS)

Hillard, R.K.

1978-10-01

The sodium fire test program and related studies at the Hanford Engineering Development Laboratory (HEDL) are described. The program is analytical and experimental in scope, with computer code development and experimental verification. Tests have ranged in size from gram quantity laboratory tests to 1600-kg sodium spills. The experimental work is performed in two facilities: the Large Sodium Fire Facility (LSFF) and the Containment Systems Test Facility (CSTF). Sodium fire extinguishment tests which verified the Fast Flux Test Facility (FFTF) secondary sodium fire protection system are described and related informaion on sodium burning rates and smoke release rates are correlated. The burning rates are compared to theoretical predictions based on heat and mass transfer analogy, with good agreement. Comparisons with the SOFIRE-II code are also made. Sodium combustion aerosol properties are defined as to chemical and physical nature, settling in closed vessels and effect of added water vapor. The HAA-3B aerosol behavior computer code is compared to tests in the 850-m 3 CSTF containment vessel. Sodium spray tests in the CSTF are compared with the SPRAY computer code. An air cleaning program is described, which has the objective of removing high mass concentration sodium combustion aerosols from vented cells and containment buildings. The aerosol mass holding capacity of commercial filters was measured and an aqueous scrubber system is described. The effects of sodium spills on cell structures were investigated, including water release from heated concrete, the reaction of sodium with concrete, the formation and spontaneous recombination of hydrogen, and the ability of steel cell liners to withstand large spills of high temperature sodium without leaking

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

UK fast reactor components - sodium removal decontamination and requalification

Energy Technology Data Exchange (ETDEWEB)

Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

1978-08-01

Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field.

UK fast reactor components - sodium removal decontamination and requalification

International Nuclear Information System (INIS)

Donaldson, D.M.; Bray, J.A.; Newson, I.H.

1978-01-01

Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

Summary of HEDL sodium fire tests

International Nuclear Information System (INIS)

Hilliard, R.K.

1979-01-01

The sodium fire test program and related studies at the Hanford Engineering Development Laboratory (HEDL), covering the period from 1972 to 1978, are described. The program is analytical and experimental in scope, with computer code development and experimental verification. Tests have ranged in size from gram quantity laboratory tests to 1600-kg sodium spills. The experimental work is performed in two facilities: the Large Sodium Fire Facility (LSFF) and the Containment Systems Test Facility (CSTF). The facilities are described and the experimental results summarized. Sodium fire extinguishment tests which verified the Fast Flux Test Facility (FFTF) secondary sodium fire protection system are described and related information on sodium burning rates and smoke release rates are correlated. The burning rates are compared to theoretical predictions based on heat and mass transfer analogy, with good agreement. Comparisons with the SOFIRE-II code are also made. Sodium combustion aerosol properties are defined as to chemical and physical nature, settling in closed vessels and effect of added water vapor. The HAA-38 aerosol behaviour computer code is compared to tests in the 850-m 3 CSTF containment vessel. Sodium spray tests in the CSTF are compared with the SPRAY computer code. An air cleaning program is described, which has the objective of removing high mass concentration sodium combustion aerosols from vented cells and containment buildings. The aerosol mass holding capacity of commercial filters was measured and an aqueous scrubber system is described. The effects of sodium spills on cell structures were investigated, including water release from heated concrete, the reaction of sodium with concrete, the formation and spontaneous recombination of hydrogen, and the ability of steel cell liners to withstand large spills of high temperature sodium without leaking. (author)

Summary of HEDL sodium fire tests

Energy Technology Data Exchange (ETDEWEB)

Hilliard, R K [Hanford Engineering Development Laboratory, Richland, WA (United States)

1979-03-01

The sodium fire test program and related studies at the Hanford Engineering Development Laboratory (HEDL), covering the period from 1972 to 1978, are described. The program is analytical and experimental in scope, with computer code development and experimental verification. Tests have ranged in size from gram quantity laboratory tests to 1600-kg sodium spills. The experimental work is performed in two facilities: the Large Sodium Fire Facility (LSFF) and the Containment Systems Test Facility (CSTF). The facilities are described and the experimental results summarized. Sodium fire extinguishment tests which verified the Fast Flux Test Facility (FFTF) secondary sodium fire protection system are described and related information on sodium burning rates and smoke release rates are correlated. The burning rates are compared to theoretical predictions based on heat and mass transfer analogy, with good agreement. Comparisons with the SOFIRE-II code are also made. Sodium combustion aerosol properties are defined as to chemical and physical nature, settling in closed vessels and effect of added water vapor. The HAA-38 aerosol behaviour computer code is compared to tests in the 850-m{sup 3} CSTF containment vessel. Sodium spray tests in the CSTF are compared with the SPRAY computer code. An air cleaning program is described, which has the objective of removing high mass concentration sodium combustion aerosols from vented cells and containment buildings. The aerosol mass holding capacity of commercial filters was measured and an aqueous scrubber system is described. The effects of sodium spills on cell structures were investigated, including water release from heated concrete, the reaction of sodium with concrete, the formation and spontaneous recombination of hydrogen, and the ability of steel cell liners to withstand large spills of high temperature sodium without leaking. (author)

Summary of results from sodium-heated steam generator test program

Energy Technology Data Exchange (ETDEWEB)

McDonald, J S

1975-07-01

A 28 MWt sodium-heated steam generator test unit developed and fabricated by Atomics International was operated in the Sodium Component Test Installation. The SCTI is located at the Liquid Metal Engineering Center which is operated for the Atomic Energy Commission by Atomics International, Reviewed in this paper are the results of the test operations and the findings of the post-test examination of the module. Testing was performed to assure the mechanical integrity of the unit over a wide range of simulated plant operating conditions and to develop a variety of performance data. Specific tests conducted included preheat, vibration, startup-shutdown, pressurization, steady state and parametric performance mapping, endurance, simulated leak injection, low- flow stability and simulated plant transients. (author)

Detection Test for Leakage of CO{sub 2} into Sodium Loop

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

This report is about the facility for the detection test for leakage of CO{sub 2} into sodium loop. The facility for the detection test for leakage of CO{sub 2} into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO{sub 2} leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO{sub 2}) heat exchanger is one of the key components for the supercritical CO{sub 2} Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO{sub 2} heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO{sub 2} exchanger, detection of CO{sub 2} leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO{sub 2} such as sodium carbonate (Na{sub 2}CO{sub 3}) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na{sub 2}CO{sub 3} in sodium loop has not been developed yet. Therefore, detection of CO{sub 2} and CO from reaction of sodium and CO{sub 2} are proper to detect CO{sub 2} leakage into sodium loop.

Investigation for the sodium leak Monju. Sodium fire test-II

International Nuclear Information System (INIS)

Uchiyama, Naoki; Takai, Toshihide; Nishimura, Masahiro; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi

2000-08-01

As a part of the work for investigating the sodium leak accident which occurred in the Monju reactor (hereinafter referred to as Monju), sodium fire test-II was carried out using the SOLFA-1 (Sodium Leak, Fire and Aerosol) facility at OEC/PNC. In the test, the piping, ventilation duct, grating and floor liner were all full-sized and arranged in a rectangular concrete cell in the same manner as in Monju. The main objectives of the test were to confirm the leak and burning behavior of sodium from the damaged thermometer, and the effects of the sodium fire on the integrity of the surrounding structure. The main conclusions obtained from the test are shown below: (1) Burning Behavior of Leaked Sodium : Images taken with a cameras in the test reveal that in the early stages of the sodium leak, the sodium dropped down out of the flexible tube in drips. (2) Damage to the Ventilation Duct and Grating : The temperature of the ventilation duct's inner surface fluctuated between approximately 600degC and 700degC. The temperature of the grating began rising at the outset of the test, then fluctuated between roughly 600degC and 900degC. The maximum temperature was about 1000degC. After the test, damage to the ventilation duct and the grating was found. Damage to the duct was greater than that at Monju. (3) Effects on the Floor Liner : The temperature of the floor liner under the leak point exceed 1,000degC at 3 hours and 20 minutes into the test. A post test inspection of the liner revealed five holes in an area about 1m x 1m square under the leak point. There was also a decrease of the liner thickness on the north and west side of the leak point. (4) Effects on Concrete : The post test inspection revealed no surface damage on either the concrete side walls or the ceiling. However, the floor concrete was eroded to a maximum depth 8 cm due to a sodium-concrete reaction. The compressive strength of the concrete was not degraded in spite of the thermal effect. (5) Chemical

Experience on sodium removal from FBTR components in its operating phase

International Nuclear Information System (INIS)

Jambunathan, D.; Rao, M.S.; Krishnamachari, V.S.; Kasiviswanathan, K.V.; Rajan, M.

1997-01-01

FBTR is a 40 MWt/13 MWe loop type, sodium cooled mixed carbide fuelled reactor. There are two primary loops, two secondary loops and a common steam-water circuit. Criticality was achieved in 1985 and during the course of the 10 years of operation phase experience has been gained on the decontamination of certain core components, primary sodium pumps, CRDM parts, handling components and cold traps. This paper deals with the decontamination aspects of these components in detail. For core subassemblies a remote sodium cleaning system was provided in the hot cell. The unique feature of this system is that it pumps alcohol into wash chamber using compressed nitrogen. During cleaning, the sub assembly is loaded into the system using master slave manipulator and compressed nitrogen gas is used to pump alcohol into the system for cleaning the outer and inner surfaces of the sub assembly in cycles. Each cycle takes about 15 minutes and around 5 to 6 cycles of cleaning were employed to remove 100 g of sodium. The level of alcohol in the tank is measured by ultrasonic level probes. The used alcohol is pumped to medium active waste storage tank. Primary and secondary sodium pumps are the large components which were cleaned off sodium using steam and water in decontamination pit. Lower part of CRDM with a failed bellow was another component cleaned in decontamination pit. An electro decontamination technique was successfully developed to remove coloration on the lower part of CRDM for reuse. A stainless steel carrier with antimony capsule was the first radio active component to undergo sodium cleaning operation in decontamination pit meant for large primary sodium circuit components after making necessary modifications. Decontamination of other components such as fingers of grippers and scrapper rings of charging and discharging flasks was carried out with alcohol under inert atmosphere. The secondary loop cold trap was successfully cleaned by hydride decomposition and vacuum

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

Experience with the instrumentation tests in large sodium test facilities

International Nuclear Information System (INIS)

Lauhoff, Th.; Ruppert, E.; Stehle, H.; Vinzens, K.

1976-01-01

A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

Failure rate evaluation for different components operating in sodium, based on operating experience of the RAPSODIE and the PHENIX reactors and the test loops

International Nuclear Information System (INIS)

Boisseau, J.; Dorey, J.; Hedin, F.; Le Floch, C.

1982-01-01

The failure rates of the following components, valves operating in sodium, mechanical and electromagnetic pumps, and heat exchangers including intermediate heat exchangers, cold traps, steam generators, are evaluated by analysing the main incidents which occurred on these components. Therefore, this paper contains an evaluation of the operating experience of components working in sodium and of the reliability of these components

Sodium component reliability data collection at CREDO

International Nuclear Information System (INIS)

Bott, T.F.; Haas, P.M.; Manning, J.J.

1979-01-01

The Centralized Reliability Data Organization (CREDO) has been established at Oak Ridge National Laboratory (ORNL) by the Department of Energy to provide a national center for collection, evaluation and dissemination of reliability data for advanced reactors. While the system is being developed and continuous data collection at the two U.S. reactor sites (EBR-II and FFTF) is being established, data on advanced reactor components which have been in use at U.S. test loops and experimental reactors have been collected and analyzed. Engineering, operating and event data on sodium valves, pumps, flow meters, rupture discs, heat exchangers and cold traps have been collected from more than a dozen sites. The results of analyses of the data performed to date are presented

Test and evaluation of Alco/BLH prototype sodium-heated steam generator. Final report

Energy Technology Data Exchange (ETDEWEB)

Kaplan, C.J.; Auge, L.J.; Cho, S.M.; Hanna, R.W.; Prevost, J.R.; Steger, N.A.; Wagner, R.K.

1971-01-31

A 30-Mwt prototype sodium-to-sodium intermediate heat exchanger and a 30-Mwt prototype sodium-heated steam generator were tested in combined operation in its Sodium Components Test Installation. This report contains the results of test and evaluation of the steam generator. During plant performance tests, performance degradation was observed, which resulted in the initiation of a diagnostic test series. This test series revealed that under certain operating conditions, the thermohydraulic characteristic of the steam generator changed either suddenly or gradually, resulting in overall performance degradation. A structural failure, requiring retirement of the unit, occurred before the diagnostic test series and analytical support effort were completed. This report describes the thermohydraulic and structural performance, including the structural failures, and related evaluation analyses of the Alco/BLH prototype steam generator performed prior to termination of the test and evaluation program. In addition, the report presents a post-test examination plan to obtain data that could possibly explain the cause of performance anomalies and structural failures experienced during testing.

C-Scan Performance Test of Under-Sodium ultrasonic Waveguide Sensor in Sodium

International Nuclear Information System (INIS)

Joo, Young Sang; Bae, Jin Ho; Kim, Jong Bum

2011-01-01

Reactor core and in-vessel structures of a sodium-cooled fast (SFR) are submerged in opaque liquid sodium in the reactor vessel. The ultrasonic inspection techniques should be applied for observing the in-vessel structures under hot liquid sodium. Ultrasonic sensors such as immersion sensors and rod-type waveguide sensors have developed in order to apply under-sodium viewing of the in-vessel structures of SFR. Recently the novel plate-type ultrasonic waveguide sensor has been developed for the versatile application of under-sodium viewing in SFR. In previous studies, the ultrasonic waveguide sensor module was designed and manufactured, and the feasibility study of the ultrasonic waveguide sensor was performed. To improve the performance of the ultrasonic waveguide sensor in the under-sodium application, a new concept of ultrasonic waveguide sensors with a Be coated SS304 plate is suggested for the effective generation of a leaky wave in liquid sodium and the non-dispersive propagation of A 0 -mode Lamb wave in an ultrasonic waveguide sensor. In this study, the C-scan performance of the under-sodium ultrasonic waveguide sensor in sodium has been investigated by the experimental test in sodium. The under-sodium ultrasonic waveguide sensor and the sodium test facility with a glove box system and a sodium tank are designed and manufactured to carry out the performance test of under-sodium ultrasonic waveguide sensor in sodium environment condition

Investigation for the sodium leak in Monju. Sodium leak and fire test-1

International Nuclear Information System (INIS)

Kawata, Koji; Ohno, Shuji; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi

2000-08-01

As a part of the work for investigating the sodium leak accident which occurred in the Monju reactor (hereinafter referred to as Monju) on December 8, 1995, three tests, (1) a sodium leak test, (2) a sodium leak and fire test-1, and (3) a sodium leak and fire test-II, were carried out at OEC/PEC. The main objectives of these tests were to confirm the leak and burning behavior of sodium from the damaged thermometer, and the effects of the sodium fire on the integrity of the surrounding structure. This report describes the results of the sodium fire test-I carried out as a preliminary test. The test was performed using the SOLFA-2 (Sodium Leak, Fire and Aerosol) facility on April 8, 1996. In this test, sodium heated to 480degC was leaked for approximately 1.5 hours from a leak simulating apparatus and caused to drop onto a ventilation duct and a grating with the same dimensions and layout as those in Monju. The main conclusions obtained from the test are shown below: 1) Observation from video cameras in the test revealed that in the early stages of the sodium leak, sodium dripped out of the flexible tube of the thermometer. This dripping and burning expanded in range as the sodium splashed on the duct. 2) No damage to the duct itself was detected. However, the aluminum louver frame of the ventilation duct's lower inlet was damaged. Its machine screws came off, leaving half of the grill (on the grating side) detached. 3) No large hole, like the one seen at Monju, was found when the grating was removed from the testing system for inspection, although the area centered on the point were the sodium dripped was damaged in a way indicating the first stages of grating failure. The 5mm square lattice was corroded through in some parts, and numerous blades (originally 3.2 mm thick) had become sharpened like the blade of a knife. 4) The burning pan underside thermocouple near the leak point measured 700degC in within approximately 10 minutes, and for the next hour remained

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)

Progress in the development and application of methods for cleaning and decontamination of components exposed to sodium

International Nuclear Information System (INIS)

Msika, D.; Lafon, A.

1978-01-01

In the technology of liquid sodium cooled fast reactors, the necessary processes for washing and decontamination have been demonstrated. For sodium removal, different solutions have been considered and tested in France. The studies have been progressively oriented toward defining a process using a fine dispersion of water in a gas (atomization). The results obtained by that method on non-radioactive components were satisfactory insofar as the efficiency and safety of the operation was concerned. The purpose of decontaminating components from the reactor primary circuits is to reduce the level of surface activity to a level compatible with personnel access without biological shielding. The treatment Is comprised of two stages: (i) washing, to remove any residual sodium, and (ii) decontamination which alternately applies alkaline and acid solutions, to dissolve the deposited radionuclides without significant attack on the surface. The treatment, recently applied to components from in-service reactors, generally met the design objective. (author)

Determination of the necessary parameters for a protection insulation disk of sodium circuit weighing system for thermomechanical and small component tests

International Nuclear Information System (INIS)

Bloch, M.; Cesar, S.B.G.

1985-01-01

The parameters requisited for a plastic disk used as thermal insulation, between feeding tank and weghing system, where the tank is supported are defined. The tank and weghing system are component parts of sodium circuit for thermomechanical and small component tests. During the circuit operation the temperature at tank reaches 600 0 C, however the temperature at weghing system should not reach 50 0 C. The temperature distribution in insulation disk is obtained by finit element method in function of thickness and thermal conductivity of material. The results obtained indicate for thickness E = 32 mm should be K 0 C and for E = 48 mm the thermal conductivity should be K 0 C. In both cases the pressure is σ > 14.5 Kgf/mm 2 . (M.C.K.) [pt

Development of sodium leak detection technology using laser resonance ionization mass spectrometry. Design and functional test using prototype sodium detection system

International Nuclear Information System (INIS)

Aoyama, Takafumi; Ito, Chikara; Harano, Hideki; Okazaki, Koki; Watanabe, Kenichi; Iguchi, Tetsuo

2009-01-01

In a sodium-cooled fast reactor, highly sensitive technology is required to detect small amounts of sodium leaking from the cooling system piping or components. The conventional sodium leak detectors have a fundamental difficulty in improving the detection sensitivity for a sodium leak because of the presence of salinity ( 23 NaCl) in the atmosphere around the components and piping of cooling systems. In order to overcome this problem, an innovative technology has been developed to selectively detect the radioactive sodium ( 22 Na) produced by a neutron reaction in the primary cooling system using Laser Resonance Ionization Mass Spectrometry (RIMS). In this method, sodium ions produced with the two processes of (1) atomization of sodium aerosols and (2) resonance ionization of sodium atom, are detected selectively using a time-of-flight mass spectrometer. The 22 Na can be distinguished from the stable isotope ( 23 Na) by mass spectrometry, which is the advantage of RIMS comparing to the other methods. The design and the construction of the prototype system based on fundamental experiments are shown in the paper. The aerodynamic lens was newly introduced, which can transfer aerosols at atmospheric pressure into a vacuum chamber while increasing the aerosol density at the same time. Furthermore, the ionization process was applied by using the external electric field after resonance exciting from the ground level to the Rydberg level in order to increase the ionization efficiency. The preliminary test results using the stable isotope ( 23 Na) showed that prototype system could easily detect sodium aerosol of 100 ppb, equivalent to the sensitivity of the conventional detectors. (author)

Development of the immersed sodium flowmeter

International Nuclear Information System (INIS)

Chen Daolong

1994-09-01

An immersed sodium flowmeter of the range 3 m 3 /h is developed. It is a flowmeter of entire-sealed construction, it can be operated in sodium. Its construction, the theoretical calculation of the calibration characteristic and the pressure loss, the test facility and the calibration test are presented in detail. It analytical expression of the calibration characteristic in the temperature limit 200∼600 degree C and the error analysis are given. The basic error of this immersed sodium flowmeter is below +-2.3% of the measuring range. The immersed sodium flowmeter can be used to resolve the sodium flowrate measuring problems of the in-reactor component of LMFBR, for example, the flowrate measuring of the in-reactor sodium purification loop, the flowrate measuring of the immersed sodium pump and the flowrate measuring of the in-reactor test component

Japanese studies on sodium fires, design and testing

International Nuclear Information System (INIS)

Mitsutsuka, N.; Yoshida, N.

1983-01-01

Considerations of sodium fires are very important for the design and licensing of LMFBRs. Continuing effort has been made in the study of sodium fires and their consequences since the beginning of the Japanese fast breeder reactor development program. Recent effort is mainly focussed on studies related to Monju, especially on the design and testing of primary cell liners against large sodium spills. Experimental and analytical studies on sodium fires, water release from concrete and sodium concrete reactions are conducted as a part of this study. Some extinguishing agents are also tested against sodium fires. In addition, considerable effort is being made in the development of detection systems for the small sodium leaks before a pipe rupture. This paper briefly summarizes the Japanese status of these sodium fire related activities conducted by Fast Breeder Reactor Development Project of the Power Reactor and Nuclear Fuel Development Corporation (PNC)

Sodium technology. 1 - FFTF support work: friction tests, January-March 1974

International Nuclear Information System (INIS)

Kurzeka, W.J.; Oliva, R.M.; Horton, P.

1974-01-01

The objective of this program is to conduct friction screening tests in an environment of high-temperature, high-purity liquid sodium or sodium vapor to: (1) develop backup materials, processes, and vendors for core component wear pads, (2) investigate material treatments and coatings for improvement of wear behavior of common LMFBR structural materials, (3) evaluate weld-deposited hardfacings and/or prefabricated bearing materials for use in long-term, high-temperature, high-fluence regions, (4) evaluate bearing materials with a low potential for change in surface composition due to corrosion or mass transfer effects, and (5) develop statistical confidence in friction values for selected material combinations

Large scale sodium interactions. Part 1. Test facility design

International Nuclear Information System (INIS)

King, D.L.; Smaardyk, J.E.; Sallach, R.A.

1977-01-01

During the design of the test facility for large scale sodium interaction testing, an attempt was made to keep the system as simple and yet versatile as possible; therefore, a once through design was employed as opposed to any type of conventional sodium ''loop.'' The initial series of tests conducted at the facility call for rapidly dropping from 20 kg to 225 kg of sodium at temperatures from 825 0 K to 1125 0 K into concrete crucibles. The basic system layout is described. A commercial drum heater is used to melt the sodium which is in 55 gallon drums and then a slight argon pressurization is used to force the liquid sodium through a metallic filter and into a dump tank. Then the sodium dump tank is heated to the desired temperature. A diaphragm is mechanically ruptured and the sodium is dumped into a crucible that is housed inside a large steel test chamber

Sodium tests on an integrated purification prototype for a sodium-cooled fast breeder reactor

International Nuclear Information System (INIS)

Abramson, R.

1984-04-01

This paper describes sodium tests performed on the integrated primary sodium purification prototype of the Creys Malville Super Phenix 1 fast breeder reactor. These tests comprised: - hydrostatic test, - thermal tests, - handling tests. They enabled a number of new technological arrangements to be qualified and provided the necessary information for the design and construction of the Super Phenix 1 purification units

The water vapor nitrogen process for removing sodium from LMFBR components

Energy Technology Data Exchange (ETDEWEB)

Crippen, M D; Funk, C W; Lutton, J M [Hanford Engineering Development Laboratory, Richland (United States)

1978-08-01

Application and operation of the Water Vapor-Nitrogen Process for removing sodium from LMFBR components is reviewed. Emphasis is placed on recent efforts to verify the technological bases of the process, to refine the values of process parameters and to ensure the utility of the process for cleaning and requalifying components. (author)

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)

Design of sodium cooled reactor systems and components for maintainability

International Nuclear Information System (INIS)

Carr, R.W.; Charnock, H.O.; McBride, J.P.

1978-09-01

Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

FY 2017 Status of Sodium Freezing and Remelting Tests

Energy Technology Data Exchange (ETDEWEB)

Lv, Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Boron, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chojnowski, D. B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. B. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-15

The Sodium Freezing and Remelting experiment facility at Argonne National Laboratory has been significantly modified and improved. The main improvement was replacement of the two original stainless steel test sections that had strain gages limited by their bonds to the stainless steel to maximum temperatures of 350°C with a single new test section with strain gages that can be utilized up to 980°C and a thin wall to enhance measured strains. Wetting of stainless steel by sodium within a practical time of one to a few days is expected to require temperatures of 450°C or greater. Thus, the higher temperature strain gages enable wetting in a short time of a few days. Wetting below 350°C would have required an impractically long time of at least weeks. Other improvements included upgrading of the loop configuration, incorporation of a cold finger to purify sodium, a new data acquisition system, and reinstallation of the many heaters, heater controllers, and thermocouples. After the loop had been heated to 400°C for about two hours, an initial sodium freezing test was conducted. It is thought that the sodium might have at least partially wetted the stainless steel wall under these conditions. The strain gage measurements indicate that an incremental step inward deformation of the test section thin wall occurred as the temperature decreased through the sodium freezing temperature. This behavior is consistent with sodium initially adhering to the stainless steel inner wall but breaking away from the wall as the freezing sodium contracted. Conduct of additional sodium freezing tests under well wetted conditions was prevented as a result of stoppage of all electrical work at Argonne by the Laboratory Director on July 25, 2017. A pathway to resuming electrical work is now in place at Argonne and additional sodium freezing testing will resume next fiscal year.

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.

Intermediate-scale sodium-concrete reaction tests with basalt and limestone concrete

International Nuclear Information System (INIS)

Hassberger, J.A.; Muhlestein, L.D.

1981-01-01

Ten tests were performed to investigate the chemical reactions and rate and extent of attack between sodium and basalt and limestone concretes. Test temperatures ranged from 510 to 870 0 C (950 to 1600 0 F) and test times from 2 to 24 hours. Sodium hydroxide was added to some of the tests to assess the impact of a sodium hydroxide-aided reaction on the overall penetration characteristics. Data suggest that the sodium penetration of concrete surfaces is limited. Penetration of basalt concrete in the presence of sodium hydroxide is shown to be less severe than attack by the metallic sodium alone. Presence of sodium hydroxide changes the characteristics of sodium penetration of limestone concrete, but no major differences in bulk penetration were observed as compared to penetration by metallic sodium

Preliminary safety evaluation (PSE) for Sodium Storage Facility at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Bowman, B.R.

1994-01-01

This evaluation was performed for the Sodium Storage Facility (SSF) which will be constructed at the Fast Flux Test Facility (FFTF) in the area adjacent to the South and West Dump Heat Exchanger (DHX) pits. The purpose of the facility is to allow unloading the sodium from the FFTF plant tanks and piping. The significant conclusion of this Preliminary Safety Evaluation (PSE) is that the only Safety Class 2 components are the four sodium storage tanks and their foundations. The building, because of its imminent risk to the tanks under an earthquake or high winds, will be Safety Class 3/2, which means the building has a Safety Class 3 function with the Safety Class 2 loads of seismic and wind factored into the design

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

Outline of sodium-water reaction test in case of large leak with SWAT-3 testing equipments

International Nuclear Information System (INIS)

Sato, Minoru

1978-01-01

The key component in sodium-cooled fast reactors in steam generators, and the sodium-water reaction owing to the break of heating tubes may cause serious damages in equipments and pipings. The main factor controlling this phenomenon is the rate of leak of water. When the rate of water leak is small, the propagation of heating tube breaking may occur owing to ''wastage phenomenon'', on the other hand, when the rate of water leak is large, the phenomena of explosive pressure and flow occur due to the reaction heat and a large quantity of hydrogen generated by the reaction. In PNC, the testing equipments of SWAT-2 for small water leak and SWAT-1 for large leak were constructed, and the development test has been carried out to establish the method of safety design experimentally. The synthetic test equipment for the safety of steam generators, SWAT-3, was constructed to carry out the large water leak test in the scale close to actual plants. The object of the test, the outline of the test equipment, the phenomena of pressure and flow in the water injection test, the confirmation of the occurrence of secondary breaking of adjacent heating tubes, and the disposal of reaction products are described in this paper. This test is till going on, and the final conclusion will be reported later. (Kako, I.)

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)

Sodium-fuel interaction: dropping experiments and subassembly test

International Nuclear Information System (INIS)

Holtbecker, H.; Schins, H.; Jorzik, E.; Klein, K.

1978-01-01

Nine dropping tests, which bring together 2 to 4 kg of molten UO 2 with 150 l sodium, showed the incoherency and non-violence of these thermal interactions. The pressures can be described by sodium incipient boiling and bubble collapse; the UO 2 fragmentation by thermal stress and bubble collapse impact forces. The mildness of the interaction is principally due to the slowness and incoherency of UO 2 fragmentation. This means that parametric models which assume instantaneous mixing and fragmentation are of no use for the interpretation of dropping experiments. One parametric model, the Caldarola Fuel Coolant Interaction Variable Mass model, is being coupled to the two dimensional time dependent hydrodynamic REXCO-H code. In a first step the coupling is applicated to a monodimensional geometry. A subassembly test is proposed to validate the model. In this test rapid mixing between UO 2 and sodium has to be obtained. Dispersed molten UO 2 fuel is obtained by flashing injected sodium drops inside a UO 2 melt. This flashing is theoretically explained and modelled as a superheat limited explosion. The measured sodium drop dwell times of two experiments are compared to results obtained from the mentioned theory, which is the basis of the Press 2 Code

A summary of sodium vapor trap experience at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Schuck, W.J.

1986-09-01

Sodium vapor trap operation at the Fast Flux Test Facility has been successful although not uneventful. Analysis and evaluation of the behavior of the vapor traps associated with reactor cover gas processing and analysis systems has confirmed their design and has led to an improved understanding of these components and the environment in which they operate. This knowledge will permit simplification and reduced costs for future designs

Ten Year Operating Test Results and Post-Test Analysis of a 1/10 Segment Stirling Sodium Heat Pipe, Phase III

Science.gov (United States)

Rosenfeld, John, H; Minnerly, Kenneth, G; Dyson, Christopher, M.

2012-01-01

High-temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, Stirling cycle heat sources; and with the resurgence of space nuclear power both as reactor heat removal elements and as radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly, long-term materials compatibility is being evaluated through the use of high-temperature life test heat pipes. Thermacore, Inc., has carried out a sodium heat pipe 10-year life test to establish long-term operating reliability. Sodium heat pipes have demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 87,000 hr (10 yr) at nearly 700 C. These life test results have demonstrated the potential for high-temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and post-test analysis of the heat pipe and sodium working fluid are described.

Fast Flux Test Facility replacement of a primary sodium pump

International Nuclear Information System (INIS)

Krieg, S.A.; Thomson, J.D.

1985-01-01

The Fast Flux Test Facility is a 400 MW Thermal Sodium Cooled Fast Reactor operated by Westinghouse Hanford Company for the US Department of Energy. During startup testing in 1979, the sodium level in one of the primary sodium pumps was inadvertently raised above the normal height. This resulted in distortion of the pump shaft. Pump replacement was carried out using special maintenance equipment. Nuclear radiation and contamination were not significant problems since replacement operations were carried out shortly after startup of the Fast Flux Test Facility

Thermal performance tests on a sodium-to-sodium heat exchanger

International Nuclear Information System (INIS)

Prahlad, B.; Kale, R.D.; Rajan, K.K.

1990-01-01

Thermal performance of a 3 MW sodium-to-sodium intermediate heat exchanger (IHX) was evaluated under temperature conditions typical of a Fast Breeder Reactor IHX. A regenerative figure of eight loop was used with the heat exchanger at the cross over point, and a 500 kW heat source and an air cooled sink to maintain the desired test conditions. The overall heat transfer coefficient was found to vary from 4.02 to 4.87 kW/m 2 ·K for Peclet numbers varying from 37 to 112.5 on the shell side and 44.4 to 133.5 on the tube side respectively. The Peclet numbers were representative of low turbulent regime in this case. While the overall heat transfer coefficient was found close to predictions using Lubarsky's correlation, it was somewhat lower than that predicted by later correlations of Spukunsky and Borishansky. The reasons for the lower overall heat transfer coefficient have been explained in terms of possible maldistribution of shell side flow in low turbulent regime reducing the effective heat transfer area and increased thermal contact resistance. Based on their findings the authors feel that heat transfer in a sodium-to-sodium heat exchanger at low Peclet numbers is expected to differ from that obtained with large Peclet numbers. (author)

A summary of sodium vapor trap experience at the Fast Flux Test Facility

Energy Technology Data Exchange (ETDEWEB)

Schuck, W J [Westinghouse Hanford Company, Richland, WA (United States)

1987-07-01

Sodium vapor trap operation at the Fast Flux Test Facility has been successful although not uneventful. Analysis and evaluation of the behavior of the vapor traps associated with reactor cover gas processing and analysis systems has confirmed their design and has led to an improved understanding of these components and the environment in which they operate. This knowledge will permit simplification and reduced costs for future designs (author)

Removal of radioactive sodium from experimental breeder reactor-II components and conversion to a disposable solid waste: alcohol recovery

International Nuclear Information System (INIS)

Krusl, J.R.; Washburn, R.A.

1985-01-01

Radioactive sodium is removed from Experimental Breeder Reactor-II components by immersing the components in denatured alcohol until the sodium has reacted with the alcohol. The resulting radioactive sodium-alcohol solution must be processed to separate and convert the sodium to a solid waste for disposal. A process was developed and is described that converts radioactive sodium dissolved in alcohol to a dry powdered carbonate waste product and recovers the alcohol for reuse. The sodium-alcohol waste solution, after adjustment for proper sodium and water content, is fed to a wiped-film evaporator operated at 190 0 C and maintained with a CO 2 atmosphere that converts the dissolved sodium to anhydrous Na 2 CO 3 . The end product, about85 to 90 wt% Na 2 CO 3 , is directed into a 208-l (55-gal) drum for disposal. Alcohol distilled during the process is condensed, collected, and dried for immediate reuse. The composition of the alcohol is not altered in the process

Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0

Energy Technology Data Exchange (ETDEWEB)

Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-11-01

The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)

Seismic tests in sodium of the SPX-1 primary pump shaft carried out in the CPV-1 test rig at ENEA-Brasimone

International Nuclear Information System (INIS)

Contardi, T.; Rapezzi, L.; Le Coz, P.; Tigeot, Y.; Partiti, C.; Zola, M.; Denimal, P.

1988-01-01

Dynamic tests were carried out by ISMES, on behalf of ENEA and CEA and in co-operation with FIAT/TTG, on a SPX-1 primary pump shaft. These tests were conducted, mainly in sodium, in the CPV-1 test rig at the ENEA Brasimone Center. The excitation was applied to the flange supporting the hydrostatic bearing. After some preliminary analysis performed in absence of liquid sodium and at ambient temperature, the following tests were performed on the rig filled with sodium at operating temperature: (A) sine sweeps between 1 and 15 Hz, (B) ambient vibration investigation, and (C) seismic tests with a SSE acceleration time-history (20 s duration) calculated by CEA at hydrostatic bearing level. Two sets of seismic tests were carried out, each time increasing amplitudes up to 70% of SSE. This value was not exceeded for safety reasons and actuator power limit. The first set of tests began in nominal operating conditions; when 70% of SSE was reached, pressure feed to hydrostatic bearing was reduced lowering its effective support. This simulated a larger earthquake. The second set of tests was representative of SPX-1 pump actual operating conditions, because both hydrostatic bearing pressure and shaft rotating speed were simultaneously reduced following the primary pump characteristic curve. The tests allowed the SPX-1 pump rotating set to be widely qualified. Among the main results, it is worth noting that the stiffness of the hydrostatic bearing system was generally compatible with seismic requirements. Finally, it is worth pointing out that, in order to allow the above-mentioned tests to be carried out, a full seismic qualification of the CPV-1 test rig was necessary: thus, this rig might be used in the future for further seismic tests on LMFBR components and systems in sodium. (author). Figs and tabs

Sodium cleaning from sodium contaminated components and operation for experimental equipment

Energy Technology Data Exchange (ETDEWEB)

Kim, B. H.; Kim, J. M.; Kim, T. J.; Nam, H. Y.; Jeong, J. Y.; Choi, B. H.; Choi, J. H

2007-11-15

An objective of washing technology development for sodium contaminated equipment is to clean and reuse safely and effectively the used equipment through a washing and maintenance, and recovery of the sodium wastes generated during washing.

Fast Flux Test Facility sodium pump operating experience - mechanical

International Nuclear Information System (INIS)

Buonamici, R.

1987-11-01

The Heat Transport System (HTS) pumps were designed, fabricated, tested, and installed in the Fast Flux Test Facility (FFTF) Plant during the period from September 1970 through July 1977. Since completion of the installation and sodium fill in December 1978, the FFTF Plant pumps have undergone extensive testing and operation with HTS testing and reactor operation. Steady-state hydraulic and mechanical performances have been and are excellent. In all, FFTF primary and secondary pumps have operated in sodium for approximately 75,000 hours and 79,000 hours, respectively, to August 24, 1987

GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

Energy Technology Data Exchange (ETDEWEB)

S. A. Eide; S. V. Chmielewski; T. D. Swantz

1990-02-01

A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

Stress corrosion on austenitic stainless steels components after sodium draining

International Nuclear Information System (INIS)

Champeix, L.; Baque, P.; Chairat, C.

1980-04-01

The damage study performed on 316 pipes of a loop after two leakages allows to conclude that a stress corrosion process in sodium hydroxide environment has induced trans-crystaline cracks. The research of conditions inducing such a phenomenon is developed, including parametric tests under uniaxial load and some tests on pipe with welded joints. In aqueous sodium hydroxide, two corrosion processes have been revealed: a general oxidization increasing with environment aeration and a transcrystalline cracking appearing for stresses of the order of yield strength. Other conditions such a temperature (upper than 100 0 C) and time exposures (some tens of hours) are necessary. Cautions in order to limit introduction of wet air into drained loop and a choice of appropriate preheating conditions when restarting the installation must permit to avoid such a type of incident

Flame photometric detection of sodium leaks: Tests on a fullscale model for the control gear sodium rig

International Nuclear Information System (INIS)

Grundy, B.R.; Knowles, P.

1971-01-01

The proposed arrangement for detecting sodium leaks from the large flanges of the Control Gear Sodium Rig (Test Section No. 8, MCTR) at REML is to jacket then in a secondary containment from which air samples will be continuously pumped. Pipework feeds the air to a flame photometer which responds if soditun is present. To prove that sodium smoke could be transferred through the system, tests were performed on a fullscale model by burning small amounts of sodium in different jackets. Large signals free from fluctuations were obtained in all tests, peak response occurring in 2 1/4 minutes or less. The signal quickly cleared after isolating the appropriate vessel. A waiting period of several hours was sufficient to reduce the signal to zero, no cleaning of pipework, etc being necessary. In contrast, samples of two lagging materials heated to 400 °C gave no response with the photometer at maximum sensitivity. (author)

Solubilities of sodium nitrate, sodium nitrite, and sodium aluminate in simulated nuclear waste

International Nuclear Information System (INIS)

Reynolds, D.A.; Herting, D.L.

1984-09-01

Solubilities were determined for sodium nitrate, sodium nitrite, and sodium aluminate in synthetic nuclear waste liquor. Solubilities were determined as a function of temperature and solution composition (concentrations of sodium hydroxide, sodium nitrate, sodium nitrite, and sodium aluminate). Temperature had the greatest effect on the solubilities of sodium nitrate and sodium nitrite and a somewhat lesser effect on sodium aluminate solubility. Hydroxide had a great effect on the solubilities of all three salts. Other solution components had minor effects. 2 references, 8 figures, 11 tables

Experimental plans for LMFBR cavity liner sodium spill test LT-1

International Nuclear Information System (INIS)

Hilliard, R.K.; Newell, G.A.

1976-01-01

Reinforced concrete is an important material of construction in LMFBR cavities and cells. Steel liners are often installed on the concrete surfaces to provide a gastight seal for minimizing air inleakage to inerted cell atmospheres and to protect the concrete from direct contact with sodium in the event of a sodium spill. In making safety assessment analyses, it is of interest to determine the adequacy of the liners to maintain their leaktightness during postulated accidents involving large sodium spills. However, data for basing analytical assessments of cell liners are very meager and an experimental program is underway at HEDL to provide some of the needed information. The HEDL cell liner evaluation program consists of both bench-scale feature tests and large-scale sodium spill demonstration tests. The plans for the first large-scale sodium spill test (LT-1) are the subject of this paper

Mechanical properties test program on structural materials in a sodium environment

International Nuclear Information System (INIS)

Natesan, K.; Chopra, O.K.; Kassner, T.F.

1979-10-01

This document describes in detail the ongoing and planned US Test program on the mechanical properties of sodium-exposed Type 316 austenitic stainless and Fe-2 1/4 Cr-1 Mo ferritic steels. The test program is based on the Development Requirement Specifications (DRS) established by the DOE/Clinch River Breeder Reactor Project (CRBRP) Program Office, the general need for the development of LMFBR structural-design criteria established by the Nuclear Systems Materials Handbook, and the need for a fundamental understanding of materials behavior in a sodium environment, which is generic to LMFBR systems. The planned test program is an extension of work based on current knowledge of sodium chemistry and the influence of sodium purity on the mechanical properties of structural materials

Sodium cleaning and disposal methods in experimental facilities

International Nuclear Information System (INIS)

Rajan, K.K.; Gurumoorthy, K.; Rajan, M.; Kale, R.D.

1997-01-01

At Indira Gandhi Centre for Atomic Research, major sodium facilities are designed and operated at Engineering Development Group as a part of development programme towards experimental and Prototype Fast Reactor. After the test programme many equipment and components were removed from the sodium facilities and sodium removal and disposal was carried out. The experience gained in different cleaning methods and waste sodium disposal are discussed. (author)

Synthesis of results obtained on sodium components and technology through the Generation IV International Forum SFR Component Design and Balance-of-Plant Project

International Nuclear Information System (INIS)

Sienicki, J.J.; Rodriguez, G.; Kisohara, N.; Kim, J. B.; Gerber, A.; Ashurko, Y.; Toyama, S.

2013-01-01

Status: The viability of designing SFR components and BOP has been demonstrated with design, construction and operation of previous sodium-cooled reactors. The main objective of this R&D project is related to system performance, or by development on the use of AECS in the BOP that could allow further cost improvements. Objective: To conduct collaborative research and development of components and BOP for the SFR System. The Project has to satisfy the GIF’s criteria of safety, economy, sustainability, proliferation resistance and physical protection. Activities within this Project are addressing experimental and analytical evaluation of advanced ISI&R, LBB assessment, development of AECS with Brayton cycles, advanced SG technologies. Project activities will be based in part on the extensive historical R&D experience with component design and balance of plant for sodium-cooled fast reactors

Prediction of EMP cavitation threshold from other than sodium testing

International Nuclear Information System (INIS)

Kambe, M.; Kamei, M.

2002-01-01

An experimental study has been performed to predict the cavitation threshold of electromagnetic pumps from measurements on test models using water and alcohol. Cavitation tests were carried out using water and alcohol test loop on subscale ducts of transparent acrylic resin with reference to an actual pump (1.1m 3 /min). These data were compared to those obtained from the in-sodium tests on the actual pump. The investigation revealed that the value of Thoma's dimensionless parameter: σ applied to the test model for water and alcohol is quite higher than that of corresponding σ on the actual pump. To minimize the incipient cavitation safety margin, more accurate prediction must be required. In view of this, the authors proposed the dimensionless parameter: σ T =σ/W-bare where W-bare denotes the Weber number. This parameter was confirmed to predict the cavitation threshold of electromagnetic pumps with much more accuracy than ever before. It can also be adopted to predict cavitation threshold of other FBR components. (author)

The re-emergence of sodium ion batteries: testing, processing, and manufacturability

Science.gov (United States)

Roberts, Samuel; Kendrick, Emma

2018-01-01

With the re-emergence of sodium ion batteries (NIBs), we discuss the reasons for the recent interests in this technology and discuss the synergies between lithium ion battery (LIB) and NIB technologies and the potential for NIB as a “drop-in” technology for LIB manufacturing. The electrochemical testing of sodium materials in sodium metal anode arrangements is reviewed. The performance, stability, and polarization of the sodium in these test cells lead to alternative testing in three-electrode and alternative anode cell configurations. NIB manufacturability is also discussed, together with the impact that the material stability has upon the electrodes and coating. Finally, full-cell NIB technologies are reviewed, and literature proof-of-concept cells give an idea of some of the key differences in the testing protocols of these batteries. For more commercially relevant formats, safety, passive voltage control through cell balancing and cell formation aspects are discussed. PMID:29910609

Sodium removal by alcohol process: Basic tests and its application

International Nuclear Information System (INIS)

Nakai, S.; Yamamoto, S.; Akai, M.; Yatabe, T.

1997-01-01

We have various methods for sodium removal; an alcohol cleaning process, a steam cleaning process and a direct burning process. Sodium removal by the alcohol process has a lot of advantages, such as causing no alkali corrosion to steel, short processing time and easy operation. Therefore the alcohol process was selected for the 1MWt double wall tube straight type steam generator. We have already had some experiences of the alcohol process, while still needed to confirm the sodium removal rate in the crevice and to develop an on-line sodium concentration monitoring method in alcohol during sodium removal. We have conducted the small scale sodium removal test with flowing alcohol where the sodium removal rate in the crevice and the alcohol conductivity were measured as functions of sodium concentration in alcohol and alcohol temperature. The sodium removal of the DWTSG was conducted by the devised alcohol process safely and efficiently. The process hour was about 1 day. Visual inspection during dismantling of the DWTSG showed no evidence of any un-reacted sodium. (author)

Flame generation of sodium chloride aerosol for filter testing

International Nuclear Information System (INIS)

Edwards, J.; Kinnear, D.I.

1975-01-01

A generator for sodium chloride aerosol is described, which when used in conjunction with a sensitive portable sodium flame detector unit, will permit the in-place testing of large filter installations having air throughputs up to about 80,000 m 3 /h, at penetrations down to at least 0.005 percent. (U.S.)

Effect of metal surface composition on deposition behavior of stainless steel component dissolved in liquid sodium

International Nuclear Information System (INIS)

Yokota, Norikatsu; Shimoyashiki, Shigehiro

1988-01-01

Deposition behavior of corrosion products has been investigated to clarify the effect of metal surface composition on the deposition process in liquid sodium. For the study a sodium loop made of Type 304 stainless steel was employed. Deposition test pieces, which were Type 304 stainless steel, iron, nickel or Inconel 718, were immersed in the sodium pool of the test pot. Corrosion test pieces, which were Type 304 stainless steel, 50 at% Fe-50 at%Mn and Inconel 718, were set in a heater pin assembly along the axial direction of the heater pin surface. Sodium temperatures at the outlet and inlet of the heater pin assembly were controlled at 943 and 833 K, respectively. Sodium was purified at a cold trap temperature of 393 K and the deposition test was carried out for 4.3 x 10 2 - 2.9 x 10 4 ks. Several crystallized particles were observed on the surface of the deposition test pieces. The particles had compositions and crystal structures which depended on both the composition of deposition test pieces and the concentration of iron and manganese in sodium. Only iron-rich particles having a polyhedral shape deposited on the iron surface. Two types of particles, iron-rich α-phase and γ-phase with nearly the same composition as stainless steel, were deposited on Type 304 stainless steel. A Ni-Mn alloy was deposited on the nickel surface in the case of a higher concentration of manganese in sodium. On the other hand, for a lower manganese concentration, a Fe-Ni alloy was precipitated on the nickel surface. Particles deposited on nickel had a γ-phase crystal structure similar to the deposition test piece of nickel. Hence, the deposition process can be explained as follows: Corrosion products in liquid sodium were deposited on the metal surface by forming a metal alloy selectively with elements of the metal surface. (author)

Large scale sodium interactions. Part 2. Preliminary test results for limestone concrete

International Nuclear Information System (INIS)

Smaardyk, J.E.; Sutherland, H.J.; King, D.L.; Dahlgren, D.A.

1977-01-01

Any sodium cooled reactor system must consider the interaction of hot sodium with cell liners, and given either a failed liner or a hypothetical core disruptive accident, the interaction of hot sodium with concrete. The data base available for safety assessments involving these interactions is limited, especially for the concrete and failed liner interactions. To better understand what happens when hot sodium comes in contact with concrete, a series of tests is being carried out to investigate sodium-concrete reactions under conditions which are similar to actual reactor accident conditions. Tests cover the cases of sodium spills on bare concrete and on cells with defective steel liners. Specific objectives have been to obtain a complete description of the sodium/concrete interaction including heat balance, gas evolution and flow, movement and heat generation of the reaction zone, reaction product formation, and the layering or movement of the products

Posterior cerebral artery Wada test: sodium amytal distribution and functional deficits

Energy Technology Data Exchange (ETDEWEB)

Urbach, H.; Schild, H.H. [Dept. of Radiology/Neuroradiology, Univ. of Bonn (Germany); Klemm, E.; Biersack, H.J. [Bonn Univ. (Germany). Klinik fuer Nuklearmedizin; Linke, D.B.; Behrends, K.; Schramm, J. [Dept. of Neurosurgery, Univ. of Bonn (Germany)

2001-04-01

Inadequate sodium amytal delivery to the posterior hippocampus during the intracarotid Wada test has led to development of selective tests. Our purpose was to show the sodium amytal distribution in the posterior cerebral artery (PCA) Wada test and to relate it to functional deficits during the test. We simultaneously injected 80 mg sodium amytal and 14.8 MBq {sup 99} {sup m}Tc-hexamethylpropyleneamine oxime (HMPAO) into the P2-segment of the PCA in 14 patients with temporal lobe epilepsy. To show the skull, we injected 116 MBq {sup 99} {sup m}Tc-HDP intravenously. Sodium amytal distribution was determined by high-resolution single-photon emission computed tomography (SPECT). In all patients, HMPAO was distributed throughout the parahippocampal gyrus and hippocampus; it was also seen in the occipital lobe in all cases and in the thalamus in 11. Eleven patients were awake and cooperative; one was slightly uncooperative due to speech comprehension difficulties and perseveration. All patients showed contralateral hemianopia during the test. Four patients had nominal dysphasia for 1-3 min. None developed motor deficits or had permanent neurological deficits. Neurological deficits due to inactivation of extrahippocampal areas thus do not grossly interfere with neuropsychological testing during the test. (orig.)

Development of blow down and sodium-water reaction jet analysis codes-Validation by sodium-water reaction tests (SWAT-1R)

International Nuclear Information System (INIS)

Hiroshi Seino; Akikazu Kurihara; Isao Ono; Koji Jitsu

2005-01-01

Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed in order to improve the evaluation method on sodium-water reaction event in the steam generator (SG) of a sodium cooled fast breeder reactor (FBR). The validation analyses by these two codes were carried out using the data of Sodium-Water Reaction Test (SWAT-1R). The following main results have been obtained through this validation: (1) The calculational results by LEAP-BLOW such as internal pressure and water flow rate show good agreement with the results of the SWAT- 1R test. (2) The LEAP-JET code can qualitatively simulate the behavior of sodium-water reaction. However, it is found that the code has tendency to overestimate the maximum temperature of the reaction jet. (authors)

Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor

International Nuclear Information System (INIS)

Nashine, B.K.; Rao, B.P.C.

2014-01-01

Highlights: • Derivation of applicable design equations. • Design of an annular induction pump based on these equations. • Testing of the designed pump in a sodium test facility. • Performance evaluation of the designed pump. - Abstract: Annular linear induction pumps (ALIPs) are used for pumping electrically conducting liquid metals. These pumps find wide application in fast reactors since the coolant in fast reactors is liquid sodium which a good conductor of electricity. The design of these pumps is usually done using equivalent circuit approach in combination with numerical simulation models. The equivalent circuit of ALIP is similar to that of an induction motor. This paper presents the derivation of equivalent circuit parameters using first principle approach. Sodium testing of designed ALIP using the equivalent circuit approach is also described and experimental results of the testing are presented. Comparison between experimental and analytical calculations has also been carried out. Some of the reasons for variation have also been listed in this paper

Compatibility of steels for fast breeder reactor in high temperature sodium

International Nuclear Information System (INIS)

Yuhara, Shunichi

1981-01-01

In recent years, considerable progress has been made and experience has been obtained for material applicability in sodium-cooled fast breeder reactors. In this report, materials, principal dimensions and sodium conditions for the reactor system components, which include fuel pin cladding, intermediate heat exchangers, steam generators and pipings, are reviewed with emphasis on the thin-walled, high temperature and high strength components. The corrosion, mechanical and tribological behavior in sodium of important materials used for the reactor components, such as Types 304 and 316 stainless steel and 2 1/4Cr-1Mo steel, are discussed on the basis of characteristic testing results. Furthermore, material requirements concerned with compatibility in sodium are summarized from this review and discussion. (author)

Sodium distiller II

International Nuclear Information System (INIS)

Goncalves, A.C.; Castro, P.M. e; Torres, A.R.; Correa, S.M.

1990-01-01

A sodium distiller allows the evaluation of the sodium purity, contained in plants and circuits of Fast Reactors. The sodium distillers of the IEN Reactor's Department was developed initially as a prototype, for the testing of the distillation process and in a second step, as a equipment dedicated to attendance the operation of these circuits. This last one was build in stainless steel, with external heat, rotating crucible of nickel for four samples, purge system for pipe cleaning and a sight glass that permits the observation of the distillation during all the operation. The major advantage of this equipment is the short time to do a distillation operation, which permits its routine utilization. As a consequence of the development of the distillers and its auxiliary systems an important amount of new information was gathered concerning components and systems behaviour under high temperature, vacuum and sodium. (author)

Testing experience with fast flux test facility

International Nuclear Information System (INIS)

Noordhoff, B.H.; McGough, C.B.; Nolan, J.E.

1975-01-01

Early FFTF project planning emphasized partial and full-scale testing of major reactor and plant prototype components under expected environmental conditions, excluding radiation fields. Confirmation of component performance during FFTF service was considered essential before actual FFTF startup, to provide increased assurance against FFTF startup delays or operational difficulties and downtime. Several new sodium facilities were constructed, and confirmation tests on the prototype components are now in progress. Test conditions and results to date are reported for the primary pump, intermediate heat exchanger, sodium-to-air dump heat exchanger, large and small sodium valves, purification cold trap, in-vessel handling machine, instrument tree, core restraint, control rod system, low-level flux monitor, closed loop ex-vessel machine, refueling equipment, and selected maintenance equipment. The significance and contribution of these tests to the FFTF and Liquid Metal Fast Breeder Reactor (LMFBR) program are summarized. (U.S.)

Accounting sodium effect in calculation of strength of nuclear reactor components

International Nuclear Information System (INIS)

Nikitin, V.I.

1981-01-01

Accounting methods of liquid sodium effect on long-term strength and creep of structural materials of nuclear reactors are considered. The decrease of pearlite steel strength at the decarburization expense and the decrease of plasticity of austenitic steels at the expense of carburization are noted. The necessity to account thermal transfer of mass is shown. Values of safety factors are presented, they are recommended for the design of reactor component parts with the thickness not less than 1 mm [ru

Air-cleaning systems for sodium-fire-aerosol control

International Nuclear Information System (INIS)

Hilliard, R.K.; Muhlestein, L.D.

1982-05-01

A development program has been carried out at the Hanford Engineering Development Laboratory (HEDL) with the purpose of developing and proof testing air cleaning components and systems for use under severe sodium fire conditions, including those involving high levels of radioactivity. The air cleaning components tested can be classified as either dry filters or aqueous scrubbers. Test results are presented

Equipment for long-term testing of material creep in liquid sodium

International Nuclear Information System (INIS)

Dufek, F.; Walder, V.; Cech, V.; Winkler, P.

1980-01-01

A sodium test plant is described and the methods are shown of securing the desired test specifications. The facility is used for the long-term testing of nonsteady material creep under the action of temperature and mechanical stress due to the static or pulsed overpressure of inert gas inside a tube specimen and a simultaneous effect of sodium flow on the outside wall. The test results are to be used for comparing Czechoslovak-made steels with foreign materials and also for testing the degradation effects of the above phenomena on the standardized long-term properties in inert and steady state conditions. (B.S.)

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 Exposure Tests on Limestone Concrete Used as Sacrificial Protection Layer in FBR

International Nuclear Information System (INIS)

Parida, F.C.; Das, S.K.; Sharma, A.K.; Rao, P.M.; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kasinathan, N.

2006-01-01

Hot sodium coming in contact with structural concrete in case of sodium leak in FBR system cause damage as a result of thermo-chemical attack by burning sodium. In addition, release of free and bound water from concrete leads to generation of hydrogen gas, which is explosive in nature. Hence limestone concrete, as sacrificial layer on the structural concrete in FBR, needs to be qualified. Four concrete blocks of dimension 600 mm x 600 mm x 300 mm with 300 mm x 300 mm x 150 mm cavity were cast and subjected to controlled sodium exposure tests. They have composition of ordinary portland cement, water, fine and coarse aggregate of limestone in the ratio of 1: 0.58: 2.547: 3.817. These blocks were subjected to preliminary inspection by ultrasonic pulse velocity technique and rebound hammer tests. Each block was exposed for 30 minutes to about 12 kg of liquid sodium (∼ 120 mm liquid column) at 550 deg. C in open air, after which sodium was sucked back from the cavity of the concrete block into a sodium tank. On-line temperature monitoring was carried out at strategic locations of sodium pool and concrete block. After removing sodium from the cavity and cleaning the surfaces, rebound hammer testing was carried out on each concrete block at the same locations where data were taken earlier at pre-exposed stage. The statistical analysis of rebound hammer data revealed that one of the concrete block alone has undergone damage to the extent of 16%. The loss of mass occurred for all the four blocks varied from 0.6 to 2.4% due to release of water during the test duration. Chemical analysis of sodium in concrete samples collected from cavity floor of each block helped in generation of depth profiles of sodium monoxide concentration for each block. From this it is concluded that a bulk penetration of sodium up to 30 mm depth has taken place. However it was also observed that at few local spots, sodium penetrated into concrete up to 50 mm. Cylindrical core samples of 50 mm x 150

Instrumentation and Control Systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA)

Energy Technology Data Exchange (ETDEWEB)

Kim, Byeong Yeon; Kim, Hyung Mo; Cho, Youn Gil; Kim, Jong Man; Ko, Yung Joo; Kang, Byeong Su; Jung, Min Hwan; Jeong, Ji Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

A forced-draft sodium-to-air heat exchanger (FHX) is a part of decay heat removal system (DHRS) in Prototype Gen-IV Sodium-cooled fast reactor (PGSFR), which is being developed at Korea Atomic Energy Research Institute (KAERI). Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA) is a test facility for verification and validation of the design code for a forced-draft sodium-to-air heat exchanger (FHX). In this paper, we have provided design and fabrication features for the instrumentation and control systems of SELFA. In general, the instrumentation systems and control systems are coupled for measurement and control of process variables. Instrumentation systems have been designed for investigating thermal-hydraulic characteristics of FHX and control systems have been designed to control the main components (e.g. electromagnetic pumps, heaters, valves etc.) required for test in SELFA. In this paper, we have provided configurations of instrumentation and control systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA). The instrumentation and control systems of SELFA have been implemented based on the expected operation ranges and lesson learned from operational experience of 'Sodium integral effect test loop for safety simulation and assessment-1' (STELLA-1)

Methods to Compose Sodium Fire Extinguishing Equipment on Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, B H; Kim, J M; Jeong, J Y; Choi, B H

2008-06-15

Sodium fire is graded 'D' and it is difficult to extinguish sodium fire. In this report, the characteristics of sodium fire and the methods composing the suitable fire extinguishing systems to suppress fire effectively were described.

Methods to Compose Sodium Fire Extinguishing Equipment on Sodium Test Facility

International Nuclear Information System (INIS)

Kim, B. H.; Kim, J. M.; Jeong, J. Y.; Choi, B. H.

2008-06-01

Sodium fire is graded 'D' and it is difficult to extinguish sodium fire. In this report, the characteristics of sodium fire and the methods composing the suitable fire extinguishing systems to suppress fire effectively were described

Cleaning of small components of complex geometry by means of the sodium-alcohol reaction

International Nuclear Information System (INIS)

De Luca, B.; Grasso, C.; Spadoni, M.

1978-01-01

The results of some experiments on the vacuum reaction between butylcellosolve and sodium, contained in small diameter capillaries, are reported. The effects on the cleaning rate of the temperature, amount of solvent, diameter and position of the capillaries are analyzed. The facility, used for the cleaning of small components of complex geometry, is described. (author)

Fast flux test facility primary sodium check valve

International Nuclear Information System (INIS)

Rabe, G.B.; Nash, C.F.

1975-01-01

The design and development of a tilting-disc check valve for the primary sodium coolant loop of the Fast Flux Test Facility is described. The demanding design requirements specified for this system dictated a design with unique features. These features, along with the structural design and analysis requirements and the testing program used to develop and justify the design, are described

Analogy of water as compared to sodium in cavitation

International Nuclear Information System (INIS)

Bisci, R.; Courbiere, P.

1976-01-01

After outlining the major aspects of the cavitation research and test program undertaken by the CEA(3) (survey of the parameters defining the onset of cavitation in water and sodium flow, and the consequences of operation under sustained cavitating conditions on sodium reactor structural components), this paper describes the test and measuring equipment that has been developed for such studies. The results of the initial tests in water and in sodium using a thin 20 mm dia.orifices plate, are then presented. Except for uncertainties about the measurements themselves, the cavitation threshold values in cold water and in sodium at temperatures above 400 0 C has shown rather good concordance. Testing is currently in progress to confirm these findings. (1) EEC engineer assigned to DRNR/SEDC, (2) CEA engineer (DRNR/STRS), (3) French Atomic Energy Commision. (author)

United States position paper on sodium fires, design and testing

International Nuclear Information System (INIS)

Hilliard, R.K.; Johnson, R.P.

1983-01-01

The first Specialists' Meeting on sodium fire technology sponsored by the International Working Group on Fast Reactors (IWGFR) was held in Richland, Washington in 1972. The group concluded that the state-of-technology at that time was inadequate to support the growing LMFBR industry. During the second IWGFR Specialists' Meeting on sodium fires, held in Cadarache, France in 1978, a large quantity of technical information was exchanged and areas were identified where additional work was needed. Advances in several important areas of sodium fire technology have been made in the United States since that time, including improved computer codes, design of a sodium fire protection system for the CRBRP, measurement of water release from heated concrete, and testing and modeling of the sodium-concrete reaction. Research in the U.S. related to sodium fire technology is performed chiefly at the Energy Systems Group of Rockwell International (including Atomics International), the Hanford Engineering Development Laboratory (HEDL), and the Sandia National Laboratories (SNL). The work at the first two laboratories is sponsored by the U.S. Department of Energy, while that at the latter is sponsored by the U.S. Nuclear Regulatory Commission. Various aspects of sodium fire related work is also performed at several other laboratories. The current status of sodium fire technology in the U.S. is summarized in this report

Experimental Facilities for Performance Evaluation of Fast Reactor Components

International Nuclear Information System (INIS)

Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

2013-01-01

Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Foust, O J [ed.

1978-01-01

The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK components and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.

Conceptual Design of Liquid Sodium Charging and Draining System of STELLA-2

Energy Technology Data Exchange (ETDEWEB)

Cho, Chungho; Nam, Ho-Yun; Kim, Jong-Man; Ko, Yung Joo; Kim, Byeongyeon; Cho, Youngil; Eoh, Jeahyuk; Yoon, Jung; Kim, Hyungmo; Lee, Hyeong-Yeon; Lee, Yong Bum; Jeong, Ji-Young; Kim, Jong-Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

STELLA program consist of two phases. In the first phase of the program, separated effects tests for demonstrating the thermal-hydraulic performances of major components such as a decay heat exchanger (DHX), Natural-draft sodium-to-air heat exchanger (AHX) of the decay heat removal system, and mechanical sodium pump of the primary heat transport system (PHTS) had been successfully performed using STELLA-1 in 2015. Now is the time for us to proceed to the second phase of STELLA program. In the second phase, to demonstrate thermal-hydraulic performances and safety features and to produce base data for the specific design approval for the PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) which will be constructed by 2028 in Korea, integral effects tests will be carried out using the STELLA-2 test facility which is underway with detailed design and will be constructed in 2018. To establish the maintenance easiness of test facility and to prevent a de-functionalization of elementary sodium component located under the main experimental sodium loops, arrangement area was separated the sodium charging and draining piping area from main system experimental piping area. The operation margin of test facility and the best use of electro-magnetic pumps were established according to employ connecting line between RV charging piping and main experimental system charging piping. To prevent liquid sodium inflow into the piping of gas supply system, the overflow line was employed in sodium expansion tanks of PHDRS and ADHRS, RV, and reservoirs of IHTS.

Conceptual Design of Liquid Sodium Charging and Draining System of STELLA-2

International Nuclear Information System (INIS)

Cho, Chungho; Nam, Ho-Yun; Kim, Jong-Man; Ko, Yung Joo; Kim, Byeongyeon; Cho, Youngil; Eoh, Jeahyuk; Yoon, Jung; Kim, Hyungmo; Lee, Hyeong-Yeon; Lee, Yong Bum; Jeong, Ji-Young; Kim, Jong-Bum

2016-01-01

STELLA program consist of two phases. In the first phase of the program, separated effects tests for demonstrating the thermal-hydraulic performances of major components such as a decay heat exchanger (DHX), Natural-draft sodium-to-air heat exchanger (AHX) of the decay heat removal system, and mechanical sodium pump of the primary heat transport system (PHTS) had been successfully performed using STELLA-1 in 2015. Now is the time for us to proceed to the second phase of STELLA program. In the second phase, to demonstrate thermal-hydraulic performances and safety features and to produce base data for the specific design approval for the PGSFR (Prototype Generation IV Sodium-cooled Fast Reactor) which will be constructed by 2028 in Korea, integral effects tests will be carried out using the STELLA-2 test facility which is underway with detailed design and will be constructed in 2018. To establish the maintenance easiness of test facility and to prevent a de-functionalization of elementary sodium component located under the main experimental sodium loops, arrangement area was separated the sodium charging and draining piping area from main system experimental piping area. The operation margin of test facility and the best use of electro-magnetic pumps were established according to employ connecting line between RV charging piping and main experimental system charging piping. To prevent liquid sodium inflow into the piping of gas supply system, the overflow line was employed in sodium expansion tanks of PHDRS and ADHRS, RV, and reservoirs of IHTS

Simulation of creep test on 316FR stainless steel in sodium environment at 550degC

International Nuclear Information System (INIS)

Satmoko, A.; Asayama, Tai

1999-04-01

In sodium environment, material 316FR stainless steel risks to suffer from carburization. In this study, an analysis using a Fortran program is conducted to evaluate the carbon influence on the creep behavior of 316FR based on experimental results from uni-axial creep test that had been performed at temperature 550degC in sodium environment simulating Fast Breeder Reactor condition. As performed in experiments, two parts are distinguished. At first, elastic-plastic behavior is used to simulate the fact that just before the beginning of creep test, specimen suffers from load or stress much higher than initial yield stress. In second part, creep condition occurs in which the applied load is kept constant. The plastic component should be included, since stresses increase due to section area reduction. For this reason, elastic-plastic-creep behavior is considered. Through time carbon penetration occurs and its concentration is evaluated empirically. This carburization phenomena are assumed to affect in increasing yield stress, decreasing creep strain rate, and increasing creep rupture strength of material. The model is capable of simulating creep test in sodium environment. Material near from surface risks to be carburized. Its material properties change leading to non-uniform distribution of stresses. Those layers of material suffer from stress concentration, and are subject to damage. By introducing a damage criteria, crack initialization can thus be predicted. And even, crack growth can be evaluated. For high stress levels, tensile strength criterion is more important than creep damage criterion. But in low stress levels, the latter gives more influence in fracture. Under high stress, time to rupture of a specimen in sodium environment is shorter than in air. But for stresses lower than 26 kgf/mm 2 , the time to rupture of creep in sodium environment is the same or little longer than in air. Quantitatively, the carburization effect at 550degC is not important. This

Unites States position paper on sodium fires. Design basis and testing

International Nuclear Information System (INIS)

Lancet, R.T.; Johnson, R.P.; Matlin, E.; Vaughan, E.U.; Fields, D.E.; Glueckler, E.; McCormack, J.D.; Miller, C.W.; Pedersen, D.R.

1989-01-01

This paper focuses on designs, analyses, and tests performed since the last Sodium Fires Meeting of the IAEA International Working Group on Fast Reactors in May 1982. Since the U.S. Liquid Metal Reactor (LMR) program is focused on the two advanced LMRs, SAFR and PRISM, the paper relates this work to these designs. First, the design philosophy and approach taken by these advanced pool reactors are described. This includes methods of leak detection, the design basis leaks, and passive accommodation of sodium fires. Then the small- and large-scale sodium fire tests performed in support of the Clinch River Breeder Reactor Plant (CRBRP) program, including post-accident cleanup, are presented and related to the advanced LMR designs. Next, the assessment and behavior of the aerosols generated are discussed including generation rate, behavior within structures, release and dispersal, and deposition on safety-grade equipment. Finally, the impact of these aerosols on the performance of safety-grade decay heat removal heat exchange surfaces is discussed including some test results as well as planned tests. (author)

CNEN experience on bellows for sodium system

International Nuclear Information System (INIS)

Gai, A.; Casalini, P.

1980-01-01

In Italy work on the breeder reactors is being carried out in three main areas: The PEC reactor; Construction of the components for the Creys-Malville plant, the testing in support of this, and work on the SPX reactor; Standardized industrial supplies. With regard to the PEC reactor, design work for the system and a part of the components is being conducted mainly in Italy by CNEN and NIRA, with the experimental testing in support being conducted in Italy and abroad. With regard to the Creys-Malville plant and the SPX reactor programs, activities for the supply of the components and testing in support of them are being conducted within a tight framework of agreements between the research organizations and Italian and French industries. Finally, the standardized industrial supplies are developing autonomously, without any major research and development involvement. The differences in methodologies, procedures and organizations involved in the three above mentioned work areas and the relative limitation of activities conducted in the breeder reactor area in Italy have made the carrying out of a systematic, organic program for the use of sodium bellows impossible. In particular, until today no sodium testing program has been carried out on this component in Italy

CNEN experience on bellows for sodium system

Energy Technology Data Exchange (ETDEWEB)

Gai, A; Casalini, P

1980-02-01

In Italy work on the breeder reactors is being carried out in three main areas: The PEC reactor; Construction of the components for the Creys-Malville plant, the testing in support of this, and work on the SPX reactor; Standardized industrial supplies. With regard to the PEC reactor, design work for the system and a part of the components is being conducted mainly in Italy by CNEN and NIRA, with the experimental testing in support being conducted in Italy and abroad. With regard to the Creys-Malville plant and the SPX reactor programs, activities for the supply of the components and testing in support of them are being conducted within a tight framework of agreements between the research organizations and Italian and French industries. Finally, the standardized industrial supplies are developing autonomously, without any major research and development involvement. The differences in methodologies, procedures and organizations involved in the three above mentioned work areas and the relative limitation of activities conducted in the breeder reactor area in Italy have made the carrying out of a systematic, organic program for the use of sodium bellows impossible. In particular, until today no sodium testing program has been carried out on this component in Italy.

Extinguishing experiments of sodium fires carried out by TNO

International Nuclear Information System (INIS)

Meijer, G.J.A.M.; Rulkens, W.H.

1979-01-01

For the collection of burning sodium from the components and pipes of the secondary circuit of SNR 300, spill-trays are foreseen which are connected to dump tanks. These spill-trays are covered with a sieve in order to reduce the flow of air to the sodium in the spill-tray and hence to reduce the burning rate. In order to further minimize the consequences of a large sodium fire for the components, the licensing authority required as a back-up the installation of a remotely operated distribution system by means of which an extinguishing powder can be sprayed upon the spill-trays. Experiments were carried out in which the effectiveness of different extinguishing powders in combination with the sieve covered spill-trays were tested in a comparative manner. Attention was paid to the question whether such a spray system would have also additional benefits in the case of smaller sodium leaks. To this purpose three commercially available extinguishing powders were tested, one on a sodium chloride, two on a carbonate base. Also the effectiveness of the sieves proper with respect to reducing the burning rate was tested without applying any extinguishing powders. Finally for a reference some tests were done on open spill-trays, i.e. spill-trays not covered with a sieve.The investigations which were carried out in 1976-1977 were limited to fire experiments, aspects of transport of the powder in the distribution system were not investigated

Phase 2 THOR Steam Reforming Tests for Sodium Bearing Waste Treatment

Energy Technology Data Exchange (ETDEWEB)

Nicholas R. Soelberg

2004-01-01

About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste is stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Steam reforming is a candidate technology being investigated for converting the waste into a road ready waste form that can be shipped to the Waste Isolation Pilot Plant in New Mexico for interment. A steam reforming technology patented by Studsvik, Inc., and licensed to THOR Treatment Technologies has been tested in two phases using a Department of Energy-owned fluidized bed test system located at the Science Applications International Corporation (SAIC) Science and Technology Applications Research Center located in Idaho Falls, Idaho. The Phase 1 tests were reported earlier in 2003. The Phase 2 tests are reported here. For Phase 2, the process feed rate, stoichiometry, and chemistry were varied to identify and demonstrate process operation and product characteristics under different operating conditions. Two test series were performed. During the first series, the process chemistry was designed to produce a sodium carbonate product. The second series was designed to produce a more leach-resistant, mineralized sodium aluminosilicate product. The tests also demonstrated the performance of a MACT-compliant off-gas system.

Operating experience in cleaning sodium-wetted components at the KNK nuclear power plant

International Nuclear Information System (INIS)

Stade, K.Ch.

1978-01-01

Since 1969, components of the KNK facility, the first sodium cooled nuclear power plant in the Federal Republic of Germany, have been cleaned both by the alcohol and the wet gas techniques. This paper outlines the experience accumulated In the application of these methods, especially in cleaning steam generators and fuel elements. Some preliminary results are indicated of the attempt to clean a cold trap from the primary circuit of the KNK facility. (author)

Description of test facilities bound to the research on sodium aerosols - some significant results

Energy Technology Data Exchange (ETDEWEB)

Dolias, M; Lafon, A; Vidard, M; Schaller, K H [DRNR/STRS - Centre de Cadarache, Saint-Paul-lez-Durance (France)

1977-01-01

This communication is dedicated to the description of the CEA (French Atomic Energy Authority) testing located at CADARACHE and which are utilized for the study of sodium aerosols behavior. These testing loops are necessary for studying the operating of equipment such as filters, sodium vapour traps, condensers and separators. It is also possible to study the effect of characteristics parameters on formation, coagulation and carrying away of sodium aerosols in the cover gas. Sodium aerosols deposits in a vertical annular space configuration with a cold area in its upper part are also studied. Some significant results emphasize the importance of operating conditions on the formation of aerosols. (author)

Sodium aerosol behavior in liquid-metal fast breeder reactor containments

International Nuclear Information System (INIS)

Jordan, S.; Cherdron, W.; Malet, J.C.; Rzekiecki, R.; Himeno, Y.

1988-01-01

A tripartite consortium DEBENE (Deutschland-Belgium-Netherlands), Japan, and France studied the sodium evaporation process of aerosols in a sodium fire. In an inert atmosphere, experimental and theoretical condensation rates were compared and indicated sodium hydride (NaH) to be the foreign nucleus for mist formation. In a normal atmosphere, the physicochemical characteristics of the aerosols produced by a sodium fire and their evolution in containment or in the environment were determined; models enabling the various countries to achieve harmonious results were derived. The proper functioning of the components, guaranteeing perfect operation during and after a sodium fire accident, was tested

Distributed temperature sensor testing in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Gerardi, Craig, E-mail: cgerardi@anl.gov; Bremer, Nathan; Lisowski, Darius; Lomperski, Stephen

2017-02-15

Highlights: • Distributed temperature sensors measured high-resolution liquid-sodium temperatures. • DTSs worked well up to 400 °C. • A single DTS simultaneously detected sodium level and temperature. - Abstract: Rayleigh-backscatter-based distributed fiber optic sensors were immersed in sodium to obtain high-resolution liquid-sodium temperature measurements. Distributed temperature sensors (DTSs) functioned well up to 400 °C in a liquid sodium environment. The DTSs measured sodium column temperature and the temperature of a complex geometrical pattern that leveraged the flexibility of fiber optics. A single Ø 360 μm OD sensor registered dozens of temperatures along a length of over one meter at 100 Hz. We also demonstrated the capability to use a single DTS to simultaneously detect thermal interfaces (e.g. sodium level) and measure temperature.

Performance Evaluation in Sodium-to-Sodium Heat Exchangers in STELLA-2

International Nuclear Information System (INIS)

Jo, Youngchul; Son, Seok-kwon; Yoon, Jung; Jeong, Jiyoung

2016-01-01

The program aiming at an integral effect test is called STELLA-2, which will be used for synthetic review of the key safety issues of PGSFR. The basic and detailed design phases of the STELLA-2 test facility are underway in accordance with the specific design requirements reflecting the whole design features of PGSFR. Based on the STELLA-2 platform, a simulation of the PGSFR transient will be made to evaluate the plant dynamic behaviors and demonstrate the decay heat removal performance. The multi-dimensional effects coming from a large sodium pool system will be identified as well. Among several components of STELLA-2, there are five different types of model heat exchangers such as IHX, DHX, FHX, AHX, and UHX. Each heat exchanger has different characteristics, and it is very important to verify the heat transfer and pressure drop performance in each heat exchanger. The performance evaluation of the sodium-to-sodium heat exchangers (IHX and DHX) in STELLA-2 is performed using CFD. Also, these results are compared with 1-D heat exchanger design code. The shell/tube outlet temperature and heat transfer rate of the heat exchanger obtained by the CFD is not significantly

Annual report 1974. Sodium technology development programme

International Nuclear Information System (INIS)

1975-01-01

The sodium technology development program comprises a number of separate research programs in the field of designing and testing parts and components for the SNR-300 reactor. Design studies and theoretical studies on cold trapping and the behavior of hydrogen in sodium circuits are reported. A preliminary test program for fighting sodium fires is completed. Results of research done on vibration measurements and counter-current mixing in a dummy tube bundle of a S.N.R. spiralized steam generator with counter-current flow are reported briefly. Research done in the field of heat transfer, pressure drop and bubble dynamics of a straight pipe steam generator are also briefly reported. To determine the influence of spiral diameter of the spiralized pipe on heat transfer in a spiralized pipe heat exchanger, a second testsection will be built in 1975. Research was reported on pump viscoseals, bearing stability, rotordynamics and bearing materials for sodium pumps. Research done on the properties of SNR-construction materials at high temperature and long time exposure and corrosion in sodium are reported. Fundamental research on corrosion accompanied this research. The report closes with results of weldability, mechanized-welding and remote welding of sodium-wetted surfaces

Fast Flux Test Facility (FFTF) feedback reactivity components

International Nuclear Information System (INIS)

Nguyen, D.H.

1988-04-01

The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

Changes in the flexural strength of engineering ceramics after high temperature sodium corrosion test. Influence after sodium exposure for 1000 hours

International Nuclear Information System (INIS)

Hayashi, Kazunori; Tachi, Yoshiaki; Kano, Shigeki; Hirakawa, Yasushi; Komine, Ryuji; Yoshida, Eiichi

1998-02-01

Engineering ceramics have excellent properties such as high strength, high hardness and high heat resistance compared with metallic materials. To apply the ceramic in fast reactor environment, it is necessary to evaluate the sodium compatibility and the influence of sodium on the mechanical properties of ceramics. In this study, the influence of high temperature sodium on the mechanical properties of sintered ceramics of conventional and high purity Al 2 O 3 , SiC, SiAlON, AlN and unidirectional solidified ceramics of Al 2 O 3 /YAG eutectic composite were investigated by means of flexure tests. Test specimens were exposed in liquid sodium at 823K and 923K for 3.6Ms. There were no changes in the flexural strength of the conventional and high purity Al 2 O 3 , AlN and Al 2 O 3 /YAG eutectic composite after the sodium exposure at 823K. On the contrary, the decrease in the flexural strength was observed in SiC and SiAlON. After the sodium exposure at 923K, there were also no changes in the flexural strength of AlN and Al 2 O 3 /YAG eutectic composite. In the conventional and high purity Al 2 O 3 and SiC, the flexural strength decreased and signs of grain boundary corrosion were detected by surface observation. The flexural strength of SiAlON after the sodium exposure at 923K increased instead of severe corrosion. In the specimens those showed no changes in the flexural strength, further exposure in sodium is needed to verify whether the mechanical properties degrade or not. For SiAlON, it is necessary to clarify the reason for the increased strength after the sodium exposure at 923K. (author)

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

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

Intermediate-scale tests of sodium interactions with calcite and dolomite aggregate concretes

International Nuclear Information System (INIS)

Randich, E.; Acton, R.U.

1983-09-01

Two intermediate-scale tests were performed to compare the behavior of calcite and dolomite aggregate concretes when attacked by molten sodium. The tests were performed as part of an interlaboratory comparison between Sandia National Laboratories and Hanford Engineering Development Laboratories. Results of the tests at Sandia National Laboratories are reported here. The results show that both concretes exhibit similar exothermic reactions with molten sodium. The large difference in reaction vigor suggested by thermodynamic considerations of CO 2 release from calcite and dolomite was not realized. Penetration rates of 1.4 to 1.7 mm/min were observed for short periods of time with reaction zone temperatures in excess of 800 0 C during the energetic attack. The penetration was not uniform over the entire sodium-concrete contact area. Rapid attack may be localized due to inhomogeneities in the concrete. The chemical reaction zone is less then one cm thick for the calcite concrete but is about seven cm thick for the dolomite concrete

Sodium technology handbook

International Nuclear Information System (INIS)

2005-09-01

This document was published as a textbook for the education and training of personnel working for operations and maintenances of sodium facilities including FBR plants and those engaged in R and D activities related to sodium technology. This handbook covers the following technical areas. Properties of sodium. Compatibilities of sodium with materials. Thermalhydraulics and structural integrity. Sodium systems and components. Sodium instrumentations. Sodium handling technology. Sodium related accident evaluation and countermeasures for FBRs. Operation, maintenance and repair technology of sodium facilities. Safety measures related to sodium. Laws, regulations and internal rules related to sodium. The plannings and discussions of the handbook were made in the Sodium Technology Education Committee organized in O-arai Engineering Center consisting of the representatives of the related departments including Tsuruga headquarters. Experts in various departments participated in writing individual technical subjects. (author)

Distributed temperature sensor testing in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Gerardi, Craig; Bremer, Nathan; Lisowski, Darius; Lomperski, Stephen

2017-02-01

Rayleigh-backscatter-based distributed fiber optic sensors were immersed in sodium to obtain high-resolution liquid-sodium temperature measurements. Distributed temperature sensors (DTSs) functioned well up to 400°C in a liquid sodium environment. The DTSs measured sodium column temperature and the temperature of a complex geometrical pattern that leveraged the flexibility of fiber optics. A single Ø 360 lm OD sensor registered dozens of temperatures along a length of over one meter at 100 Hz. We also demonstrated the capability to use a single DTS to simultaneously detect thermal interfaces (e.g. sodium level) and measure temperature.

Preliminary Plugging tests in Narrow Sodium Channels by Sodium and Carbon Dioxide reaction

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

This report is on the investigation of the physical/chemical phenomena that a slow loss of CO{sub 2} inventory into sodium after the sodium-CO{sub 2} boundary failure in PCHEs in realistic operating conditions. The first phenomenon is potential channel plugging inside the narrow PCHE channel. Unlike a conventional shell and- tube type HXs, failures in a PCHE are expected to be small cracks. If the faulted channel is blocked, it may have a positive function for plant safety because the pressure boundary would automatically recover due to this self-plugging. The other one is damage propagation on pressure boundary, which is referred to as potential wastage with combined corrosion/erosion effect. Physical/chemical phenomena that a slow loss of CO{sub 2} inventory into sodium after the sodium-CO{sub 2} boundary failure in printed circuit heat exchangers (PCHEs) were investigated. Our preliminary experimental results of plugging show that sodium flow immediately stopped as CO{sub 2} was injected through the nozzle at 300-400 .deg. C in 3 mm sodium channels, whereas sodium flow stopped about 60 min after CO{sub 2} injection in 5 mm sodium channels.

Wireless sensor network for sodium leak detection

International Nuclear Information System (INIS)

Satya Murty, S.A.V.; Raj, Baldev; Sivalingam, Krishna M.; Ebenezer, Jemimah; Chandran, T.; Shanmugavel, M.; Rajan, K.K.

2012-01-01

Highlights: ► Early detection of sodium leak is mandatory in any reactor handling liquid sodium. ► Wireless sensor networking technology has been introduced for detecting sodium leak. ► We designed and developed a wireless sensor node in-house. ► We deployed a pilot wireless sensor network for handling nine sodium leak signals. - Abstract: To study the mechanical properties of Prototype Fast Breeder Reactor component materials under the influence of sodium, the IN Sodium Test (INSOT) facility has been erected and commissioned at Indira Gandhi Centre for Atomic Research. Sodium reacts violently with air/moisture leading to fire. Hence early detection of sodium leak if any is mandatory for such plants and almost 140 sodium leak detectors are placed throughout the loop. All these detectors are wired to the control room for data collection and monitoring. To reduce the cost, space and maintenance that are involved in cabling, the wireless sensor networking technology has been introduced in the sodium leak detection system of INSOT. This paper describes about the deployment details of the pilot wireless sensor network and the measures taken for the successful deployment.

Core Seismic Tests for a Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Lee, J. H

2007-01-15

This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

Contribution to perfecting eddy current testing of steam generator tubes of sodium cooled breeders: description of the Monacault loop for the study of sodium deposit influence

International Nuclear Information System (INIS)

Lapicore, A.; Lemarquis, J.C.; Oberlin, C.; Pigeon, M.

1981-12-01

In the event of sodium-water reaction in the steam generator of a sodium cooled breeder reactor, it is essential to be able to monitor the local loss of thickness of the tubes located in the reaction area. A method for monitoring the tubes by an eddy current probe is being developed for Super Phenix. The sodium deposits on the outer wall of the tubes, as well as their prolonged contact with high temperature sodium are likely to bring about a change in the signals picked up. A test loop, Monacault, has been built in order to clarify the importance of these parameters (effect of sodium deposits, reproducibility of the wetting at different temperatures). It includes three test cells containing the sample tubes having a total of 61 standard defects to be tested. The first results on the wetting of tubes are given and discussed [fr

Dismantling of the 50 MW steam generator test facility

International Nuclear Information System (INIS)

Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

1997-01-01

We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

Shaking table testing of mechanical components

International Nuclear Information System (INIS)

Jurukovski, D.; Taskov, Lj.; Mamucevski, D.; Petrovski, D.

1995-01-01

Presented is the experience of the Institute of Earthquake Engineering and Engineering Seismology, Skopje, Republic of Macedonia in seismic qualification of mechanical components by shaking table testing. Technical data and characteristics for the three shaking tables available at the Institute are given. Also, for characteristic mechanical components tested at the Institute laboratories, basic data such as producer, testing investor, description of the component, testing regulation, testing equipment and final user of the results. (author)

Test results of sodium-water reaction testing in near prototypical LMR steam generator

International Nuclear Information System (INIS)

Boardman, C.E.; Hui, M.; Neely, H.H.

1990-01-01

An extensive test program has been performed in the United States to investigate the effects of large sodium-water reaction events in LMFBR steam generators. Tests were conducted in the Large Leak Test Rig (LLTR) located at the Energy Technology Engineering Center (ETEC). The program was divided into two phases, Series I and Series II, for the purpose of satisfying near-term and long-term needs. Series II was further subdivided into large and intermediate leak tests. This paper will emphasize the Series II intermediate leak tests and resulting conclusions for steam generator design and operation. 11 figs, 2 tabs

Some new fatigue tests in high temperature water and liquid sodium environment

International Nuclear Information System (INIS)

Hattori, Takahiro; Yamauchi, Takayoshi; Kanasaki, Hiroshi; Kondo, Yoshiyuki; Endo, Tadayoshi.

1987-01-01

To evaluate the fatigue strength of structural materials for PWR or FBR plants, fatigue test data must be obtained in an environment of simulated primary and secondary water for PWR or of high temperature liquid sodium for FBR. Generally, such tests make it necessary to prepare expensive facilities, so when large amount of fatigue data are required, it is necessary to rationalize and simplify the fatigue tests while maintaining high accuracy. At the Takasago Research Development Center, efforts to rationalize facilities and maintain accuracy in fatigue tests have been made by developing new test methods and improving conventional techniques. This paper introduces a new method of low cycle fatigue test in high temperature water, techniques for automatic measurement of crack initiation and propagation in high temperature water environment and a multiple type fatigue testing machine for high temperature liquid sodium. (author)

Test of fuel handling machine for Monju in sodium

International Nuclear Information System (INIS)

Ishii, Yoichiro; Masuda, Yoichi; Kataoka, Hajime

1980-01-01

Various types of fuel handling machines were studied, and under-the-plug method of fuel exchange and the fuel handling machine of single turning plug, fixed arm type were selected for the prototype reactor ''Monju'', because the turning plug is relatively small, and the rate of operation, safety, operational ability, maintainability and reliability required for the reactor are satisfied, moreover, the extrapolation to the demonstration reactor was considered. Attention must be paid to the points that the fuel handling machine is very long and invisible from outside, and the smooth operation and endurance in sodium are required for it. The full mock-up testing facility of single turning plug, fixed arm type was installed in 1974, and the full mock-up test has been carried out since 1975 in Oarai. Fuel exchange is carried out at about 6 months intervals in Monju, and about 20 to 30% of core and blanket fuels are exchanged for about one month period. The functions required for the fuel handling machine for Monju, the outline of the testing facility, the schedule of the testing, the items of testing and the results, and the matters to be specially written are described. The full mock-up test in sodium has been carried out for 5 years, and the functions and the endurance have been proved sufficiently. (Kako, I.)

Developing works to detect fatigue cracks (small sodium leak detector and acoustic emission

International Nuclear Information System (INIS)

Kikuchi, M.; Sakakibara, Y.; Nagata, T.

1980-01-01

Continuous monitoring of fatigue cracks was performed (using both sodium leak detector and AE measuring system) through the creep-fatigue test of 304 stainless steel long elbow as part of the test series to establish the structural reliability of the Prototype FBR primary heat transport piping system. The sodium leak detector was a system composed mainly of SID (Sodium Ionization Detector) and DPD (Deferential Pressure Detector), that was developed by HITACHI Ltd. under a contract with PNC. The AE system was Synthetic AE Measuring and Analyzing system that was developed at FBR Safety Section to measure and analyze AE at various piping component tests. The test was continued until a sodium leakage was detected by the contact-type sodium leak detector attached to the test assembly, after about 4 weeks operation under cyclic loading at 600 deg. C. The following conclusions were obtained: (1) The sodium leak detector, both SID and DPD, indicated sodium leakage clearly, some hours before the contact-type detector did, even under an environment of air that contains ordinary humidity (Leaked sodium was estimated to be less than 15 grams after completion of the test); (2) The AE method indicated location and seriousness of the fatigue cracks, apparently before the crack penetration occurred. (author)

Control of sodium vapor transport in annuli

International Nuclear Information System (INIS)

Meadows, G.E.; Bohringer, A.P.

1983-11-01

The method used to control sodium vapor transport in the annuli of various components at the Fast Flux Test Facility (FFTF) is a downward purge of the annuli with high purity argon. The purge rates for the FFTF were selected by calculating the gas velocity required to overcome thermal convection transport in the annuli. To evaluate the effectiveness of the gas purge, laboratory apparatus was fabricated which simulated selected annuli in the FFTF In-Vessel Handling Machine (IVHM) and the Instrument Tree (IT) annuli. Tests were conducted at temperatures similar to FFTF conditions. Gas purge rates ranged from zero to 130% of FFTF flow rates. Test results show the effectiveness of a high purity gas purge in decreasing the accumulation of sodium vapor deposits in an annulus. The presence of water vapor and oxygen in the purge gas increased the sodium deposition rate by a factor of three over other tests usig high purity argon. The presence of a vapor control collar used in the IT annulus was shown to be beneficial for controlling vapor transport into the upper region of the annulus

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.

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

these robots, three main configurations have been considered, depending on the adopted solution for robot component seclusion: Zone 1: Tight surrounding shell cooled by argon gas flow. Constraints: irradiation and 70 deg. C temperature, Zone 2: Tight surrounding shell (not cooled). Constraints: irradiation and 180 deg. C-200 deg. C temperature, Zone 3: No tight surrounding shell. Constraints: irradiation and 180 deg. C-200 deg. C temperature and immersion within liquid sodium. It appears that some technical solutions do exist for future in sodium carriers, using available trade components. Qualification tests will be necessary in order to confirm some specific components (such as polymers, greases, sensors, reducers, motors, bearings). For bearings, a specific test program has to be performed in order to check their capability for static and dynamic loadings during in sodium operation (without lubricant). Electrical motor development for 200 deg. C operation seems possible and R and D work is being done in 2013 with already available components. (authors)

Experimental evaluation of sodium to air heat exchanger performance

International Nuclear Information System (INIS)

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

2013-01-01

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

The Self-Calibration Test of flowmeter installed in STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) facility

Energy Technology Data Exchange (ETDEWEB)

Jung, Minhwan; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The objective of this study is to describe the procedure of the self-calibration test for the flowmeters and to analyze the result of the test. In this work, the test procedure of the self-calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility was described and the test result was analyzed. In regard to the long-term SFR development plan, a large-scale sodium thermal-hydraulic test project is being progressed by KAERI. This project is called STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment), and it is proceeding by adopting the QA (Quality Assurance) program. Due to the specificity of an experiment using sodium(Na) categorized as Class 3(pyrophoric material and water-prohibiting substance) by the Safety Control of Dangerous Substances Act, it is necessary to apply QA in consideration of the sodium experiment environment in certain parts. The one of them is about calibration of measuring instrument such as a flowmeter, thermocouple and pressure gauge. It is described in the QAP (Quality Assurance Procedures) of KAERI that calibration work should be conducted in accordance with self-calibration procedures in a special case where conventional calibration is not practicable. The calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility is the typical example. As a result of test, it was confirmed that the flowmeters meet the pass criterion. Therefore, it was concluded that the flowmeters maintain instrument capacity a year ago.

Comparison of cavitation tests on the SNR 300 prototype sodium pump, carried out using water at room temperature and liquid sodium at 5800C

International Nuclear Information System (INIS)

Fakkel, R.H.; Hoornweg, C.J.; Kamerling, B.; Ten Wolde, Tj.; Heslenfeld, M.W.; Mendte, W.K.; Bunjies, J.H.

1976-01-01

This paper gives results of tests carried out on a centrifugal pump in both a water and a sodium test facility. The pump tested is a prototype of the primary circulation pumps intended for a LMFBR 300 MWe nuclear power station (the SNR reactor) under construction at Kalkar, West Germany. The pump characteristics under various NPSH-conditions were investigated, and a comparison is made in the paper between the results of cavitation tests using water and liquid sodium. An attempt is made to account for differences in pump characteristics in both types of tests by referring to the physical properties of the liquids used (water and sodium). An attempt is also made to correlate the results obtained with the full-scale prototype to those obtained previously with a half-scale model of the impeller. The various test circuits used should be made identical, in order to avoid that differences in geometry should obscure essential features resulting from using different liquids. Yet, it showed that in this respect, the tests did not obey to this key-rule, reasons why the test results did not fully reveal the essential physical properties of either fluid under cavitating conditions. (author)

Component evaluation testing and analysis algorithms.

Energy Technology Data Exchange (ETDEWEB)

Hart, Darren M.; Merchant, Bion John

2011-10-01

The Ground-Based Monitoring R&E Component Evaluation project performs testing on the hardware components that make up Seismic and Infrasound monitoring systems. The majority of the testing is focused on the Digital Waveform Recorder (DWR), Seismic Sensor, and Infrasound Sensor. In order to guarantee consistency, traceability, and visibility into the results of the testing process, it is necessary to document the test and analysis procedures that are in place. Other reports document the testing procedures that are in place (Kromer, 2007). This document serves to provide a comprehensive overview of the analysis and the algorithms that are applied to the Component Evaluation testing. A brief summary of each test is included to provide the context for the analysis that is to be performed.

Specialists meeting on sodium removal and decontamination. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-08-01

This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes.

Specialists meeting on sodium removal and decontamination. Summary report

International Nuclear Information System (INIS)

1978-08-01

This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes

Large scale sodium-water reaction tests for Monju steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Hori, M.

1976-01-01

To demonstrate the safe design of the steam generator system of the prototype fast reactor Monju against the postulated large leak sodium-water reaction, a large scale test facility SWAT-3 was constructed. SWAT-3 is a 1/2.5 scale model of the Monju secondary loop on the basis of the iso-velocity modeling. Two tests have been conducted in SWAT-3 since its construction. The test items using SWAT-3 are discussed, and the description of the facility and the test results are presented

In-test and post-test analyses of sodium-sulfur cells

Energy Technology Data Exchange (ETDEWEB)

Wada, Motoi; Kawamoto, Hiroyuki; Hatoh, Hisamitsu

1986-01-15

Cell life of sodium-sulfur cells is often determined by the degradation of the solid electrolyte. Solid electrolyte degradation will cause an increase of electrolyte resistivity, decrease of faradic efficiency, or even an electrolyte rupture which leads to a cell temperature rise due to direct reaction of reactants. Electrolyte degradation in actual sodium-sulfur cells is believed to be caused by the passage of sodium ion current across the solid electrolyte. The degree of degradation has been reported to be a function of amount of charge passed through the electrolyte, and the breakdown of the solid electrolyte was observed to occur above some threshold. For this reason, the concentration of sodium ion current density is to be avoided to prevent solid electrolyte from premature degradation and rupture, and the electrode structure for a sodium-sulfur cell should be determined with enough care to homogenize the current density distribution on the electrolyte. The longitudinal current density distribution of a sodium-sulfur cell was measured by attaching probing terminals on the electrode container. It was found that the current density distribution of a vertically supported cell was inhomogeneous due to the effect of gravity. This setup can be used as a way to locate the place where the first electrolyte cracking occurs. It was also found that the electrolyte cracking accompanies a fluctuation of cycling cell voltage that starts to appear several cycles before the noticeable break down of the electrolyte.

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

Energy Technology Data Exchange (ETDEWEB)

Le Bourdais, F.; Le Polles, T. [Non Destructive Testing Department at the French Atomic Energy Commission (CEA), Saclay, 91191 Gif sur Yvette CEDEX, (France); Baque, F. [Department of Sodium Technology at the French Atomic Energy Commission (CEA), Cadarache, 13108 St Paul lez Durance CEDEX, (France)

2015-07-01

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

International Nuclear Information System (INIS)

Le Bourdais, F.; Le Polles, T.; Baque, F.

2015-01-01

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

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.)

Profile of irritant patch testing with detergents: sodium lauryl sulfate, sodium laureth sulfate and alkyl polyglucoside.

Science.gov (United States)

Löffler, H; Happle, R

2003-01-01

The cutaneous reaction to detergents follows distinct kinetic rules: the duration of application and the irritant concentration are of major importance. The aim of this study was to evaluate the differences in kinetics of skin reaction between the standard irritant sodium lauryl sulfate (SLS), and 2 modern detergents: sodium laureth sulfate (SLES) and alkyl polyglucoside (APG). We performed patch testing with SLS and SLES (or APG) at different concentrations (0.125, 0.25, 0.5, 1.0 and 2.0%) and with different exposure times (6, 12 and 24 h). Evaluation was conducted by measurement of transepidermal water loss (TEWL) and laser Doppler flowmetry (LD) 24 h, 7 and 10 days after patch removal. We found a pronounced reaction to SLS, and a far milder one to SLES. Even at the highest concentration the skin reaction to APG was hard to detect. During the regeneration period (day 3-10) SLS showed even at day 10 an increased TEWL at all concentrations tested. The irritation due to SLES was convincingly detectable only up to day 7, whereas the APG-tested skin areas showed no significant reaction even at day 3. These results demonstrate the improvement in reduction of skin irritation achieved by development of novel detergents.

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.)

Postaccident heat removal: large-scale molten-fuel-sodium interaction experiments

International Nuclear Information System (INIS)

Johnson, T.R.; Pavlik, J.R.; Baker, L. Jr.

1975-02-01

Kilogram-scale interactions between molten UO 2 and sodium were performed in an unrestrained geometry to study the resulting energetics and fragmentation. The molten UO 2 was producted by the exothrmic reaction between uranium and MoO 3 powders. Under the conditions of the experiments completed to date, the short-rise-time pressure pulses created in the liquid phase had negligible work potential, and their magnitude did not increase with the amount of molten fuel. No significant gas-phase shock pressures were generated. The largest potential for mechanical work was the sodium vapor generated over a period of roughly 1 sec. About 20 percent of the heat was effective in generating vapor. The ex- perimental results show a marked tendency of molten UO 2 to form particulate after passage through only a few inches of sodium. Particle size distributions obtained under the conditions of the experiments were not significantly different from those obtained in prior small-scale tests and in TREAT tests. Also, the results indicate that the metallic component of the molten mixture formed larger particles than the oxide component. (U.S.)

Sodium leakage and combustion tests. Measurement and distribution of droplet size using various spray nozzles

International Nuclear Information System (INIS)

Nagai, Keiichi; Hirabayashi, Masaru; Onojima, T.; Gunji, Minoru; Ara, Kuniaki; Oki, Yoshihisa

1999-04-01

In order to develop a numerical code simulating sodium fires initiated frame dispersion of droplets, measured data of droplet diameter as well as its distribution are needed. In the present experiment the distribution of droplet diameter was measured using water, oil and sodium. The tests elucidated the influential factors with respect to the droplet diameter. In addition, we sought to develop a similarity law between water and sodium. The droplet size distribution of sodium using the large diameter droplet (Elnozzle) was predicted. (J.P.N.)

Ultrasonic sweep arm for sodium cooled reactors

International Nuclear Information System (INIS)

Rohrbacher, H.A.; Bartholomay, R.

1975-05-01

This report describes experience in the use of a new type of monitoring and testing device to be applied in conjunction with components under sodium. In the method outlined, ultrasonic pulses are used which are emitted into the sodium plenum of fast breeder reactors by newly developed high temperature transducers. The basic work was conducted under out-of-pile conditions in a sodium tank of the sodium tank facility of the Karlsruhe Institute for Reactor Development. The sensor development, which preceded this phase, resulted in the use of soldered lithium niobate crystals whose operating characteristics were improved by the preliminary treatment outlined in the report. Special materials and techniques suitable for sensor fabrication are proposed. An alternative to soldering is suggested for contacting the crystals with their diaphragms, i.e. a contact pressure concept for the range of application up to 2 MHz. (orig.) [de

KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

International Nuclear Information System (INIS)

1978-01-01

The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

Acoustic analysis of sodium boiling stability tests using THORS bundle 6A

International Nuclear Information System (INIS)

Sheen, S.H.; Bobis, J.P.; Carey, W.M.

1977-01-01

Acoustic data from boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility are presented and discussed. The THORS sodium loop is a high temperature test facility that contains the bundle 6A, a full length stimulated fuel subassembly with nineteen electrically heated pins. Boiling stability tests on the THORS facility were designed to determine if a stable boiling region exists during the thermal hydraulic test at normal and off-normal conditions. Boiling was observed and the stable boiling region was determined. The acoustic data observed by three ANL sodium-immersible microphones have provided the following information: (1) the boiling signal is clearly observed and shows a correlation with the inlet flow fluctuations; (2) the signal level and the repetition rate of the boiling signal are directly related to the applied heat flux; (3) a typical boiling pulse consists of a high frequency signal due mainly to the bubble collapse and a low frequency (approximately 75 Hz) void oscillation; (4) a boiling pulse yields a frequency spectrum with significant amplitudes up to 80 KHz as compared with 4 KHz for background pulses; and (5) the frequency content of a boiling pulse can be mostly explained in terms of various resonance frequencies of the loop. The characterization of these data is pertinent to the design of sodium boiling detection systems

Compatibility and testing of electronic components

CERN Document Server

Jowett, C E

2013-01-01

Compatibility and Testing of Electronic Components outlines the concepts of component part life according to thresholds of failure; the advantages that result from identifying such thresholds; their identification; and the various tests used in their detection. The book covers topics such as the interconnection of miniature passive components; the integrated circuit compatibility and its components; the semiconductor joining techniques; and the thin film hybrid approach in integrated circuits. Also covered are topics such as thick film resistors, conductors, and insulators; thin inlays for el

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

Lubrication of nuclear reactor components

International Nuclear Information System (INIS)

Wild, E.; Mack, K.J.

1978-01-01

Safe operation of liquid metal cooled nuclear reactors requires a knowledge of the tribological behaviour of contacting components at high temperatures with slow relative movement at high frictional loads in a chemically aggressive environment. Experiments have been performed on various material combinations in liquid sodium and argon. Because of the small sliding movements, hydrodynamic lubrication is not expected and thus surface finish is an important factor. Tests have been performed on brushed, ground and lapped surfaces. Among the material combinations tested a CrC-coating on a 1.4961 stainless steel substrate performed well. Friction coefficients of 0.35-0.5 in argon and 0.1-1.2 in liquid sodium were recorded. (author)

Cast Stone Formulation At Higher Sodium Concentrations

International Nuclear Information System (INIS)

Fox, K. M.; Edwards, T. A.; Roberts, K. B.

2013-01-01

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium

Cast Stone Formulation At Higher Sodium Concentrations

International Nuclear Information System (INIS)

Fox, K. M.; Roberts, K. A.; Edwards, T. B.

2014-01-01

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium

Cast Stone Formulation At Higher Sodium Concentrations

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Roberts, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-02-28

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium

Cast Stone Formulation At Higher Sodium Concentrations

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M.; Edwards, T. A.; Roberts, K. B.

2013-10-02

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium

Cast Stone Formulation At Higher Sodium Concentrations

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M.; Roberts, K. A.; Edwards, T. B.

2013-09-17

A low temperature waste form known as Cast Stone is being considered to provide supplemental Low Activity Waste (LAW) immobilization capacity for the Hanford site. Formulation of Cast Stone at high sodium concentrations is of interest since a significant reduction in the necessary volume of Cast Stone and subsequent disposal costs could be achieved if an acceptable waste form can be produced with a high sodium molarity salt solution combined with a high water to premix (or dry blend) ratio. The objectives of this study were to evaluate the factors involved with increasing the sodium concentration in Cast Stone, including production and performance properties and the retention and release of specific components of interest. Three factors were identified for the experimental matrix: the concentration of sodium in the simulated salt solution, the water to premix ratio, and the blast furnace slag portion of the premix. The salt solution simulants used in this study were formulated to represent the overall average waste composition. The cement, blast furnace slag, and fly ash were sourced from a supplier in the Hanford area in order to be representative. The test mixes were prepared in the laboratory and fresh properties were measured. Fresh density increased with increasing sodium molarity and with decreasing water to premix ratio, as expected given the individual densities of these components. Rheology measurements showed that all of the test mixes produced very fluid slurries. The fresh density and rheology data are of potential value in designing a future Cast Stone production facility. Standing water and density gradient testing showed that settling is not of particular concern for the high sodium compositions studied. Heat of hydration measurements may provide some insight into the reactions that occur within the test mixes, which may in turn be related to the properties and performance of the waste form. These measurements showed that increased sodium

Development of LEAP-JET code for sodium-water reaction analysis. Validation by sodium-water reaction tests (SWAT-1R)

International Nuclear Information System (INIS)

Seino, Hiroshi; Hamada, Hirotsugu

2004-03-01

The sodium-water reaction event in an FBR steam generator (SG) has influence on the safety, economical efficiency, etc. of the plant, so that the selection of design base leak (DBL) of the SG is considered as one of the important matters. The clarification of the sodium-water reaction phenomenon and the development of an analysis model are necessary to estimate the sodium-water reaction event with high accuracy and rationality in selecting the DBL. The reaction jet model is pointed out as a part of the necessary improvements to evaluate the overheating tube rupture of large SGs, since the behavior of overheating tube rupture is largely affected by the reaction jet conditions outside the tube. Therefore, LEAP-JET has been developed as an analysis code for the simulation of sodium-water reactions. This document shows the validation of the LEAP-JET code by the Sodium-Water Reaction Test (SWAT-1R). The following results have been obtained: (1) The reaction rate constant, K, is estimated at between 0.001≤K≤0.1 from the LEAP-JET analysis of the SWAT-1R data. (2) The analytical results on the high-temperature region and the behaviors of reaction consumption (Na, H 2 O) and products (H 2 , NaOH, Na 2 O) are considered to be physically reasonable. (3) The LEAP-JET analysis shows the tendency of overestimation in the maximum temperature and temperature distribution of the reaction jet. (4) In the LEAP-JET analysis, the numerical calculation becomes unstably, especially in the mesh containing quite small sodium mass. Therefore, it is necessary to modify the computational algorism to stabilize it and obtain the optimum value of K in sodium-water reactions. (author)

Intelligent type sodium level gauge and its graduation calibration facility

International Nuclear Information System (INIS)

Chen Daolong; Wang Xuan; Li Xinying; Sun Huiqing; Zhu Jie

1998-04-01

The component construction and their performances of the intelligent type sodium level gauge newly-developed and its graduation calibration facility are presented. They can be operated in the temperature limit 100∼550 degree C. Its graduation characteristic calibration test is described. The temperature effect is analyzed. The graduation characteristic equation using the medium temperature as the parameter is given. The calibration errors are analyzed. The measurement system using this sodium level gauge is presented. The tests show that the intelligent type sodium level gauge possesses good linearity. The accurate sodium level measurement data can be obtained by means of its on-line compensation function of the temperature effect. Moreover, it possesses the self-inspection, the electric shutoff protection, the setting of full-scale, the thermocouple breaking alarm, the two upper limits and two lower limits alarms, the standard analog output signal and the digital output signal. Therefore, it is applicable particularly to the instrument, control and protection systems of LMFBR. The basic error of this intelligent type sodium level gauge is +-1.9% of measuring range

Sodium test of the Super-Phenix full size primary pump shaft on the CPV-1 test rig at ENEA-Brasimone

International Nuclear Information System (INIS)

Contardi, T.; Rapezzi, L.; Partiti, C.; Zola, M.; Denimal, P.

1984-01-01

Tests on FBR Superphenix primary pump shaft were performed within the sodium-cooled FBR common research and development programs provided for by the cooperation agreement between ENEA and CEA. These tests were performed in CPV-1 plant ENEA - Brasimone Energy Research Center. The CPV-1 rig was built by FIAT-TTG and reproduces the reactor operating conditions (sodium-temperature and level, shaft inclination, etc..). Furthermore, CPV-1 rig's most interesting feature is its possibility to apply seismic stresses to test section by means of an oleodynamic actuator. Pivoterie-1 test section was made by JEUMONT-SCHNEIDER which built Superphenix pumps too; it was given to ENEA by FIAT-TTG. Seismic tests were performed with the cooperation of ISMES and FIAT-TTG. (author)

Empirical usability testing in a component-based environment : improving test efficiency with component-specific usability measures

NARCIS (Netherlands)

Brinkman, W.P.; Haakma, R.; Bouwhuis, D.G.; Bastide, R.; Palanque, P.; Roth, J.

2005-01-01

This paper addresses the issue of usability testing in a component-based software engineering environment, specifically measuring the usability of different versions of a component in a more powerful manner than other, more holistic, usability methods. Three component-specific usability measures are

21 CFR 176.320 - Sodium nitrate-urea complex.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Sodium nitrate-urea complex. 176.320 Section 176... Substances for Use Only as Components of Paper and Paperboard § 176.320 Sodium nitrate-urea complex. Sodium nitrate-urea complex may be safely used as a component of articles intended for use in producing...

Metallurgical Laboratory and Components Testing

Data.gov (United States)

Federal Laboratory Consortium — In the field of metallurgy, TTC is equipped to run laboratory tests on track and rolling stock components and materials. The testing lab contains scanning-electron,...

Environmental effects of high temperature sodium of fatigue crack characteristics

International Nuclear Information System (INIS)

Abe, Hideaki; Takahashi, Kazuo; Ozawa, Kazumasa; Takahashi, Yukio

2004-01-01

In order to study fatigue crack growth characteristics in the components used in liquid sodium, fatigue tests were carried out at 550degC. This is near the system temperature used for sodium coolant in fast breeder reactors (FBRs). The factors influencing fatigue lifetime in sodium compared with that in air were investigated by observation of surface cracks in 316FR steel. Furthermore, the effects of sodium environment on fatigue were investigated based on examining the results of thermal striping tests, etc., obtained up to now. The results of the fatigue tests show that many micro cracks in the shearing direction were produced by the mid-lifetime, and micro cracks connected quickly after that. This is because an oxidation film was not formed, since sodium is of a reductive nature, and strain of the material surface tends to distribute equally. During crack progression there is no oxide formed on broken surfaces. Therefore re-combination between broken surfaces takes place, and crack progression rate falls. Furthermore, in non-propagating crack, the wedge effect by oxide between broken surfaces at the time of compression is small. Therefore, the crack closure angle is small, compression strain generated in the crack tip becomes large, and the crack cannot stop easily. As mentioned above, the main sodium influence on the fatigue characteristics are because of its reductive nature. In summary, in sodium environment, it is hard to form a crack and to get it to grow. Once started, however, it is hard to stop the crack in sodium compared with in the case of the air. (author)

Sodium nitrate combustion limit tests

International Nuclear Information System (INIS)

Beitel, G.A.

1976-04-01

Sodium nitrate is a powerful solid oxidant. Energetically, it is capable of exothermically oxidizing almost any organic material. Rate-controlling variables such as temperature, concentration of oxidant, concentration of fuel, thermal conductivity, moisture content, size, and pressure severely limit the possibility of a self-supported exothermic reaction (combustion). The tests reported in this document were conducted on one-gram samples at atmospheric pressure. Below 380 0 C, NaNO 3 was stable and did not support combustion. At moisture concentrations above 22 wt percent, exothermic reactions did not propagate in even the most energetic and reactive compositions. Fresh resin and paraffin were too volatile to enable a NaNO 2 -supported combustion process to propagate. Concentrations of NaNO 3 above 95 wt percent or below 35 wt percent did not react with enough energy release to support combustion. The influence of sample size and confining pressure, both important factors, was not investigated in this study

Experimental study of sodium droplet burning in free fall. Evaluation of preliminary test results

International Nuclear Information System (INIS)

Miyahara, Shinya; Ara, Kuniaki

1998-08-01

To study a sodium leak and combustion behavior phenomenologically and to construct the mechanistic evaluation method, an experimental series of a sodium droplet burning in free fall is under way. In this study, the accuracy of measurement technique used in the preliminary test was assessed and the modified technique was proposed for the next test series. Analytical study of the test results was also conducted to deduce dominant parameters and important measurement items which would play an important role in the droplet combustion behavior. The results and conclusions are as follows: (1) Assessment of measurement accuracy and modified technique proposed for the next test series. a) Control accuracy of sodium supply system using β-alumina solid electrolyte was sufficient for generation of objective size of single droplet. However, it is necessary to calibrate the correlation between the quantity of electric charge for sodium supply system and that of supplied sodium. b) Measurement accuracy of falling velocity using high-speed video was ±0.33 m/s at an upper part and ±0.48 m/s at a lower part of the measurement. To reduce the error, a high-speed stroboscopic method is recommended to measure the falling velocity of droplet. (2) Results of analytical study and deduced dominant parameters and important measurement items. a) The falling behavior of a burning droplet was described solving the equation of free falling motion for a rigid sphere. In the case of higher falling height, it is necessary to study the burning effects on the falling behavior. b) The mass burned of a falling droplet was calculated using the combustion model according to 'D 2 ' law during the full burning phase. It is necessary to study the dominant chemical reaction in the burning flame because the mass burned depends on the composition of the reaction products. c) The mass burned was calculated using surface oxidation model for preignition phase together with above model. However, it is

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

Liquid sodium technology research

International Nuclear Information System (INIS)

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

1982-01-01

This report describes the technology of impurity control and measurement of liquid sodium, problems associated with material degradation and change of heat transfer characteristics in liquid sodium, and the conceptual design of multipurpose sodium test loop. Discussion and the subsequent analysis are also made with regard to the test results for the sodium-H 2 0 reaction and its effects on the system. (author)

Measurement of the Residual Sodium and Reaction Compounds on a Cleaned Cold Trap

International Nuclear Information System (INIS)

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

2006-01-01

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. 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 a intergranular penetration characteristic of a high-oxygen sodium or an accelerated corrosion due to oxygen. Cleaning methods are needed that will avoid a deleterious local overheating, material surface degradation or deposits, a chemical, physical, or mechanical damage, and external effects. It is important to determine the levels of residual sodium that can be accepted so that those deleterious effects will not negate the reuse of the component. The purpose of this paper is to measure the amount of the sodium and the reaction compounds remaining on a component after a cleaning and prepare acceptable criteria for the reuse of components which have been subjected to a sodium cleaning

Numerical simulation of alteration of sodium bentonite by diffusion of ionic groundwater components

International Nuclear Information System (INIS)

Jacobsen, J.S.; Carnahan, C.L.

1987-12-01

Experiments measuring the movement of trace amounts of radionuclides through compacted bentonite have typically used unaltered bentonite. Models based on experiments such as these may not lead to accurate predictions of the migration through altered or partially altered bentonite of radionuclides that undergo ion exchange. To address this problem, we have modified an existing transport code to include ion exchange and aqueous complexation reactions. The code is thus able to simulate the diffusion of major ionic groundwater components through bentonite and reactions between the bentonite and groundwater. Numerical simulations have been made to investigate the conversion of sodium bentonite to calcium bentonite for a reference groundwater characteristic of deep granitic formations. 20 refs., 2 figs., 2 tabs

HG ion thruster component testing

Science.gov (United States)

Mantenieks, M. A.

1979-01-01

Cathodes, isolators, and vaporizers are critical components in determining the performance and lifetime of mercury ion thrusters. The results of life tests of several of these components are reported. A 30-cm thruster CIV test in a bell jar has successfully accumulated over 26,000 hours. The cathode has undergone 65 restarts during the life test without requiring any appreciable increases in starting power. Recently, all restarts have been achieved with only the 44 volt keeper supply with no change required in the starting power. Another ongoing 30-cm Hg thruster cathode test has successfully passed the 10,000 hour mark. A solid-insert, 8-cm thruster cathode has accumulated over 4,000 hours of thruster operation. All starts have been achieved without the use of a high voltage ignitor. The results of this test indicate that the solid impregnated insert is a viable neutralizer cathode for the 8-cm thruster.

Concrete protection from sodium spills by intentionally defected liners, small-scale tests S9 and S10

International Nuclear Information System (INIS)

Hilliard, R.K.; Boehmer, W.D.

1975-07-01

Two small scale tests were performed to determine the protection against sodium attack afforded to a concrete surface by a defected steel liner. An inert atmosphere was maintained over the sodium pool, which was heated electrically to 1600 0 F for 2--6 hrs in one test, to 1380 0 F for 19 hrs in the other. The 10 inch diameter vertical concrete surface was separated from the sodium by a liner plate in which small defects had been drilled. The plates provided significant protection against direct chemical attack, but most of the water was released from the concrete through the defects to react in the sodium pool region. The liners were corroded significantly in the defect areas. (U.S.)

Sodium-concrete reaction model development

International Nuclear Information System (INIS)

Nguyen, D.H.; Muhlestein, L.D.; Postma, A.K.

1982-07-01

Major observations have been formulated after reviewing test results for over 100 sodium-concrete reaction tests. The observations form the basis for developing a mechanistic model to predict the transient behavior of sodium-concrete reactions. The major observations are listed. Mechanisms associated with sodium and water transport to the reaction zone are identified, and represented by appropriate mathematical expressions. The model attempts to explain large-scale, long-term (100 h) test results were sodium-concrete reactions terminated even in the presence of unreacted sodium and concrete

Integrated dry NO{sub x}/SO{sub 2} emissions control system sodium-based dry sorbent injection test report. Test period: August 4, 1993--July 29, 1995

Energy Technology Data Exchange (ETDEWEB)

Smith, R.A.; Shimoto, G.H.; Muzio, L.J. [Fossil Energy Research Corp., Laguna Hills, CA (United States); Hunt, T. [Public Service Co. of Colorado, Denver, CO (United States)

1997-04-01

The project goal is to demonstrate up to 70% reductions in NOx and SO{sub 2} emissions through the integration of: (1) down-fired low-NOx burners with overfire air; (2) Selective Non-Catalytic Reduction (SNCR) for additional NOx removal; and (3) dry sorbent injection and duct humidification for SO{sub 2} removal. This report documents the sixth phase of the test program, where the performance of dry sorbent injection with sodium compounds was evaluated as a SO{sub 2} removal technique. Dry sorbent injection was performed in-duct downstream of the air heater (ahead of the fabric filter), as well as at a higher temperature location between the economizer and air heater. Two sodium compounds were evaluated during this phase of testing: sodium sesquicarbonate and sodium bicarbonate. In-duct sodium injection with low levels of humidification was also investigated. This sixth test phase was primarily focused on a parametric investigation of sorbent type and feed rate, although boiler load and sorbent preparation parameters were also varied.

Impact test of components

International Nuclear Information System (INIS)

Borsoi, L.; Buland, P.; Labbe, P.

1987-01-01

Stops with gaps are currently used to support components and piping: it is simple, low cost, efficient and permits free thermal expansion. In order to keep the nonlinear nature of stops, such design is often modeled by beam elements (for the component) and nonlinear springs (for the stops). This paper deals with the validity and the limits of these models through the comparison of computational and experimental results. The experimental results come from impact laboratory tests on a simplified mockup. (orig.)

Sodium pool combustion test for small-scale leakage. Run-F7-4 and Run-F8-2

International Nuclear Information System (INIS)

Futagami, Satoshi; Ohno, Shuji

2003-06-01

Since 1998, the test (Run-F7 series) was performed to acquire the fundamental knowledge about the sodium pool growth and floor liner temperature in the case of small-scale leakage of sodium. And the test (Run-F8 series) was performed to know the floor liner material corrosion mechanism under high moisture conditions. In both test series, those influences are investigated by making the rate of sodium leakage, and moisture conditions of supply air into main parameters. As the last test, (1) Run-F7-4 (June 28, 2000) and (2) Run-F8-2 (January 26, 2000) were carried out. The conclusion of the following which receives sodium small-scale leakage (about 10 kg/h) was obtained from these experiments and the result of old Run-F7 and Run-F8 series. The peak temperature of a catch pan tends to become lower with decrease of sodium leak rate. Moreover, height of leak point and moisture conditions also become the factor which raises the catch pan peak temperature. Although it grows up in proportion [almost]to time in early stages of leakage about growth of a sodium pool, growth stops during the leakage. Moreover, the final growth area is mostly proportional to the rate of sodium leakage. It was suggested by the measured value of catch pan corrosion thickness and a material analysis result that the dominant corrosion mechanism was relatively slow Na-Fe double oxidization type corrosion even under the high moisture condition of 4.6 to 4.8%. And the chemical analysis result of a deposits also suggested that the catch pan material was in the environment in which molten salt type corrosion was not easy to occur. (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)

Characterization of Sm14 related components in different helminths by sodium dodecyl sulphate-polyacrylamide gel electrophoresis and Western blotting analysis

Directory of Open Access Journals (Sweden)

Nilton Thaumaturgo

2002-10-01

Full Text Available Sm14 was the first fatty acid-binding protein homologue identified in helminths. Thereafter, members of the same family were identified in several helminth species, with high aminoacid sequence homology between them. In addition, immune crossprotection was also reported against Fasciola hepatica infection, in animals previously immunized with the Schistosoma mansoni vaccine candidate, r-Sm14. In the present study, data on preliminary sodium dodecyl sulphate-polyacrylamide gel electrophoresis and Western blotting analysis of nine different helminth extracts focusing the identification of Sm14 related proteins, is reported. Out of these, three extracts - Ascaris suum (males and females, Echinostoma paraensei, and Taenia saginata - presented components that comigrated with Sm14 in SDS-PAGE, and that were recognized by anti-rSm14 policlonal serum, in Western blotting tests.

Report of sodium cavitation

International Nuclear Information System (INIS)

Murai, Hitoshi; Shima, Akira; Oba, Toshisaburo; Kobayashi, Ryoji; Hashimoto, Hiroyuki

1975-01-01

The damage of components for LMFBRs due to sodium cavitation is serious problem. This report summarizes the following items, (1) mechanism of the incipience of sodium cavitation, (2) damage due to sodium cavitation, (3) detection method for sodium cavitation, and (4) estimation method for sodium cavitation by the comparison with water cavitation. Materials were collected from the reports on liquid metal cavitation, sodium cavitation and water cavitation published from 1965 to now. The mechanism of the incipience of sodium cavitation cavitation parameters (mean location, distributed amount or occurrence aspect and stability), experiment of causing cavitation with Venturi tube, and growth of bubbles within superheated sodium. The sodium cavitation damage was caused by magnetostriction vibration method and with Venturi tube. The state of damage was investigated with the cavitation performance of a sodium pump, and the damage was examined in view of the safety of LMFBR plants. Sodium cavitation was detected with acoustic method, radiation method, and electric method. The effect of physical property of liquid on incipient cavitation was studied. These are thermodynamic effect based on quasistatic thermal equilibrium condition and the effect of the physical property of liquid based on bubble dynamics. (Iwase, T.)

Standard test method for evaluating stress-corrosion cracking of stainless alloys with different nickel content in boiling acidified sodium chloride solution

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

1.1 This test method describes a procedure for conducting stress-corrosion cracking tests in an acidified boiling sodium chloride solution. This test method is performed in 25% (by mass ) sodium chloride acidified to pH 1.5 with phosphoric acid. This test method is concerned primarily with the test solution and glassware, although a specific style of U-bend test specimen is suggested. 1.2 This test method is designed to provide better correlation with chemical process industry experience for stainless steels than the more severe boiling magnesium chloride test of Practice G36. Some stainless steels which have provided satisfactory service in many environments readily crack in Practice G36, but have not cracked during interlaboratory testing using this sodium chloride test method. 1.3 This boiling sodium chloride test method was used in an interlaboratory test program to evaluate wrought stainless steels, including duplex (ferrite-austenite) stainless and an alloy with up to about 33% nickel. It may also b...

U.S. position paper on sodium fires, design and testing

International Nuclear Information System (INIS)

Hilliard, R.K.; Johnson, R.P.; Powers, D.A.

1982-05-01

Sodium combustion phenomena and U.S. computer codes developed for sodium fires are discussed. Ways of preventing and mitigating sodium fires are described. Effects of sodium fires and spills on LMFBR structural materials, thermal insulation materials, and equipment/instrumentation are considered

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

Tests of a sodium bonded type control element of 'Monju' with the actual conditions. Pt. 1

International Nuclear Information System (INIS)

Katoh, Jungo; Tanaka, Masako; Ikarimoto, Iwao

2004-03-01

This report provides a summary of the following three experimental and analytical results conducted on the sodium bonded double porous-plug type control element: the pretest of retention of the absorber in the double C-type shroud tube (the ASMI test); the stress analysis of the double C-type shroud tube; and the long-term submersion test of the sodium bonded type control element. The objective of the ASMI test and the stress analysis is to define the shroud tube's bottom structure. First, simulating the absorber shroud mechanical interaction (ASMI), we checked the function of absorber holding of the double C-type shroud tube. Second, we conducted the stress analysis, changing the distance between the end plug of the tube and the bottom end of the absorber, to determine proper position of the absorber. In addition, we conducted the long-term submersion test of the control element on which sodium residues had been left, and confirmed that no trouble occurred even though the element was submerged for a long time. The results of the above tests and analysis are as follows: (1) On the conditions that the end plug of the shroud tube was fitted from outside of the double C-type shroud tube was double-layered, we checked through the ASMI test that no absorber leaked and the holding property was secured even though the absorber be powdery. (2) The stress analysis showed that the structural integrity was satisfied when the distance between the end plug of the tube and the bottom end of the absorber was 40 mm or longer, and that the stress in the shroud tube was within the elastic range when the distance was 50 mm or longer. From these results, we could assure the structure integrity of 50 mm distance case. (3) The result of the long-term submersion test showed that no trouble occurred when the sodium bonded type control rod assembly (control element) is stored under water with sodium residues after cleaning in the sodium cleansing facility, the temperature increase and the

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)

Surveillance test of the JMTR core components

International Nuclear Information System (INIS)

Takeda, Takashi; Amezawa, Hiroo; Tobita, Kenji

1986-02-01

Surveillance test for the core components of Japan Materials Testing Reactor (JMTR) was started in 1966, and completed in 1985 without one capsule. Most of capsules in the program, except one beryllium specimens, were removed from the core, and carred out the post-irradiation tests at the JMTR Hot Laboratory. The data is applied to review of JMTR core components management plan. JMTR surveillance test was carried out with several kind of materials of JMTR core components, Berylium as the reflector, Hafnium as the neutron absorber of control rod, 17-4PH stainless steel as a roller spring of the control rod, and 304 stainless steel as the grid plate. Results are described in this report. (author)

Development of Sodium Technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, B. H.; Nam, H. Y.; Kim, T. J. (and others)

2007-03-15

In the experiments to investigate the characteristics of the free surface fluctuation in a vessel, the experimental correlation was modeled to describe the free surface fluctuation in the upper plenum of a liquid metal reactor within 95% reliability and 2.4% error. The correlation was used to verify the computational model. The new conceptual flowmeters were suggested to measure the sodium flow for the reliability enhancement. The electromagnetic flowmeter with permanent magnet showed a good linearity and repeatability. For reuse of the sodium contaminated component, CO{sub 2} bubbling method was developed. Sodium in 0.3mm crevice specimen was removed completely. The optimum condition for the used sodium treatment was deduced to estimate which reaction is more safe and adequate for operation condition by analyzing the reactivity alleviation condition and the reaction rate with the control of sodium hydroxide concentration A series of tests were carried out to investigate the enlargement rate of the nozzle hole itself and the sodium-water reaction temperature associated with needle-like jets of a high-pressure water/steam into the sodium side of a steam generator. The size of the nozzle hole became larger with an increased duration of the steam injection both for the 2.25Cr-1Mo and M9Cr-1Mo steels by a self-wastage phenomenon. For developing the SWR acoustic leak detection technology, the tool prepared by the LabVIEW was installed with the system, and confirmed the performance of the on-line acoustic leak detection tool using the SWR leak signal acquired in the KAERI facility.

Sodium corrosion tests in the ML 1 circuit

International Nuclear Information System (INIS)

Borgstedt, H.U.

1977-01-01

In the ML-1 circuit of the 'Juan Vigon' research centre in Madrid, sodium corrosion tests are being carried out on the austenitic steels DIN 1.4970 (X10NiCrMoTiB1515) and DIN 1.4301 (X5CrNi189) at temperatures between 500 and 700 0 C. The exposure time of the samples amounts to 6,000 h by now. Every 1,000 h, the samples were weighed in order to measure corrosion and deposition effects. After 3,000 and 6,000 h, some selected samples were destroyed for inspection. The results are given. (GSC) [de

Development of sodium disposal technology. Experiment of sodium compound solidification process

International Nuclear Information System (INIS)

Matsumoto, Toshiyuki; Ohura, Masato; Yatoh, Yasuo

2007-07-01

A large amount of sodium containing radioactive waste will come up at the time of final shutdown/decommission of FBR plant. The radioactive waste is managed as solid state material in a closed can in Japan. As for the sodium, there is no established method to convert the radioactive sodium to solid waste. Further, the sodium is highly reactive. Thus, it is recommended to convert the sodium to a stable substance before the solidification process. One of the stabilizing methods is conversion of sodium into sodium hydroxide solution. These stabilization and solidification processes should be safe, economical, and efficient. In order to develop such sodium disposal technology, nonradioactive sodium was used and a basic experiment was performed. Waste-fluid Slag Solidification method was employed as the solidification process of sodium hydroxide solution. Experimental parameters were mixing ratio of the sodium hydroxide and the slag solidification material, temperature and concentration of the sodium hydroxide. The best parameters were obtained to achieve the maximum filling ratio of the sodium hydroxide under a condition of enough high compressive strength of the solidified waste. In a beaker level test, the solidified waste was kept in a long term and it was shown that there was no change of appearance, density, and also the compressive strength was kept at a target value. In a real scale test, homogeneous profiles of the density and the compressive strength were obtained. The compressive strength was higher than the target value. It was shown that the Waste-fluid Slag Solidification method can be applied to the solidification process of the sodium hydroxide solution, which was produced by the stabilization process. (author)

Utilization of sodium bicarbonate for the neutralization of acid components in exhaust gases; Verwendung von Natriumbicarbonat zur Neutralisation saurer Bestandteile in Abgasen

Energy Technology Data Exchange (ETDEWEB)

Bauer, Thomas [Solvay Chemicals GmbH, Rheinberg (Germany). Technisches Marketing

2013-03-01

The SOLVAIR dry process using sodium bicarbonate as absorbent facilitates an uncomplicated and efficient purification of exhaust gases from different processes. The products from exhaust gas purification can be used either directly in a producing process or subsequently to a treatment process in the chemical process. The author of this contribution reports on the utilization of sodium bicarbonate in the neutralisation of pour components in exhaust gases such as hydrochloric acid, sulphur dioxide and nitrous dioxides. Further aspects of this contribution are the energy efficiency and the recycling of reaction products.

Final report on the safety assessment of sodium sulfite, potassium sulfite, ammonium sulfite, sodium bisulfite, ammonium bisulfite, sodium metabisulfite and potassium metabisulfite.

Science.gov (United States)

Nair, Bindu; Elmore, Amy R

2003-01-01

Sodium Sulfite, Ammonium Sulfite, Sodium Bisulfite, Potassium Bisulfite, Ammonium Bisulfite, Sodium Metabisulfite, and Potassium Metabisulfite are inorganic salts that function as reducing agents in cosmetic formulations. All except Sodium Metabisulfite also function as hair-waving/straightening agents. In addition, Sodium Sulfite, Potassium Sulfite, Sodium Bisulfite, and Sodium Metabisulfite function as antioxidants. Although Ammonium Sulfite is not in current use, the others are widely used in hair care products. Sulfites that enter mammals via ingestion, inhalation, or injection are metabolized by sulfite oxidase to sulfate. In oral-dose animal toxicity studies, hyperplastic changes in the gastric mucosa were the most common findings at high doses. Ammonium Sulfite aerosol had an acute LC(50) of >400 mg/m(3) in guinea pigs. A single exposure to low concentrations of a Sodium Sulfite fine aerosol produced dose-related changes in the lung capacity parameters of guinea pigs. A 3-day exposure of rats to a Sodium Sulfite fine aerosol produced mild pulmonary edema and irritation of the tracheal epithelium. Severe epithelial changes were observed in dogs exposed for 290 days to 1 mg/m(3) of a Sodium Metabisulfite fine aerosol. These fine aerosols contained fine respirable particle sizes that are not found in cosmetic aerosols or pump sprays. None of the cosmetic product types, however, in which these ingredients are used are aerosolized. Sodium Bisulfite (tested at 38%) and Sodium Metabisulfite (undiluted) were not irritants to rabbits following occlusive exposures. Sodium Metabisulfite (tested at 50%) was irritating to guinea pigs following repeated exposure. In rats, Sodium Sulfite heptahydrate at large doses (up to 3.3 g/kg) produced fetal toxicity but not teratogenicity. Sodium Bisulfite, Sodium Metabisulfite, and Potassium Metabisulfite were not teratogenic for mice, rats, hamsters, or rabbits at doses up to 160 mg/kg. Generally, Sodium Sulfite, Sodium

French position paper on sodium fires, design and testing

International Nuclear Information System (INIS)

Malet, J.C.; Casselman, C.; Charpenel, J.; Duverger de Cuy, G.; Rzekiecki, R.; Dufresne, J.; Lo Pinto, P.; Montaignac de; Herault, A.

1983-01-01

This document gives an up-to-date account of studies carried out in France in the fields of sodium fires, extinction, filtration and prefiltration of aerosols originating from a sodium fire, and of sodium concrete reactions. It also indicates how the CEA, in collaboration with the CNEN, intends to deal with scale effects. (author)

Analysis of acoustic data from UK sodium/water reaction test facilities

International Nuclear Information System (INIS)

Rowley, R.; Mcknight, J.A.; Airey, J.

1990-01-01

This paper describes acoustic measurements made during a number of sodium/water reaction experiments in the UK. The tests have included water and steam injections through both realistic (fatigue crack) defects and machined orifices and have covered a range of experimental conditions including those appropriate to the inlet and outlet regions of the EFR steam generators. Injection rates were typically in the range 0.1 to 30 g/s. Where possible, gas injections were also included in the test programme for comparison, since it is anticipated that a practical SGU acoustic leak detection system would include a facility for gas injections to allow system calibration, and to confirm transmission properties within the SGU. The test sections were instrumented with accelerometers on waveguides and in some cases included an under-sodium microphone situated about 300mm above the reaction zone. Tape recordings were made during the tests and used for detailed analysis off-line, although an audible output from one of the acoustic channels was used to monitor the progress of the injections and provide information for the rig operators. A comparison of the signal amplitudes measured during the experiments with typical reactor background noise was made and an estimate of the detection sensitivity of an acoustic monitoring system was deduced. 3 refs, 5 figs, 1 tab

Investigation of Plugging of Narrow Sodium Channels by Sodium and Carbon Dioxide Interaction

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong; Kim, Tae-joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The supercritical CO{sub 2} Brayton cycle system is known to be a promising power conversion system for improving the efficiency and preventing the sodium water reaction (SWR) of the current SFR concept using a Rankine steam cycle. PCHEs are known to have potential for reducing the volume occupied by the sodium-to-CO{sub 2} exchangers as well as the heat exchanger mass relative to traditional shell-and-tube heat exchangers. Here, we report a study on a plugging test by the interaction of sodium and CO{sub 2} to investigate design parameters of sodium channels in the realistic operating conditions. We investigated a plugging test by an interaction of sodium and CO{sub 2} with different cross sectional areas of the sodium channels. It was found that the flow rate of sodium decreased earlier and faster with a narrower cross sectional area compared to a wider one. Our experimental results are expected to be used for determining the sodium channel areas of PCHEs.

Sodium fill of FFTF

International Nuclear Information System (INIS)

Waldo, J.B.; Greenwell, R.K.; Keasling, T.A.; Collins, J.R.; Klos, D.B.

1980-02-01

With construction of the Fast Flux Test Facility (FFTF) completed, the first major objective in the startup program was to fill the sodium systems. A sodium fill sequence was developed to match construction completion, and as systems became available, they were inerted, preheated, and filled with sodium. The secondary sodium systems were filled first while dry refueling system testing was in progress in the reactor vessel. The reactor vessel and the primary loops were filled last. This paper describes the methods used and some of the key results achieved for this major FFTF objective

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.)

Design and test of a vacuum distillation method for determining carbon in sodium

International Nuclear Information System (INIS)

Irmisch, R.; Rettig, D.; Woelke, K.

1976-08-01

A method is described for determining total and carbonate carbon in sodium samples until 10 g. Sodium is removed by vacuum distillation at 300 0 C and the carbon in the residue is converted to carbon dioxide by combustion in a stream of air or thermic splitting in a stream of cover gas at 1000 0 C. The carbon dioxide is measured manometrically. It is therefore not necessary to carry out calibration. Distillation and combustion rig are combined with inertgas filled transfer box. Therefore the sodium sample does not get into touch with air. Test of this method was carried out with Na 2 CO 3 and WC. Carbon recoveries were for Na 2 CO 3 between 103 and 107% and for WC between 92 and 96%. The blank value found being 9 μg C and sensitivity 3 μg C. (author)

Friction behavior of cobalt base and nickel base hardfacing materials in high temperature sodium

International Nuclear Information System (INIS)

Mizobuchi, Syotaro; Kano, Shigeki; Nakayama, Kohichi; Atsumo, Hideo

1980-01-01

A friction behavior of the hardfacing materials such as cobalt base alloy ''Stellite'' and nickel base alloy ''Colmonoy'' used in the sliding components of a sodium cooled fast breeder reactor was investigated in various sodium environments. Also, friction tests on these materials were carried out in argon environment. And they were compared with those in sodium environment. The results obtained are as follows: (1) In argon, the cobalt base hardfacing alloy showed better friction behavior than the nickel base hardfacing alloy. In sodium, the latter was observed to have the better friction behavior being independent of the sodium temperature. (2) The friction coefficient of each material tends to become lower by pre-exposure in sodium. Particularly, this tendency was remarkable for the nickel base hardfacing alloy. (3) The friction coefficient between SUS 316 and one of these hardfacing materials was higher than that between latter materials. Also, some elements of hardfacing alloys were recognized to transfer on the friction surface of SUS 316 material. (4) It was observed that each tested material has a greater friction coefficient with a decrease of the oxygen content in sodium. (author)

Liquid sodium pool fires

Energy Technology Data Exchange (ETDEWEB)

Casselman, C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

1979-03-01

Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

Liquid sodium pool fires

International Nuclear Information System (INIS)

Casselman, C.

1979-01-01

Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

EBR-II facility for cleaning and maintenance of LMR components

International Nuclear Information System (INIS)

Washburn, R.A.

1986-01-01

The cleaning and maintenance of EBR-II sodium wetted components is accomplished in a separate hands-on maintenance facility known as the Sodium Components Maintenance Shop (SCMS). Sodium removal is mostly done using alcohol but steam or water is used. The SCMS has three alcohol cleaning systems: one for small nonradioactive components, one for small radioactive components, and one for large radioactive components. The SCMS also has a water-wash station for the removal of sodium with steam or water. An Alcohol Recovery Facility removes radioactive contaminants from the alcohol and reclaims the alcohol for reuse. Associated with the large components cleaning system is a major component handling system

The Initial Comparison Study of Sodium Lignosulfonate, Sodium Dodecyl Benzene Sulfonate, and Sodium p-Toluene Sulfonate Surfactant for Enhanced Oil Recovery

Science.gov (United States)

Khoirul Anas, Argo; Iman Prakoso, Nurcahyo; Sasvita, Dilla

2018-04-01

Surfactant (surface active agent) exhibit numerous interesting properties that enable their use as additional component in mobilising of residual oil from capillary pore after secondary recovery process using gas injection and water flooding. In this study, Sodium Lignosulfonate (SLS) surfactant was successfully synthesized by applying batch method using lignin from oil palm empty fruit bunches as precursor. Furthermore, its performance in reducing interfacial tension of crude oil and formation water colloidal system was compared with commercial available surfactant including Sodium Dodecyl Benzene Sulfonate (SDBS) and Sodium p-Toluene Sulfonate (SpTS). The synthesized SLS surfactant was characterized by using Fourier Transform Infrared (FTIR) spectroscopy. Meanwhile, its performance in reducing interfacial tension of crude oil and formation water colloidal system was analyzed by using compatibility test, phase behaviour analysis, and interfacial tension (IFT) measurement. The compatibility test shows that SLS, SDBS, and SpTS surfactants were compatible with formation water. In addition, the phase behaviour analysis shows that SLS surfactant was better than SpTS surfactant, while SDBS surfactant generates the highest performance proved by the best microemulsion formation resulted by SDBS. Furthermore, the optimum concentration of SLS, SDBS, and SpTS surfactants in reducing the interfacial tension of crude oil and formation water was 1.0%. The IFT measurement indicates that the performance of SLS with the value of 1.67 mN/m was also better than SpTS surfactant with the value of 3.59 mN/m. Meanwhile, SDBS surfactant shows the best performance with the IFT value of 0.47 mN/m.

Effects of dietary sodium on metabolites: the Dietary Approaches to Stop Hypertension (DASH)-Sodium Feeding Study.

Science.gov (United States)

Derkach, Andriy; Sampson, Joshua; Joseph, Justin; Playdon, Mary C; Stolzenberg-Solomon, Rachael Z

2017-10-01

Background: High sodium intake is known to increase blood pressure and is difficult to measure in epidemiologic studies. Objective: We examined the effect of sodium intake on metabolites within the DASH (Dietary Approaches to Stop Hypertension Trial)-Sodium Trial to further our understanding of the biological effects of sodium intake beyond blood pressure. Design: The DASH-Sodium Trial randomly assigned individuals to either the DASH diet (low in fat and high in protein, low-fat dairy, and fruits and vegetables) or a control diet for 12 wk. Participants within each diet arm received, in random order, diets containing high (150 nmol or 3450 mg), medium (100 nmol or 2300 mg), and low (50 nmol or 1150 mg) amounts of sodium for 30 d (crossover design). Fasting blood samples were collected at the end of each sodium intervention. We measured 531 identified plasma metabolites in 73 participants at the end of their high- and low-sodium interventions and in 46 participants at the end of their high- and medium-sodium interventions ( N = 119). We used linear mixed-effects regression to model the relation between each log-transformed metabolite and sodium intake. We also combined the resulting P values with Fisher's method to estimate the association between sodium intake and 38 metabolic pathways or groups. Results: Six pathways were associated with sodium intake at a Bonferroni-corrected threshold of 0.0013 (e.g., fatty acid, food component or plant, benzoate, γ-glutamyl amino acid, methionine, and tryptophan). Although 82 metabolites were associated with sodium intake at a false discovery rate ≤0.10, only 4-ethylphenylsufate, a xenobiotic related to benzoate metabolism, was significant at a Bonferroni-corrected threshold ( P Sodium intake is associated with changes in circulating metabolites, including gut microbial, tryptophan, plant component, and γ-glutamyl amino acid-related metabolites. This trial was registered at clinicaltrials.gov as NCT00000608. © 2017

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)

The report of inspection and repair technology of sodium cooled reactors

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru

2002-12-01

Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. 1) Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow system must be evaluated to realize its performance. 2) In-service inspection (ISI and R). The viewpoint of the commercialized plant's ISI and R was organized by comparing with the prototype reactor's ISI and R method. We also investigated short-term ISI and R method without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize the with the present technology. Hereafter, the ISI and R of the commercialized plants must be defined by considering its characteristics. 3) Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was

Annular sodium flowsensor

International Nuclear Information System (INIS)

Kaiser, W.C.; Brewer, J.; Forster, G.A.

1983-01-01

This paper describes a unique eddy-current type liquid sodium flowsensor, designed as a joint effort between Argonne National Laboratory and Kaman Instrumentation Corp. Test results are included for operation of the flowsensor mounted on a sodium test loop whose configuration simulates the actual operating conditions, except for the magnetic field of the ALIP

An experience with in-service fabrication and inspection of austenitic stainless steel piping in high temperature sodium system

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-01

Highlights: • Procedure for changing 304L SS pipe to 316L SS in sodium loop has been established. • Hot leg made of 304L SS was isolated from existing cold leg made of 316LN SS. • Innovative welding was used in joining the new 316L SS pipe with existing 316LN SS. • The old components of 304L SS piping have been integrated with the new piping. - Abstract: A creep testing facility along with dynamic sodium loop was installed at Indira Gandhi Centre for Atomic Research, Kalpakkam, India to assess the creep behavior of fast reactor structural materials in flowing sodium. Type 304L austenitic stainless steel was used in the low cross section piping of hot-leg whereas 316LN austenitic stainless steel in the high cross section cold-leg of the sodium loop. The intended service life of the sodium loop was 10 years. The loop has performed successfully in the stipulated time period. To enhance its life time, it has been decided to replace the 304L piping with 316L piping in the hot-leg. There were more than 300 welding joints involved in the integration of cold-leg with the new 316L hot-leg. Continuous argon gas flow was maintained in the loop during welding to avoid contamination of sodium residue with air. Several innovative welding procedures have been adopted for joining the new hot-leg with the existing cold-leg in the presence of sodium residue adopting TIG welding technique. The joints were inspected for 100% X-ray radiography and qualified by performing tensile tests. The components used in the discarded hot-leg were retrieved, cleaned and integrated in the renovated loop. A method of cleaning component of sodium residue has been established. This paper highlights the in-service fabrication and inspection of the renovation.

On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

Directory of Open Access Journals (Sweden)

Jong-Bum Kim

2016-10-01

Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

International Nuclear Information System (INIS)

Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

2016-01-01

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

Sodium concrete reaction - Structural considerations

International Nuclear Information System (INIS)

Ferskakis, G.N.

1984-01-01

An overview of the sodium concrete reaction phenomenon, with emphasis on structural considerations, is presented. Available test results for limestone, basalt, and magnetite concrete with various test article configurations are reviewed. Generally, tests indicate reaction is self limiting before all sodium is used. Uncertainties, however, concerning the mechanism for penetration of sodium into concrete have resulted in different theories about a reaction model. Structural behavior may be significant in the progression of the reaction due to thermal-structuralchemical interactions involving tensile cracking, compressive crushing, or general deterioration of concrete and the exposure of fresh concrete surfaces to react with sodium. Structural behavior of test articles and potential factors that could enhance the progression of the reaction are discussed

Fiber Laser Component Testing for Space Qualification Protocol Development

Science.gov (United States)

Falvey, S.; Buelow, M.; Nelson, B.; Starcher, Y.; Thienel, L.; Rhodes, C.; Tull, Jackson; Drape, T.; Westfall, C.

A test protocol for the space qualifying of Ytterbium-doped diode-pumped fiber laser (DPFL) components was developed under the Bright Light effort, sponsored by AFRL/VSE. A literature search was performed and summarized in an AMOS 2005 conference paper that formed the building blocks for the development of the test protocol. The test protocol was developed from the experience of the Bright Light team, the information in the literature search, and the results of a study of the Telcordia standards. Based on this protocol developed, test procedures and acceptance criteria for a series of vibration, thermal/vacuum, and radiation exposure tests were developed for selected fiber laser components. Northrop Grumman led the effort in vibration and thermal testing of these components at the Aerospace Engineering Facility on Kirtland Air Force Base, NM. The results of the tests conducted have been evaluated. This paper discusses the vibration and thermal testing that was executed to validate the test protocol. The lessons learned will aid in future assessments and definition of space qualification protocols. Components representative of major items within a Ytterbium-doped diode-pumped fiber laser were selected for testing; including fibers, isolators, combiners, fiber Bragg gratings, and laser diodes. Selection of the components was based on guidelines to test multiple models of typical fiber laser components. A goal of the effort was to test two models (i.e. different manufacturers) of each type of article selected, representing different technologies for the same type of device. The test articles did not include subsystems or systems. These components and parts may not be available commercial-off-the-shelf (COTS), and, in fact, many are custom articles, or newly developed by the manufacturer. The primary goal for this effort is a completed taxonomy that lists all relevant laser components, modules, subsystems, and interfaces, and cites the documentation for space

The influence of particles of a minor component on the matrix strength of sodium chloride

NARCIS (Netherlands)

Van Veen, B.; van der Voort Maarschalk, Kees; Bolhuis, G.K; Gons, M.; Zuurman, K.; Frijlink, H.W

2002-01-01

This paper deals with the matrix strength of sodium chloride particles in pure sodium chloride tablets and in tablets compressed from binary mixtures of sodium chloride with low concentrations of pregelatinised starch. Because this study concerns the strength of the sodium chloride matrix, the

Sodium oxide and uranium oxide aerosol experiments: NSPP Tests 106-108 and Tests 204-207, data record report

Energy Technology Data Exchange (ETDEWEB)

Adams, R.E.; Kress, T.S.; Tobias, M.L.

1981-03-01

This data record report describes three sodium oxide aerosol tests and four uranium oxide aerosol tests conducted in the Nuclear Safety Pilot Plant project at Oak Ridge National Laboratory. The goal of this project is to establish the validity (or level of conservatism) of the aerosol behavioral code, HAARM-3, and follow-on codes under development at the Battelle Columbus Laboratories for the US Nuclear Regulatory Commission. Descriptions of the seven tests with tables and graphs summarizing the results are included. 92 figs.

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

An automatic sodium-loop for testing the lon-term behaviour of sintered bodies flowed through by gas

International Nuclear Information System (INIS)

Barkleit, G.; George, G.; Haase, I.; Kiessling, W.

1980-08-01

An automatic sodium loop NAKOS for testing the long-term behaviour of porous stainless steel bodies which are flowed through by gas is described. The loop using a special safety protection system is capable of working without control up to 1000 h. During a 500 h-experiment the safety system and the gas permeability measuring method for testing the porous bodies were tested. Both first results of the behaviour of sintered bodies in liquid sodium of high purity and temperatures of about 850 K and some details of the production of these bodies are given. (author)

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

Sodium characterization during the starting period of a sodium loop

International Nuclear Information System (INIS)

Lievens, F.; Parmentier, C.; Soenen, M.

1976-01-01

A sodium loop for analytical chemistry studies has been built by S.C.K./C.E.N. at Mol Belgium. Its first working period was used to test analytical methods, to characterize the sodium and to define the operating parameters of the loop. This report covers the working parameters of the loop, the characterization of the filling sodium and its purity evolution during the first working period of the loop

Characteristic test technology for PWR fuel and its components

Energy Technology Data Exchange (ETDEWEB)

Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-01-01

Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

Patterns in the Parathyroid Response to Sodium Bicarbonate Infusion Test in Healthy Volunteers

Directory of Open Access Journals (Sweden)

Theodossis S. Papavramidis

2014-01-01

Full Text Available Background. The sodium bicarbonate infusion test evaluates the function of the parathyroid glands. The present study aims to evaluate the range of parathyroid response in healthy individuals and the potential influence of various factors. Methods. Fifty healthy volunteers were subjected to the test. Levels of vitamin D, calcium, albumin, and PTH were measured before infusion. PTH was measured at 3, 5, 10, 30, and 60 minutes after infusion. Results. A curve describing the response of parathyroids to the test was drawn. Twenty percent of the subjects had blunted PTH response. No significant difference was observed between normal and blunted responders concerning age, BMI, baseline PTH, or calcium levels. Nonetheless, there was a significant difference in vitamin D levels (P=0.024. Interpretation. The test is easy to perform and may be used for everyday screening. It has to be clarified whether our observations are, at least partly, produced due to the presence of individuals with a constitutively blunted response or if low levels of vitamin D decrease the ability of the parathyroids to respond. Whichever the case, PTH response of normal individuals to sodium bicarbonate infusion test is more varied than previously thought and vitamin D levels influence it.

Characterization of ion distributions near the surface of sodium-containing and sodium-depleted calcium aluminosilicate glass melts

International Nuclear Information System (INIS)

Corrales, Louis R.; Du, Jincheng

2006-01-01

The distribution of cation and anion components of sodium containing calcium aluminosilicate glass was studied by classical molecular dynamics simulations in a high temperature melt in the bulk and at the vacuum-melt interface. A significant redistribution of the sodium and non-bridging oxygen ions was observed. Subsequently, a sodium depleted calcium aluminosilicate glass melt was simulated to determine the sensitivity of the redistribution of ions near the vacuum-melt interface to the presence of sodium ions. It is found that the thermodynamic equilibrium condition near a surface favors the enrichment of non-bridging oxygen ions that is closely associated with enrichment of the sodium ions

Final report on the safety assessment of potassium silicate, sodium metasilicate, and sodium silicate.

Science.gov (United States)

Elmore, Amy R

2005-01-01

Metasilicate, and Sodium Silicate ranged from negligible to severe, depending on the species tested and the molar ratio and concentration tested. Sodium Metasilicate was negative in the local lymph node assay (LLNA), but a delayed-type hypersensitivity response was observed in mice. Potassium Silicate was nonirritating in two acute eye irritation studies in rabbits. Sodium Metasilicate (42.4% H2O) was corrosive to the rabbit eye. Sodium Silicate was a severe eye irritant in some eye irritation studies, but was irritating or nonirritating in others. A skin freshener containing Sodium Silicate was nonirritating. Sodium Metasilicate was nonmutagenic in bacterial cells. Rats given Sodium Silicate (600 and 1200 ppm of added silica) in the drinking water in reproductive studies produced a reduced number of offspring: to 67% of controls at 600 ppm and to 80% of controls at 1200 ppm. Three adult rats injected intratesticularly and subcutaneously with 0.8 mM/kg of Sodium Silicate showed no morphological changes in the testes and no effect on the residual spermatozoa in the ductus deferens. Sodium Metasilicate (37% in a detergent) mixed with water was a severe skin irritant when tested on intact and abraded human skin, but 6%, 7%, and 13% Sodium Silicate were negligible skin irritants to intact and abraded human skin. Sodium Silicate (10% of a 40% aqueous solution) was negative in a repeat-insult predictive patch test in humans. The same aqueous solution of Sodium Silicate was considered a mild irritant under normal use conditions in a study of cumulative irritant properties. The Cosmetic Ingredient Review (CIR) Expert Panel recognized the irritation potential of these ingredients, especially in leave-on products. However, because these ingredients have limited dermal absorption and Sodium Metasilicate is a GRAS direct food substance, the Panel deemed the ingredients safe for use in cosmetic products in the practices of use and concentration described in this safety assessment, when

On Perturbation Components Correspondence between Diffusion and Transport

Energy Technology Data Exchange (ETDEWEB)

G. Palmiotti

2012-11-01

We have established a correspondence between perturbation components in diffusion and transport theory. In particular we have established the correspondence between the leakage perturbation component of the diffusion theory to that of the group self scattering in transport theory. This has been confirmed by practical applications on sodium void reactivity calculations of fast reactors. Why this is important for current investigations? Recently, there has been a renewed interest in designing fast reactors where the sodium void reactivity coefficient is minimized. In particular the ASTRID8,9 reactor concept has been optimized with this goal in mind. The correspondence on the leakage term that has been established here has a twofold implication for the design of this kind of reactors. First, this type of reactor has a radial reflector; therefore, as shown before, the sodium void reactivity coefficient calculation requires the use of transport theory. The minimization of the sodium reactivity coefficient is normally done by increasing the leakage component that has a negative sign. The correspondence established in this paper allows to directly look at this component in transport theory. The second implication is related to the uncertainty evaluation on sodium void reactivity. As it has shown before, the total sodium void reactivity effect is the result of a large compensation (opposite sign) between the scattering (called often spectral) component and the leakage one. Consequently, one has to evaluate separately the uncertainty on each separate component and then combine them statistically. If one wants to compute the cross section sensitivity coefficients of the two different components, the formulation established in this paper allows to achieve this goal by playing on the contribution to the sodium void reactivity coming from the group self scattering of the sodium cross section.

Fire protection at the Fast Flux Test Facility (a sodium cooled test reactor)

International Nuclear Information System (INIS)

Bell, J.R.

1980-01-01

For purposes of this presentation, fire protection at the FFTF is subdivided into two catagories; protection for non-sodium areas and protection for areas containing sodium. Fire protection systems and philosophies for non-sodium areas at the FFTF are very similar to those used at conventional power plants being constructed throughout the country. They follow, essentially, the NRC rules and guidelines and ANSI 59.4 Generic Requirements for Light Water Nuclear Power Plant Fire Protection. The FFTF with its support facilities have their own water system comprised of a looped 8'' and 10'' underground distribution system, three 1500 GPM fire pumps and three ground level storage tanks totaling 736,000 gallons with 420,000 reserved for fire protection. Fire hydrants are enclosed with hose houses outfitted for use by the Emergency Response Team (ERT). Fire prevention systems for sodium areas of the FFTF are also described

Technology for analysis of sodium pool fire characteristics

Energy Technology Data Exchange (ETDEWEB)

Kim, Byung Ho; Jeong, K C; Jeong, J Y; Kim, T J; Choi, J H; Choi, Y D; Hwang, S T

2000-09-01

Basic and detail design for medium sodium fire test facility was carried out and medium sodium fire test facility was constructed. Design data is as follows. - Test cell material : Concrete with high strength - Test cell dimension ; 48m{sup 3} (3x4x4m) - Design temp. ; 700 deg C - Operation temp. ; 530 deg C - Design pressure ; 1 bar (max.) - Dimension(Inside) : 3 x 4 x 4(m) - Test cell thickness ; 45cm - Liner plate with (Thickness : 3mm) In this study, sodium fire characteristics was analyzed and data for validation of computer code was produced. Oxygen and sodium filled in pool pan didn't burns instantly, but pool fire occurred through pre-ignition. Distribution of temperature in test cell was divided by two parts, and temperature at upper position appeared to be higher than temperature at lower position. The temperature in test cell increased with the feed of sodium. The pressure in test cell increased with the feed of sodium. When the feed of sodium was 8kg, peak pressure was 0.075 bar. Peak temperature in sodium pool appeared to be 854 deg C regardless of the feed of sodium. Decrease of 1% in oxygen concentration showed the rise of 0.036bar in pressure.

Technology for analysis of sodium pool fire characteristics

International Nuclear Information System (INIS)

Kim, Byung Ho; Jeong, K. C.; Jeong, J. Y.; Kim, T. J.; Choi, J. H.; Choi, Y. D.; Hwang, S. T.

2000-09-01

Basic and detail design for medium sodium fire test facility was carried out and medium sodium fire test facility was constructed. Design data is as follows. - Test cell material : Concrete with high strength - Test cell dimension ; 48m 3 (3x4x4m) - Design temp. ; 700 deg C - Operation temp. ; 530 deg C - Design pressure ; 1 bar (max.) - Dimension(Inside) : 3 x 4 x 4(m) - Test cell thickness ; 45cm - Liner plate with (Thickness : 3mm) In this study, sodium fire characteristics was analyzed and data for validation of computer code was produced. Oxygen and sodium filled in pool pan didn't burns instantly, but pool fire occurred through pre-ignition. Distribution of temperature in test cell was divided by two parts, and temperature at upper position appeared to be higher than temperature at lower position. The temperature in test cell increased with the feed of sodium. The pressure in test cell increased with the feed of sodium. When the feed of sodium was 8kg, peak pressure was 0.075 bar. Peak temperature in sodium pool appeared to be 854 deg C regardless of the feed of sodium. Decrease of 1% in oxygen concentration showed the rise of 0.036bar in pressure

Technology for analysis of sodium pool fire characteristics

Energy Technology Data Exchange (ETDEWEB)

Kim, Byung Ho; Jeong, K. C.; Jeong, J. Y.; Kim, T. J.; Choi, J. H.; Choi, Y. D.; Hwang, S. T

2000-09-01

Basic and detail design for medium sodium fire test facility was carried out and medium sodium fire test facility was constructed. Design data is as follows. - Test cell material : Concrete with high strength - Test cell dimension ; 48m{sup 3} (3x4x4m) - Design temp. ; 700 deg C - Operation temp. ; 530 deg C - Design pressure ; 1 bar (max.) - Dimension(Inside) : 3 x 4 x 4(m) - Test cell thickness ; 45cm - Liner plate with (Thickness : 3mm) In this study, sodium fire characteristics was analyzed and data for validation of computer code was produced. Oxygen and sodium filled in pool pan didn't burns instantly, but pool fire occurred through pre-ignition. Distribution of temperature in test cell was divided by two parts, and temperature at upper position appeared to be higher than temperature at lower position. The temperature in test cell increased with the feed of sodium. The pressure in test cell increased with the feed of sodium. When the feed of sodium was 8kg, peak pressure was 0.075 bar. Peak temperature in sodium pool appeared to be 854 deg C regardless of the feed of sodium. Decrease of 1% in oxygen concentration showed the rise of 0.036bar in pressure.

Mixed waste chemical compatibility with packaging components

International Nuclear Information System (INIS)

Nigrey, P.J.; Conroy, M.; Blalock, L.B.

1994-01-01

In this paper, a chemical compatibility testing program for packaging of mixed wastes at will be described. We will discuss the choice of four y-radiation doses, four time durations, four temperatures and four waste solutions to simulate the hazardous waste components of mixed wastes for testing materials compatibility of polymers. The selected simulant wastes are (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. A selection of 10 polymers with anticipated high resistance to one or more of these types of environments are proposed for testing as potential liner or seal materials. These polymers are butadiene acrylonitrile copolymer, cross-linked polyethylene, epichlorhyarin, ethylene-propylene rubber, fluorocarbon, glass-filled tetrafluoroethylene, high-density poly-ethylene, isobutylene-isoprene copolymer, polypropylene, and styrene-butadiene rubber. We will describe the elements of the testing plan along with a metric for establishing time resistance of the packaging materials to radiation and chemicals

Large-component handling equipment and its use

International Nuclear Information System (INIS)

Krieg, S.A.; Swannack, D.L.

1983-01-01

The Fast Flux Test Facility (FFTF) reactor systems have special requirements for component replacements during maintenance servicing. Replacement operations must address handling of equipment within shielded metal containers while maintaining an inert atmosphere to prevent reaction of sodium with air. Plant identification of a failed component results in selecting and assembling the maintenance cask and equipment transport system for transfer from the storage facility to the Reactor Containment Building (RCB). This includes a proper diameter and length cask, inert atmosphere control consoles, component lift fixture and support structure for interface with the facility area surrounding the component. This equipment is staged in modular groups in the Reactor Service Building for transfer through the equipment airlock to the containment interior. The failed component is generally prepared for replacement by installation of the special lifting fixture attachment. Assembly of the cask support structure is performed over the component position on the containment building operating floor. The cask and shroud from the reactor interface are inerted after all manual service connections and handling attachments are completed. The component is lifted from the reactor and into the cask interior through a floor valve which is then closed to isolate the component reactor port. The cask with sodium wetted component is transferred to a service/repair location, either within containment or outside, to the Maintenance Facility cleaning and repair area. The complete equipment and handling operations for replacement of a large reactor component are described

Fast Flux Test Facility, Sodium Storage Facility project-specific project management plan

International Nuclear Information System (INIS)

Shank, D.R.

1994-01-01

This Project-Specific Project Management Plan describes the project management methods and controls used by the WHC Projects Department to manage Project 03-F-031. The Sodium Storage Facility provides for storage of the 260,000 gallons of sodium presently in the FFTF Plant. The facility will accept the molten sodium transferred from the FFTF sodium systems, and store the sodium in a solid state under an inert cover gas until such time as a Sodium Reaction Facility is available for final disposal of the sodium

Fast Flux Test Facility, Sodium Storage Facility project-specific project management plan

Energy Technology Data Exchange (ETDEWEB)

Shank, D.R.

1994-12-29

This Project-Specific Project Management Plan describes the project management methods and controls used by the WHC Projects Department to manage Project 03-F-031. The Sodium Storage Facility provides for storage of the 260,000 gallons of sodium presently in the FFTF Plant. The facility will accept the molten sodium transferred from the FFTF sodium systems, and store the sodium in a solid state under an inert cover gas until such time as a Sodium Reaction Facility is available for final disposal of the sodium.

Sodium-23 magnetic resonance brain imaging

International Nuclear Information System (INIS)

Winkler, S.S.; Wisconsin Univ., Madison

1990-01-01

This is a review of recent work in 23 Na MR imaging. The main emphasis of recent papers has been pulse sequences that, with appropriate postprocessing, give images of the fast, slow, and intermediate components of T 2 decay. The assignment of compartmental designation to the T 2 component remains a problem except for homogeneous structures easily identifiable anatomically (ventricles, superior sagittal sinus, globe of the eye). Compartmental distribution of sodium is described. The predominance of the interstitial and plasma compartment, the invisibility of part of the intracellular sodium, and the difficulty in imaging the very fast T 2 component of visible intracellular sodium make the usual Na spin-echo image essentially an image of the interstitial and plasma space. Use of paramagnetic iron oxide coupled to dextran as a contrast medium may help to identify the plasma compartment. Because the usual Na MR images are essentially interstitial and plasma images, our own interest is in observing functional changes in these compartments. Another proposed application is the detection of the very fast T 2 component in brain tumors to aid in defining tumor grade and extent. (orig.)

Technical meeting on decommissioning of fast reactors after sodium draining. Working material

International Nuclear Information System (INIS)

2005-01-01

The objective of the technical meeting was to provide a forum for in-depth scientific and technical exchange on topics related to the decommissioning experience with fast reactors, in particular with regard to the decommissioning of components after sodium draining. Accordingly, the scope of the meeting covers the review and analyses of the experience gained from the decommissioning of both active sodium loops and sodium cooled fast reactors (e.g., KNK II, Superphenix, RAPSODIE, EBR-II, FERMI, BN-350, BR-10). It is expected that the outcome of the meeting will contribute to the Agency initiative to preserve fast reactor data and knowledge. The main focus of the technical meeting was given on the decommissioning of both active loop and reactor components (e.g., the primary vessel of a sodium-cooled reactor) that have been drained of sodium, but that still conserve some residual amounts of sodium (e.g., films covering the entire surface of the component, or particular sodium heels that cannot be drained)

Studies on generation and transport of sodium aerosols in some test facilities

International Nuclear Information System (INIS)

Sano, T.; Shimomura, T.; Hattori, N.

1986-01-01

Technical experiences that have been obtained during the course of the experiments to determine the sodium aerosol concentration, to study the deposition of sodium aerosol, and to predict mechanical properties of sodium vapor deposits are presented. In the first study, the sodium aerosol concentrations in an inert cover gas space over a sodium pool and those following a sodium spray injection into an inert atmosphere were determined. The results from the two different experiments were compared with each other and were discussed in comparison with those from the literature. In the second study, deposition of sodium aerosol following a sodium spray injection into an inert atmosphere was examined. The deposition rates on the walls and the floor of a closed concrete cell were measured, and the results obtained were discussed. The third study relates to the sodium vapor deposition within a narrow annulus. In the experiments, a downward argon gas flow that passes the annulus was fed to prevent sodium vapor deposition. Average sodium vapor deposition rates on the walls of the annulus were determined, then the effect of the downward feed gas was discussed. The last study relates to one of the mechanical properties and the deformation rate of solid sodium being compressed. The purpose of the experiments were to obtain data to predict deformation rate of the sodium deposits. (author)

Testing and performance of electrolytic oxygen meters for use in liquid sodium

International Nuclear Information System (INIS)

Taylor, R.G.; Thompson, R.

1983-01-01

The performance of yttria-doped thoria ceramic electrochemical oxygen meters in liquid sodium is described. Tests were carried out using laboratory loops. Temperature coefficients of the oxygen meters have been measured between 380 0 C and 480 0 C, and the response to changes in oxygen level using cold-trap temperatures from 125 0 C to 250 0 C was determined. The ceramic has been shown to give good performance over lifetimes exceeding 400 days in some cases. The temperature coefficients and response to oxygen level changes are in good agreement with thermodynamic predictions. The effect of running the meters in high-oxygen sodium has been studied and a general mode of failure has been shown to be grain-boundary attack by oxygen/sodium solutions. The effect of #betta#-radiation on the meters has been studied. The meters with a metal/metal oxide reference electrode were unaffected by dose rates up to 52860 mGy h - 1 . Meters with an air reference electrode do show an effect as a voltage reduction at levels down to 2420 mGy h - 1 . This effect was temperature-dependent and was insignificant at 500 0 C. (orig.)

Generation IV SFR Nuclear Reactors: Under-Sodium Repair for ASTRID

International Nuclear Information System (INIS)

Baque, F.; Chagnot, C.; Bruguiere, L.; Augem, J.M.; Delalande, V.; Sibilo, J.

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. In-pile examination or repair requires robotic carriers. These carriers have to be compatible with the sodium environment: either in the cover-gas plenum or in gas after sodium draining, or even under liquid sodium. This R and D programme has been divided into nine parts in order to provide an overall design of the required robotic carriers and to develop technological solutions for their components: detailed definition for SFR carrier needs (access to internal structures, possible defects to be detected/repaired), definition and specifications of carrier architecture (depending on inspection and repair scenarios), in-sodium leak-tightness of carrier components, carrier material compatibility with sodium, temperature resistance (200 deg. C), irradiation resistance (depending on the location of the main vessel), gas-tight bell for operations under liquid sodium, carrier positioning control in liquid sodium, development, validation and qualification of technological solutions, for future SFRs, and worldwide benchmark regarding the previous areas of investigation. (authors)

ORNL facilities for testing first-wall components

International Nuclear Information System (INIS)

Tsai, C.C.; Becraft, W.R.; Gardner, W.L.; Haselton, H.H.; Hoffman, D.J.; Menon, M.M.; Stirling, W.L.

1985-01-01

Future long-impulse magnetic fusion devices will have operating characteristics similar to those described in the design studies of the Tokamak Fusion Core Experiment (TFCX), the Fusion Engineering Device (FED), and the International Tokamak Reactor (INTOR). Their first-wall components (pumped limiters, divertor plates, and rf waveguide launchers with Faraday shields) will be subjected to intense bombardment by energetic particles exhausted from the plasma, including fusion products. These particles are expected to have particle energies of approx.100 eV, particle fluxes of approx.10 18 cm -2 .s -1 , and heat fluxes of approx.1 kW/cm 2 CW to approx.100 kW/cm 2 transient. No components are available to simultaneously handle these particle and heat fluxes, survive the resulting sputtering erosion, and remove exhaust gas without degrading plasma quality. Critical issues for research and development of first-wall components have been identified in the INTOR Activity. Test facilities are needed to qualify candidate materials and develop components. At Oak Ridge National Laboratory (ORNL), existing neutral beam and wave heating test facilities can be modified to simulate first-wall environments with heat fluxes up to 30 kW/cm 2 , particle fluxes of approx.10 18 cm -2 .s -1 , and pulse lengths up to 30 s, within test volumes up to approx.100 L. The characteristics of these test facilities are described, with particular attention to the areas of particle flux, heat flux, particle energy, pulse length, and duty cycle, and the potential applications of these facilities for first-wall component development are discussed

A Small-Scale Capsule Test for Investigating the Sodium-Carbon Dioxide Reaction

International Nuclear Information System (INIS)

Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H.

2007-01-01

The utilization of modular sodium-to-supercritical CO 2 heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO 2 heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO 2 in a sodium pool, causing a pressurization which may threaten the structural integrity of the heat exchanger. Available data seems to indicate that the chemical reaction between sodium and CO 2 would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO 2 reaction is virtually non-existent

Design Evaluation of Thermal-hydraulic Test Facility for a Finned-tube Sodium-to-Air Heat Exchanger

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyungmo; Kim, Byeong-Yeon; Ko, Yung Joo; Cho, Youngil; Kim, Jong-Man; Son, Seok-Kwon; Jo, Youngchul; Kang, Byeong Su; Jung, Minhwan; Eoh, Jaehyuk; Lee, Hyeong-Yeon; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

This paper introduces the recent progress of overall design phase for the SELFA facility and deals with basic thermal-hydraulic design parameters and its design validation as well. For more reliable design of the safety-grade decay heat removal system (DHRS) in PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), two kinds of sodium-to-air heat exchangers have been employed in the system as an ultimate heat sink. One is a natural draft sodium-to-air heat exchanger (AHX) with helically-coiled sodium tubes, and the other is a forced draft sodium-to-air heat exchanger (FHX) with finned tubes with a straight-type arranged. Since the FHX is normally operated in an active mode with a forced air draft conditions, its performance should be verified for any anticipated operating conditions. To validate the test section design, evaluations of both thermal-hydraulic and mechanical design have been carried out, and some new concepts or devices were newly employed to replicate the prototypic conditions as closely as possible.

SUBMERGED GRAVEL SCRUBBER DEMONSTRATION AS A PASSIVE AIR CLEANER FOR CONTAINMENT VENTING AND PURGING WITH SODIUM AEROSOLS -- CSTF TESTS AC7 - AC10

Energy Technology Data Exchange (ETDEWEB)

HILLIARD, R K.; MCCORMACK, J D.; POSTMA, A K.

1981-11-01

Four large-scale air cleaning tests (AC7 - AC10) were performed in the Containment Systems Test Facility (CS'lF) to demonstrate the performance of a Submerged Gravel Scrubber for cleaning the effluent gas from a vented and purged breeder reactor containment vessel. The test article, comprised of a Submerged Gravel Scrubber (SGS) followed by a high efficiency fiber demister, had a design gas flow rate of 0.47 m{sup 3}/s (1000 ft{sup 3}/min) at a pressure drop of 9.0 kPa (36 in. H{sub 2}O). The test aerosol was sodium oxide, sodium hydroxide, or sodium carbonate generated in the 850-m{sup 3} CSTF vessel by continuously spraying sodium into the air-filled vessel while adding steam or carbon dioxide. Approximately 4500 kg (10,000 lb) of sodium was sprayed over a total period of 100 h during the tests. The SGS/Demister system was shown to be highly efficient (removing ~99.98% of the entering sodium aerosol mass), had a high mass loading capacity, and operated in a passive manner, with no electrical requirement. Models for predicting aerosol capture, gas cooling, and pressure drop are developed and compared with experimental results.

Peroxide formation and kinetics of sodium dissolution in alcohols

International Nuclear Information System (INIS)

Muralidaran, P.; Chandran, K.; Ganesan, V.; Periaswami, G.

1997-01-01

Suitable techniques for sodium removal and decontamination of sodium wetted components of Liquid Metal Fast Reactors (LMFRs) are necessary both for repair, reuse and decommissioning of such components. Among the methods followed for sodium removal, alcohol dissolution is usually employed for small components like bellow sealed valves, gripping tools to handle core components and sodium sampling devices (primary and secondary). One of the concerns in the alcohol dissolution method is the possible role of peroxide formation in the ethoxy group during storage and handling leading to explosion. This paper describes the study of peroxide formation in ethyl carbitol and butyl cellosolve as well as some of the results of dissolution kinetic studies carried out in our laboratory using different alcohols. The peroxide formation of ethyl carbitol and butyl cellosolve were studied by iodometric technique. It has been found that the peroxide formation is less in sodium containing alcohol than in pure one. Ethyl carbitol, butyl cellosolve and Jaysol-SS (mixture of ethyl alcohol, methyl alcohol, isopropyl alcohol and methyl isobutyl ketone) were used in dissolution kinetics studies. The effects due to area and orientation of the fresh sodium surface have also been investigated. The reaction rates were studied in the temperature range of 303-343 K. The rate of dissolution was estimated by measuring the sodium content of alcohol at periodic intervals. It is found that the reaction rate varies in the order of ethyl alcohol-water mixture > Jaysol-SS > butyl cellosolve > ethyl carbitol. While cleaning sodium using alcohol, the concentration of alcohol is held essentially constant throughout the process. The rate of reaction depends only on the amount of sodium and follows pseudo-first order kinetics. Increase in surface area has a marked impact on the dissolution rate at lower temperatures while at higher temperatures, the temperature factor overrides the effect due to surface area

High power rf component testing for the NLC

International Nuclear Information System (INIS)

Vlieks, A.E.; Fowkes, W.R.; Loewen, R.J.; Tantawi, S.G.

1998-09-01

In the Next Linear Collider (NLC), the high power rf components must be capable of handling peak rf power levels in excess of 600 MW. In the current view of the NLC, even the rectangular waveguide components must transmit at least 300 MW rf power. At this power level, peak rf fields can greatly exceed 100 MV/m. The authors present recent results of high power tests performed at the Accelerator Structure Test Area (ASTA) at SLAC. These tests are designed to investigate the rf breakdown limits of several new components potentially useful for the NLC. In particular, the authors tested a new TE 01 --TE 10 circular to rectangular wrap-around mode converter, a modified (internal fin) Magic Tee hybrid, and an upgraded flower petal mode converter

Sodium removal from CSRDM lower part by water vapour - CO2 process

International Nuclear Information System (INIS)

Sundar Raj, S.I.; Sreedhar, B.K.; Gurumoorthy, K.; Rajan, K.K.; Kalyanasundaram, P.; Rajan, M.; Vaidyanathan, G.

2006-01-01

Sodium is the primary and secondary coolant in fast reactors. Primary and secondary circuits components like Control and Safety Rod Drive Mechanism (CSRDM), pumps, heat exchangers etc. handle liquid sodium. Sodium has good affinity to oxygen and reacts vigorously with water. Hence sodium cleaning is the first and important activity in the maintenance of the components. In reactor components this cleaning process also helps in removing a major part of radioactive contaminants after which they are subjected to chemical decontamination. There are several methods available for removing sodium from components. Out of these, the water vapour-CO 2 process is selected for large components such as pumps, heat exchangers etc. while steam cleaning is used for the core sub assemblies. The cleaning processes are to be closely monitored to ensure safety because the release of hydrogen is to be kept below 4 % during the process. This paper discusses the in house facility and the experience in the successful use of the process in the cleaning of CSRDM. (author)

EBR-II experience with sodium cleaning and radioactivity decontamination

International Nuclear Information System (INIS)

Ruther, W.E.; Smith, C.R.F.

1978-01-01

The EBR-II is now in Its 13th year of operation. During that period more than 2400 subassemblies have been cleaned of sodium without a serious incident of any kind by a two-step process developed at Argonne. Sodium cleaning and decontamination of other reactor components has been performed only on the relatively few occasions in which a repair or replacement has been required. A summary of the EBR-II experience will be presented. A new facility will be described for the improved cleaning and maintenance of sodium-wetted primary components

EBR-II experience with sodium cleaning and radioactivity decontamination

Energy Technology Data Exchange (ETDEWEB)

Ruther, W E; Smith, C R.F. [Argonne National Laboratory, Argonne (United States)

1978-08-01

The EBR-II is now in Its 13th year of operation. During that period more than 2400 subassemblies have been cleaned of sodium without a serious incident of any kind by a two-step process developed at Argonne. Sodium cleaning and decontamination of other reactor components has been performed only on the relatively few occasions in which a repair or replacement has been required. A summary of the EBR-II experience will be presented. A new facility will be described for the improved cleaning and maintenance of sodium-wetted primary components.

Influence of dynamic sodium environment on the creep-fatigue behaviour of Modified 9Cr-1Mo ferritic-martensitic steel

International Nuclear Information System (INIS)

Kannan, R.; Ganesan, V.; Mariappan, K.; Sukumaran, G.; Sandhya, R.; Mathew, M.D.; Bhanu Sankara Rao, K.

2011-01-01

Highlights: → The effects of dynamic sodium on the CFI behaviour of Mod. 9Cr-1Mo steel has investigated. → The cyclic stress response of Mod. 9Cr-1Mo steel under flowing sodium environment is similar to that of air environment. → The creep-fatigue endurance of the alloy is found to decrease with introduction of hold time and with increase in the duration of hold time and the factor of life increase in sodium compared to air environment is reduced with increase in hold time. → In contrast to air environment, tensile holds were found to be more damaging than compression hold in sodium environment. → Design rules based on air environment can be safely applied for the components operating in sodium environment. - Abstract: The use of liquid sodium as a heat transfer medium for sodium-cooled fast reactors (SFRs) necessitates a clear understanding of the effects of dynamic sodium on low cycle fatigue (LCF), creep and creep-fatigue interaction (CFI) behaviour of reactor structural materials. Mod. 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of sodium cooled fast reactors. The steam generator has a design life of 30-40 years. The effects of dynamic sodium on the LCF and CFI behaviour of Mod. 9Cr-1Mo steel have been investigated at 823 and 873 K. The CFI life of the steel showed marginal increase under flowing sodium environment when compared to air environment. Hence, the design rules for creep-fatigue interaction based on air tests can be safely applied for components operating in sodium environment. This paper attempts to explain the observed LCF and CFI results based on the detailed metallography and fractography conducted on the failed samples.

Sodium Is Not Essential for High Bioactivity of Glasses

Science.gov (United States)

Chen, Xiaojing; Chen, Xiaohui; Brauer, Delia S.; Wilson, Rory M.; Law, Robert V.; Hill, Robert G.; Karpukhina, Natalia

2017-01-01

This study aims to demonstrate that excellent bioactivity of glass can be achieved without the presence of an alkali metal component in glass composition. In vitro bioactivity of two sodium-free glasses based on the quaternary system SiO2-P2O5-CaO-CaF2 with 0 and 4.5 mol% CaF2 content was investigated and compared with the sodium containing glasses with equivalent amount of CaF2. The formation of apatite after immersion in Tris buffer was followed by Fourier transform infrared spectroscopy (FTIR), X-ray diffraction (XRD), 31P and 19F solid state MAS-NMR. The dissolution study was completed by ion release measurements in Tris buffer. The results show that sodium free bioactive glasses formed apatite at 3 hours of immersion in Tris buffer, which is as fast as the corresponding sodium containing composition. This signifies that sodium is not an essential component in bioactive glasses and it is possible to make equally degradable bioactive glasses with or without sodium. The results presented here also emphasize the central role of the glass compositions design which is based on understanding of structural role of components and/or predicting the network connectivity of glasses. PMID:29271977

Impact Of Sodium Oxalate, Sodium Aluminosilicate, and Gibbsite/Boehmite on ARP Filter Performance

Energy Technology Data Exchange (ETDEWEB)

Poirier, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-11-01

The Savannah River Site (SRS) is currently treating radioactive liquid waste with the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction Unit (MCU). Recently, the low filter flux through the ARP of approximately 5 gallons per minute has limited the rate at which radioactive liquid waste can be treated. Salt Batch 6 had a lower processing rate and required frequent filter cleaning. Savannah River Remediation (SRR) has a desire to understand the causes of the low filter flux and to increase ARP/MCU throughput. SRR requested SRNL to conduct bench-scale filter tests to evaluate whether sodium oxalate, sodium aluminosilicate, or aluminum solids (i.e., gibbsite and boehmite) could be the cause of excessive fouling of the crossflow or secondary filter at ARP. The authors conducted the tests by preparing slurries containing 6.6 M sodium Salt Batch 6 supernate, 2.5 g MST/L slurry, and varying concentrations of sodium oxalate, sodium aluminosilicate, and aluminum solids, processing the slurry through a bench-scale filter unit that contains a crossflow primary filter and a dead-end secondary filter, and measuring filter flux and transmembrane pressure as a function of time. Among the conclusions drwn from this work are the following: (1) All of the tests showed some evidence of fouling the secondary filter. This fouling could be from fine particles passing through the crossflow filter. (2) The sodium oxalate-containing feeds behaved differently from the sodium aluminosilicate- and gibbsite/boehmite-containing feeds.

THORS: a high-temperature sodium test facility rated at 2.0 MW

International Nuclear Information System (INIS)

Gnadt, P.A.; Anderson, A.H.; Clapp, N.E.; Montgomery, B.H.; Collins, C.W.; Stulting, R.D.

1979-01-01

The Thermal--Hydraulic Out-of-Reactor Safety (THORS) facility at Oak Ridge Naitonal Laboratory (ORNL) is a high-temperature sodium test facility operated for the United States Breeder Reactor Safety Program. The facility is primarily used for testing large simulated Liquid-Metal Fast Breeder Reactor (LMFBR) fuel subassemblies. The facility has recently been upgraded to provide a 2.0-MW test bundle power input and heat removal capability. A new test section, which will be capable of operating at 980 0 C and which will accommodate a 217-pin bundle, has also been added. A 61-pin bundle is currently under test in the facility. A description of the test facility is presented, along with a brief summary of the 8-year operating history of this safety-related test facility

A Small-Scale Capsule Test for Investigating the Sodium-Carbon Dioxide Reaction

Energy Technology Data Exchange (ETDEWEB)

Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H

2007-01-15

The utilization of modular sodium-to-supercritical CO{sub 2} heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO{sub 2} heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO{sub 2} in a sodium pool, causing a pressurization which may threaten the structural integrity of the heat exchanger. Available data seems to indicate that the chemical reaction between sodium and CO{sub 2} would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO{sub 2} reaction is virtually non-existent.

Safety problems encountered in construction and operation of the sodium test facilities of the Institute of Reactor Development (IRD) at the Karlsruhe Nuclear Research Center

International Nuclear Information System (INIS)

Schleisiek, K.

1971-01-01

In this report the safety aspects of the design and construction of a sodium boiling loop and a sodium tank test facility are discussed. Subsequently two experiments concerning the safety of the facilities are described: the testing of a drip basin to collect the sodium and to limit the rate of burning in the case of a leak, and the investigation of the chemical reaction of sodium with the insulating materials. Finally some general emergency procedures in the case of sodium incidents are discussed. A 16 mm-film demonstrating sodium fires and fire fighting methods will be shown. (author)

Test facility for the evaluation of microwave transmission components

International Nuclear Information System (INIS)

Fong, C.G.; Poole, B.R.

1985-01-01

A Low Power Test Facility (LPTF) was developed to evaluate the performance of Electron Cyclotron Resonance Heating (ECRH) microwave transmission components for the Mirror Fusion Test Facility (MFTF-B). The facility generates 26 to 60 GHz in modes of TE 01 , TE 02 , or TE 03 launched at power levels of 1/2 milliwatt. The propagation of the rf as it radiates from either transmitting or secondary reflecting microwave transmission components is recorded by a discriminating crystal detector mechanically manipulated at constant radius in spherical coordinates. The facility is used to test, calibrate, and verify the design of overmoded, circular waveguide components, quasi-optical reflecting elements before high power use. The test facility consists of microwave sources and metering components, such as VSWR, power and frequency meters, a rectangular TE 10 to circular TE 01 mode transducer, mode filter, circular TE 01 to 2.5 in. diameter overmoded waveguide with mode converters for combination of TE 01 to TE 03 modes. This assembly then connects to a circular waveguide launcher or the waveguide component under test

Shaking table qualification tests of mechanical and electrical components

International Nuclear Information System (INIS)

Jurukovski, D.

1993-01-01

This presentation covers the experience of the Institute of Earthquake Engineering and Engineering Seismology, Skopje, Republic of Macedonia in seismic qualification of mechanical components by shaking table testing. The characteristics of the biaxial seismic and single component shaking tables used at the Institute are given. Some examples of the experience from performed test for reactor components are included

Sodium waste technology: A summary report

International Nuclear Information System (INIS)

Abrams, C.S.; Witbeck, L.C.

1987-01-01

The Sodium Waste Technology (SWT) Program was established to resolve long-standing issues regarding disposal of sodium-bearing waste and equipment. Comprehensive SWT research programs investigated a variety of approaches for either removing sodium from sodium-bearing items, or disposal of items containing sodium residuals. The most successful of these programs was the design, test, and the production operation of the Sodium Process Demonstration Facility at ANL-W. The technology used was a series of melt-drain-evaporate operations to remove nonradioactive sodium from sodium-bearing items and then converting the sodium to storable compounds

Current status of sodium fire and aerosol research in Japan

International Nuclear Information System (INIS)

Himeno, Yoshiaki

1989-01-01

State-of-the-art of the research and development related to sodium fire and aerosol behaviour is presented. This paper covers the Japanese work on sodium leak, leak detector, sodium oxidation and combustion, sodium aerosol release, fire mitigation, reliabilities of the electrical instruments and the reactor components under the sodium aerosols suspended atmosphere, aerosol plugging in a leak path, and the computer codes are presented. (author)

Investigation of sodium area conflagrations and testing of a protective system

Energy Technology Data Exchange (ETDEWEB)

Huber, F; Menzenhauer, P; Peppler, W [Kernforschungszentrum Karlsruhe (F.R. Germany). Inst. fuer Reaktorentwicklung

1975-12-01

During research and development work on the SNR-300 sodium-cooled fast reactor the consequences and confinement of sodium fires occurring in enclosures were studied. The behavior of liquid sodium during fires and the behavior of an inherently ready-for-operation protective system are described. Theoretical considerations on the behavior of burning liquid sodium are compared with experimental results. A protective system for large facilities is presented and the use of extinguishing powders is reviewed.

A stronger patch test elicitation reaction to the allergen hydroxycitronellal plus the irritant sodium lauryl sulfate

DEFF Research Database (Denmark)

Heydorn, S; Andersen, K E; Johansen, J D

2003-01-01

Household and cleaning products often contain both allergens and irritants. The aim of this double-blinded, randomized, paired study was to determine whether patch testing with an allergen (hydroxycitronellal) combined with an irritant [sodium lauryl sulfate (SLS)] cause a stronger patch test...

Non-destructive testing of electronic component packages

International Nuclear Information System (INIS)

Anderle, C.

1975-01-01

A non-destructive method of investigating packaged parts of semiconductor components by X radiation is described and the relevant theoretical relations limiting this technique are derived. The application of the technique is demonstrated in testing several components. The described method is iNsimple and quick. (author)

Sodium aerosol recovering device

International Nuclear Information System (INIS)

Fujimori, Koji; Ueda, Mitsuo; Tanaka, Kazuhisa.

1997-01-01

A main body of a recovering device is disposed in a sodium cooled reactor or a sodium cooled test device. Air containing sodium aerosol is sucked into the main body of the recovering device by a recycling fan and introduced to a multi-staged metal mesh filter portion. The air about against each of the metal mesh filters, and the sodium aerosol in the air is collected. The air having a reduced sodium aerosol concentration circulates passing through a recycling fan and pipelines to form a circulation air streams. Sodium aerosol deposited on each of the metal mesh filters is scraped off periodically by a scraper driving device to prevent clogging of each of the metal filters. (I.N.)

Combustion of sodium in the open atmosphere

Energy Technology Data Exchange (ETDEWEB)

Morewith, H A; Johnson, R P; Nelson, C T; Otter, J M [Energy System Group, Rockwell International, Rockwell (United States)

1979-03-01

A series of sodium fire tests has been conducted in ambient air at a meteorological test site. This test series was designed to simulate hypothetical accidents which might occur in the heat transport system of an LMFBR. Measurements of concentration, agglomeration, fallout, and chemical species of the sodium combustion products were made as a function of downwind distance. In each of the first two tests, {approx}23 kg of 540 deg. C sodium was sprayed as a fan of 250-{mu}m sodium drops across the wind, from heights of 5 or 6 m. Each release took a few minutes. A dense sodium combustion product aerosol was formed, and quickly agglomerated to large (100 to 660 {mu}m) diameter particles. More than 50% of the aerosol mass fell out within several hundred meters of the release point. Two additional tests were performed by releasing sodium through 9.5-mm diameter jets at a height of 30 m. In each test, the sodium jet was aimed horizontally across the wind, and followed a downward parabolic trajectory, releasing burning sodium drops along its track. Again, close-in fallout due to large agglomerates was observed. A substantial amount of unburned sodium fell 30 m to the ground, where it burned. In a third type of test, sodium was burned for 60 min as pool in a 1.5m{sup 2} burn pan at 9 m/s wind velocity. Approximately 30% of the combustion products became airborne. Large agglomerates fell out as they moved downwind, depositing 1 kg/m{sup 2} at 1 m downwind from the edge of the pan. Chemical analysis of the samples indicated that the sodium fires produced mainly Na{sub 2}O, and that the conversion of NaOH was slow. Comparison were made with COMRADEX-IV code models, which are appropriate for calculating deposition and concentrations for downwind distances between 10{sup 2} and 10{sup 4} m. (author)

Combustion of sodium in the open atmosphere

International Nuclear Information System (INIS)

Morewith, H.A.; Johnson, R.P.; Nelson, C.T.; Otter, J.M.

1979-01-01

A series of sodium fire tests has been conducted in ambient air at a meteorological test site. This test series was designed to simulate hypothetical accidents which might occur in the heat transport system of an LMFBR. Measurements of concentration, agglomeration, fallout, and chemical species of the sodium combustion products were made as a function of downwind distance. In each of the first two tests, ∼23 kg of 540 deg. C sodium was sprayed as a fan of 250-μm sodium drops across the wind, from heights of 5 or 6 m. Each release took a few minutes. A dense sodium combustion product aerosol was formed, and quickly agglomerated to large (100 to 660 μm) diameter particles. More than 50% of the aerosol mass fell out within several hundred meters of the release point. Two additional tests were performed by releasing sodium through 9.5-mm diameter jets at a height of 30 m. In each test, the sodium jet was aimed horizontally across the wind, and followed a downward parabolic trajectory, releasing burning sodium drops along its track. Again, close-in fallout due to large agglomerates was observed. A substantial amount of unburned sodium fell 30 m to the ground, where it burned. In a third type of test, sodium was burned for 60 min as pool in a 1.5m 2 burn pan at 9 m/s wind velocity. Approximately 30% of the combustion products became airborne. Large agglomerates fell out as they moved downwind, depositing 1 kg/m 2 at 1 m downwind from the edge of the pan. Chemical analysis of the samples indicated that the sodium fires produced mainly Na 2 O, and that the conversion of NaOH was slow. Comparison were made with COMRADEX-IV code models, which are appropriate for calculating deposition and concentrations for downwind distances between 10 2 and 10 4 m. (author)

Sodium-concrete reactions experiments and code development

International Nuclear Information System (INIS)

Casselman, C.; Malet, J.C.; Dufresne, J.; Bolvin, M.

1988-01-01

Hypothesis of hot sodium leak in a fast breeder reactor implies, for the safety organism to consider spillage of sodium on concrete. This safety analysis involves the understanding of sodium-concrete reactions, the knowledge of their consequences and to test the choiced preventive solutions. In association with EDF, the nuclear safety department had carried out an extensive experimental program, the different parts of which are connected with each aspect of this problem: - firstly, interaction between sodium and bare surface of usual concrete; - secondly, the case of a sodium spillage on a concrete surface covered with a defected liner; - thirdly, special concrete tests for a comparison with usual concrete behavior, in direct contact with hot sodium; - at last, a test which concerns a new design with a layer of the selected concrete protected with a defected liner. On the same time, theoretical work leads to elaborate a physical model to describe temporal evolution of thermal and chemical decomposition of a concrete slab under hot sodium action. SORBET-REBUS system will use quoted above test results to its validation

Radiochemical surveillance of KNK primary sodium

International Nuclear Information System (INIS)

Stamm, H.-H.; Stade, K.Ch.

1987-05-01

Radiochemical surveillance of the KNK primary sodium has been performed now for 15 years with 953 effective full-power days. The overflow method used for sodium sampling proved to be reliable. Different crucible materials have been used for different analytical tasks. The amount of radionuclides in the primary system has not given restrictions to plant operation at any time. On-line gamma spectroscopy on pipings and components of the primary circuits was accomplished in reactor downtimes. Activity depositions on the walls were dominated by Ta-182 after KNK I operation. Main deposited activities at KNK II were Mn-54 (fresh core) and after operation with failed fuel Cs-137, in cover gas areas together with Zn-65. Efficient experimental radionuclide traps for the removal of Mn-54, Zn-65 and Cs-137 from the primary coolant were tested successfully. The dose rates on primary pipes and components of KNK I and KNK II were lower by an order of magnitude compared to water-cooled reactors. This is in good agreement with experiences from LMFBR's in other countries. The resulting average yearly accumulated personal dose rate was 0.21 man-Sv at KNK, compared to 3.9 man-Sv at German light-water-cooled power reactors

Design Evaluation of a Piping System in the SELFA Sodium Test Facility

Energy Technology Data Exchange (ETDEWEB)

Son, Seok-Kwon; Jo, Young-Chul; Lee, Hyeong-Yeon; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, design evaluations on the SELFA piping system has been conducted according to the ASME B31.1 and RCC-MRx RD-3600. The conservatism of the two codes was quantified based on the evaluation results. It was shown that B31.1 was more conservative for the sustained loads while less conservative for thermal expansion loads when compare with those of RD-3600. However, all the evaluation results according to the two codes were within the code allowables. There are two main piping systems in the SELFA test loop. In this study, the integrity of the SELFA piping system has been evaluated according to the two design-by-rule (DBR) codes of ASME B31.1 and RCC-MRx RD-3600. B31.1 is an industry design code for power piping while RD-3600 is a class 3 nuclear DBR code. The conservatism of the two codes was quantified based on the evaluation results as per the two DBR codes. The sodium test facility of the SELFA is under construction at KAERI for the investigation of thermo-hydraulic behavior of finned-tube sodium-to-air heat exchanger.

Wastage-resistant characteristics of 12Cr steel tube material. Small leak sodium-water reaction test

International Nuclear Information System (INIS)

Shimoyama, Kazuhito

2004-03-01

In the water leak accident of a steam generator designed for a sodium cooled reactor in the Feasibility Study, the localization of tube failure propagation by using an advanced water leak detector will be required from the viewpoints of the safety and economical efficiency of the plant. So far, the conventional knowledge and analytical tools have been used in the investigation and evaluation of water leak phenomenon; nevertheless, there was neither test data nor the study of quantitative evaluation on the corrosion behavior, so-called wastage-resistant characteristics, of 12Cr steel tube material in sodium-water reactions. Wastage tests for the 12Cr steel tube material were conducted in small water leaks by use of the Sodium-Water Reaction Test Rig (SWAT-1R), and the data of wastage rate were obtained in the parameter of water leak rate under the constant sodium temperature and distance between leak and target tubes. The test results lead to the following conclusions: (1) The wastage-resistibility of 12Cr steel is 1.6 times greater than that of 9Cr steel and is 2.7 times greater than that of 2.25Cr-1Mo steel. (2)The wastage-resistibility of 12Cr steel increases in smaller water leaks; especially in water leak rates of 1 g/sec or less, it is more excellent than that of SUS321 stainless steel used as Monju superheater tube material. (3) Based on the correlation of wastage rate for the 9Cr steel, the correlation for the 12Cr steel has been obtained to be used for the evaluation of tube failure propagation. As the correlation of wastage rate for the 12Cr steel is based on the correlation for the 9Cr steel, it gives enough conservatism in smaller water leaks. To serve in accurately evaluating the tube failure propagation in smaller water leaks, it is necessary to obtain new correlation of wastage rate for the 12Cr steel based on the data in the wide range of water leak rates. (author)

Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-06-29

The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na₂O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO₂ (sCO₂) Brayton cycle power conversion system. In SFRs, Na₂O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na₂O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na₂O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na₂O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO₂ systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

Sodium fire protection

International Nuclear Information System (INIS)

Raju, C.; Kale, R.D.

1979-01-01

Results of experiments carried out with sodium fires to develop extinguishment techniques are presented. Characteristics, ignition temperature, heat evolution and other aspects of sodium fires are described. Out of the powders tested for extinguishment of 10 Kg sodium fires, sodium bi-carbonate based dry chemical powder has been found to be the best extinguisher followed by large sized vermiculite and then calcium carbonate powders distributed by spray nozzles. Powders, however, do not extinguish large fires effectively due to sodium-concrete reaction. To control large scale fires in a LMFBR, collection trays with protective cover have been found to cause oxygen starvation better than flooding with inert gas. This system has an added advantage in that there is no damage to the sodium facilities as has been in the case of powders which often contain chlorine compounds and cause stress corrosion cracking. (M.G.B.)

Sodium ionization detector and sensor

International Nuclear Information System (INIS)

Hrizo, J.; Bauerle, J.E.

1979-01-01

Work conducted on a basic technology development effort with the Westinghouse Sodium Ionization Detector (SID) sensor is reported. Included are results obtained for three task areas: (1) On-line operational response testing - in-situ calibration techniques; (2) Performance-reliability characteristics of aged filaments; and (3) Evaluation of chemical interference effects. The results showed that a calibrator filament coated with a sodium compound, when activated, does supply the necessary sodium atoms to provide a valid operational in-situ test. The life time of new Cr 2 0 3 -protected SID sensor filaments can be extended by operating at a reduced temperature. However, there also is a reduction in the sensitivity. Non-sodium species, such as products from a smoldering fire and organic aerosols, produce an interference response from the sensor comparable to a typical sodium response

Estimation of component failure probability from masked binomial system testing data

International Nuclear Information System (INIS)

Tan Zhibin

2005-01-01

The component failure probability estimates from analysis of binomial system testing data are very useful because they reflect the operational failure probability of components in the field which is similar to the test environment. In practice, this type of analysis is often confounded by the problem of data masking: the status of tested components is unknown. Methods in considering this type of uncertainty are usually computationally intensive and not practical to solve the problem for complex systems. In this paper, we consider masked binomial system testing data and develop a probabilistic model to efficiently estimate component failure probabilities. In the model, all system tests are classified into test categories based on component coverage. Component coverage of test categories is modeled by a bipartite graph. Test category failure probabilities conditional on the status of covered components are defined. An EM algorithm to estimate component failure probabilities is developed based on a simple but powerful concept: equivalent failures and tests. By simulation we not only demonstrate the convergence and accuracy of the algorithm but also show that the probabilistic model is capable of analyzing systems in series, parallel and any other user defined structures. A case study illustrates an application in test case prioritization

Sodium experiments in support of PFBR design safety

International Nuclear Information System (INIS)

Rajan, M.; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Kale, R.D.

2002-01-01

Full text: The pool type PFBR uses sodium as coolant. It has a roof slab of about 16 metres in diameter with large rotating plug, small rotating plug and control plug. The plugs rotate to orient. the fuel handling canal above the core and the operating and control instruments are supported in the control plug. The roof slab is maintained at 120 degC. The sodium temperature is 550 degC when the reactor is in operation. The axial temperature difference between the sodium and roof slab leads to intense convection of argon cover gas in the annular space of the rotating plugs causing large circumferential temperature difference and unwanted thermal stresses in the plugs and other components supported on roof slab. Experiments were carried out simulating reactor conditions in sodium facilities to control the circumferential temperature difference. The argon cover gas above sodium (550 degC) is at a temperature of about 350 degC. The sodium vapour generated, condenses as sodium mist in the cover gas. Due to convection currents the sodium mist is carried and deposited in cooler areas. The process creates problem for components like CSRDM which has long vertical annular gaps with a large axial temperature gradient. Experiments were carried out to reduce the convection currents inside the CSRDM annular space. A multiple sleeve was developed which reduces convection and hence the sodium deposition in the annular space

Dynamic leaching test of personal computer components.

Science.gov (United States)

Li, Yadong; Richardson, Jay B; Niu, Xiaojun; Jackson, Ollie J; Laster, Jeremy D; Walker, Aaron K

2009-11-15

A dynamic leaching test (DLT) was developed and used to evaluate the leaching of toxic substances for electronic waste in the environment. The major components in personal computers (PCs) including motherboards, hard disc drives, floppy disc drives, and compact disc drives were tested. The tests lasted for 2 years for motherboards and 1.5 year for the disc drives. The extraction fluids for the standard toxicity characteristic leaching procedure (TCLP) and synthetic precipitation leaching procedure (SPLP) were used as the DLT leaching solutions. A total of 18 elements including Ag, Al, As, Au, Ba, Be, Cd, Cr, Cu, Fe, Ga, Ni, Pd, Pb, Sb, Se, Sn, and Zn were analyzed in the DLT leachates. Only Al, Cu, Fe, Ni, Pb, and Zn were commonly found in the DLT leachates of the PC components. Their leaching levels were much higher in TCLP extraction fluid than in SPLP extraction fluid. The toxic heavy metal Pb was found to continuously leach out of the components over the entire test periods. The cumulative amounts of Pb leached out of the motherboards in TCLP extraction fluid reached 2.0 g per motherboard over the 2-year test period, and that in SPLP extraction fluid were 75-90% less. The leaching rates or levels of Pb were largely affected by the content of galvanized steel in the PC components. The higher was the steel content, the lower the Pb leaching rate would be. The findings suggest that the obsolete PCs disposed of in landfills or discarded in the environment continuously release Pb for years when subjected to landfill leachate or rains.

Conceptual design report, Sodium Storage Facility, Fast Flux Test Facility, Project F-031

International Nuclear Information System (INIS)

Shank, D.R.

1995-01-01

The Sodium Storage Facility Conceptual Design Report provides conceptual design for construction of a new facility for storage of the 260,000 gallons of sodium presently in the FFTF plant. The facility will accept the molten sodium transferred from the FFTF sodium systems, and store the sodium in a solid state under an inert cover gas until such time as a Sodium Reaction Facility is available for final disposal of the sodium

Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

International Nuclear Information System (INIS)

Tomabechi, K.

1975-01-01

The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

Japanese position paper on sodium-water reaction testing and design

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Tanabe, H.; Miyake, O.; Kuroha, M.; Hoshi, Y.

1984-01-01

PNC has been developing the steam generator with helically coiled heat transfer tube bundle and downcommer tubes for the prototype fast reactor Monju since 1968. To establish the safety design against the sodium-water reaction accident was one of the most important R and D items at the start of the development. PNC started the experimental study initially in the large leak region in 1970. Until now, during twelve years, the experimental studies have been performed, which covers the phenomena from a micro leak to a large one, with the use of the SWAT-1 rig, SWAT-2 loop, SWAT-3 loop, and SWAT-4 rigs. The reliable leak detection system is necessary to minimize the damage by the sodium-water reaction. Two groups of efforts have been paid for developing the detection system. One is to develop the leak detector itself, and another is to grasp the hydrogen transport behavior in the sodium in the steam generator and the secondary piping system. Four sodium loops have been used for the development. The development of computer codes has also progressed in parallel with the sodium-water reaction experiments. Three codes have been accomplished for the design tools against the sodium-water reaction. Through the efforts mentioned above, sufficient experiences were obtained for designing and operating the Monju steam generator system

Abrasion Testing of Critical Components of Hydrokinetic Devices

Energy Technology Data Exchange (ETDEWEB)

Worthington, Monty [ORPC Alaska; Ali, Muhammad [Ohio University; Ravens, Tom [University of Alaska Anchorage

2013-12-06

The objective of the Abrasion Testing of Critical Components of Hydrokinetic Devices (Project) was to test critical components of hydrokinetic devices in waters with high levels of suspended sediment – information that is widely applicable to the hydrokinetic industry. Tidal and river sites in Alaska typically have high suspended sediment concentrations. High suspended sediment also occurs in major rivers and estuaries throughout the world and throughout high latitude locations where glacial inputs introduce silt into water bodies. In assessing the vulnerability of technology components to sediment induced abrasion, one of the greatest concerns is the impact that the sediment may have on device components such as bearings and seals, failures of which could lead to both efficiency loss and catastrophic system failures.

Acceptance test for graphite components and construction status of HTTR

International Nuclear Information System (INIS)

Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

Assessment of sodium status in large ruminants by measuring the sodium-to-potassium ratio in muzzle secretions.

Science.gov (United States)

Singh, S P; Rani, D

1999-09-01

To develop a simple diagnostic test to assess sodium status in large ruminants on the basis of the sodium-to-potassium ratio (Na:K) and to determine its relevance. 7 buffalo heifers and 21 lactating, pregnant, and nonpregnant dairy cows and heifers. Buffalo heifers were subjected in 2 experiments to variable dietary sodium intake or sodium depletion and changes in sodium and potassium concentrations; Na:K was simultaneously monitored in various body fluids to study its value for indicating sodium status. Validity of the muzzle secretion test was assessed. Muzzle secretion and urinary Na:K and sodium concentration, but not serum electrolyte concentrations, reflected the sodium status of buffalo heifers in response to the widely variable intake of sodium (0.03 to 0.16% of dry matter [DM]). Progressive sodium depletion during an 11-day period, using saliva deprivation caused reciprocal changes in sodium and potassium concentrations in saliva and muzzle secretion, but not in urine. Decreasing urine sodium concentration was associated with decreasing urine potassium concentration. Saliva, urine, and muzzle secretion Na:K closely reflected the degree of sodium deficit. Buffaloes or dairy cows maintained on optimal sodium intake had muzzle secretion and urine Na:K > 0.30. Muzzle secretion or urine Na:K muzzle secretion Na:K, and to a large extent urine Na:K, may be used as a convenient diagnostic tool to assess sodium status in large ruminants. It has accuracy similar to that of saliva Na:K.

Sodium vapour deposition studies related to the development of components for LMFBR in Japan

Energy Technology Data Exchange (ETDEWEB)

Himeno, Y; Yamamoto, K; Takahashi, J; Mochizuki, K; Saito, R [O-arai Engineering Centre, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai Machi, Ibaraki-ken (Japan)

1977-01-01

During the course of research and development for the experimental fast breeder reactor JOYO, performance tests for such components as rotating plug and vapor trap have been carried out using prototype models. In recent years, as the development effort has been concentrated to the prototype fast reactor MONJU, performance tests for the components mentioned above in large scale are now in progress. In this report, brief description of the test results and experiences obtained from these studies are presented, with the results of basic studies related to these. (author)

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)

Monitoring sodium in commercially processed foods from stores and restaurants

Science.gov (United States)

Most of the sodium we eat comes from commercially processed foods from stores and restaurants. Sodium reduction in these foods is a key component of several recent public health efforts. Agricultural Research Service (ARS) of USDA, CDC and FDA have launched a collaborative program to monitor sodium ...

Cavitation erosion scaling: tests on a pump impeller in water and in sodium

International Nuclear Information System (INIS)

Dorey, J.M.; Rascalou, T.

1992-01-01

Tests to quantify cavitation agressivity carried out in water and in sodium (400 deg) on a model pump impeller are presented. The polished samples method has been used. It can be now applied to curved surfaces such as impeller blades with the help of new measurement devices. Results are discussed regarding scaling laws for fluid-to-fluid transposition

Water simulation of sodium reactors

International Nuclear Information System (INIS)

Grewal, S.S.; Gluekler, E.L.

1981-01-01

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

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)

Large scale sodium interactions. Part 3. Chemical phenomena with limestone concrete

International Nuclear Information System (INIS)

Sallach, R.A.

1977-01-01

The description of the chemical processes and reaction products resulting from the exposure of concrete to molten sodium metal is important for a thorough, realistic assessment of the safety of CRBR-type reactors. Concretes are in general complex heterogenous substances whose ingredients can be derived from many sources. Consequently a wide variety of reaction processes and products might be anticipated. Initial attention has focused on a concrete in which both the aggregate and sandy components are derived from limestone. Presented are the chemical observations and experimental data from tests in which molten sodium metal at approximately 500 0 C is dropped into cold limestone concrete crucibles. Thermocouples immersed in the sodium pool indicate that the reaction proceeds in two stages. In the first stage which lasts 5 to 8 minutes, the temperature of the reacting mass hovers around 500 0 C. This stage is followed by a second stage of longer duration--greater than 100 minutes--where the temperature is 700 to 800 0 C. The main reaction product is a hard, fused, black slag which contains about 3/4 of the sodium in the initial charge. A secondary product is sodium oxide aerosol which accounts for the remaining 1/4 of the charge. It is significant that no free sodium metal is found in the slag; all sodium has completely reacted

Design, fabrication, and testing of a sodium evaporator for the STM4-120 kinematic Stirling engine

Energy Technology Data Exchange (ETDEWEB)

Rawlinson, K.S.; Adkins, D.R.

1995-05-01

This report describes the development and testing of a compact heat-pipe heat exchanger kW(e) designed to transfer thermal energy from hot combustion gases to the heater tubes of a 25-kW(e) Stirling engine. In this system, sodium evaporates from a surface that is heated by a stream of hot gases. The liquid metal then condenses on the heater tubes of a Stirling engine, where energy is transferred to the engine`s helium working fluid. Tests on a prototype unit illustrated that a compact (8 cm {times} 13 cm {times} 16 cm) sodium evaporator can routinely transfer 15 kW(t) of energy at an operating vapor temperature of 760 C. Four of these prototype units were eventually used to power a 25-kW(e) Stirling engine system. Design details and test results from the prototype unit are presented in this report.

Some techniques for sodium removal in CIAE

International Nuclear Information System (INIS)

Yuan Waimai; Ding Dejun; Guo Huanfang; Hong Shuzhang; Zhou Shuxia; Shen Fenyang; Yang Zhongmin; Xu Yongxing

1997-01-01

In this paper the experiment and application on sodium removal and sodium disposal are presented. Steam-nitrogen process was used in CIAE for cleaning cold traps, sodium vapor traps, a sodium tank. Atomized water-nitrogen process was used for cleaning dummy fuel assembly for CEFR and a sintered stainless steel filter. Sprinkle process was used for cleaning some tubes. Bultylcellosolve was used for cleaning sintered stainless steel filter and sodium flow measurement device. Ethanol alcohol was used for cleaning electromagnetic pump. Paraffin, transformer-oil or their mixture was used for cleaning sodium valves, a sodium vapor trap and sodium-potassium alloy absorber. A small sintered stainless steel filter was distillated in vacuum. A simple sodium disposal device has been served for several years in CIA.E. It can dispose about 10 Kg sodium each time and the disposal process is no-aerosol. It operates in open air for non-radioactive sodium. In recent years a small sodium cleaning plant has been built. It can use atomized water, steam or organic alcohol to removal of sodium. The LAVEL cleaning plant and SLAPSO cleaning plant were introduced from Italy. And CEFR preliminary design on sodium cleaning for spent fuel assembly and on sodium removal-decontamination for large reactor components is introduced. Vapour-nitrogen process is planned to use in them. (author)

Sodium modulates opioid receptors through a membrane component different from G-proteins. Demonstration by target size analysis

International Nuclear Information System (INIS)

Ott, S.; Costa, T.; Herz, A.

1988-01-01

The target size for opioid receptor binding was studied after manipulations known to affect the interactions between receptor and GTP-binding regulatory proteins (G-proteins). Addition of GTP or its analogs to the binding reaction, exposure of intact cells to pertussis toxin prior to irradiation, or treatment of irradiated membranes with N-ethylmaleimide did not change the target size (approximately equal to 100 kDa) for opioid receptors in NG 108-15 cells and rat brain. These data suggest that the 100-kDa species does not include an active subunit of a G-protein or alternatively that GTP does not promote the dissociation of the receptor-G-protein complex. The presence of Na+ (100 mM) in the radioligand binding assay induced a biphasic decay curve for agonist binding and a flattening of the monoexponential decay curve for a partial agonist. In both cases the effect was explained by an irradiation-induced loss of the low affinity state of the opioid receptor produced by the addition of Na+. This suggests that an allosteric inhibitor that mediates the effect of sodium on the receptor is destroyed at low doses of irradiation, leaving receptors which are no longer regulated by sodium. The effect of Na+ on target size was slightly increased by the simultaneous addition of GTP but was not altered by pertussis toxin treatment. Thus, the sodium unit is distinct from G-proteins and may represent a new component of the opioid receptor complex. Assuming a simple bimolecular model of one Na+ unit/receptor, the size of this inhibitor can be measured as 168 kDa

High Sodium Simulant Testing To Support SB8 Sludge Preparation

International Nuclear Information System (INIS)

Newell, J. D.

2012-01-01

Scoping studies were completed for high sodium simulant SRAT/SME cycles to determine any impact to CPC processing. Two SRAT/SME cycles were performed with simulant having sodium supernate concentration of 1.9M at 130% and 100% of the Koopman Minimum Acid requirement. Both of these failed to meet DWPF processing objectives related to nitrite destruction and hydrogen generation. Another set of SRAT/SME cycles were performed with simulant having a sodium supernate concentration of 1.6M at 130%, 125%, 110%, and 100% of the Koopman Minimum Acid requirement. Only the run at 110% met DWPF processing objectives. Neither simulant had a stoichiometric factor window of 30% between nitrite destruction and excessive hydrogen generation. Based on the 2M-110 results it was anticipated that the 2.5M stoichiometric window for processing would likely be smaller than from 110-130%, since it appeared that it would be necessary to increase the KMA factor by at least 10% above the minimum calculated requirement to achieve nitrite destruction due to the high oxalate content. The 2.5M-130 run exceeded the DWPF hydrogen limits in both the SRAT and SME cycle. Therefore, testing of this wash endpoint was halted. This wash endpoint with this minimum acid requirement and mercury-noble metal concentration profile appears to be something DWPF should not process due to an overly narrow window of stoichiometry. The 2M case was potentially processable in DWPF, but modifications would likely be needed in DWPF such as occasionally accepting SRAT batches with undestroyed nitrite for further acid addition and reprocessing, running near the bottom of the as yet ill-defined window of allowable stoichiometric factors, potentially extending the SRAT cycle to burn off unreacted formic acid before transferring to the SME cycle, and eliminating formic acid additions in the frit slurry

NUCLEBRAS' installations for tests of nuclear power plants components

International Nuclear Information System (INIS)

Vasconcelos Paiva, I.P. de; Horta, J.A.L.; Avelar Esteves, F. de; Pinheiro, R.B.

1983-05-01

The reasons for NUCLEBRAS' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufacturers, giving to them the means for performing functional tests of industrial products, are presented. A brief description of the facilities under construction: the Components Test Loop and the Facility for Testing N.P.P. Components under Accident Conditions, and of other already in operation, is given, as well as its objectives and main technical characteristics. Some test results already obtained are also presented. (Author) [pt

Immunogenicity test of tetanus component in adsorbed vaccines by toxin binding inhibition test

Directory of Open Access Journals (Sweden)

Denise Cristina Souza Matos

2002-09-01

Full Text Available Samples from 20 lots of diphtheria-tetanus (adult use dT vaccine and from 20 lots of diphtheria-tetanus-pertussis (DTP vaccine were used to standardize and validate the in vitro toxin binding inhibition (ToBI test for the immunogenicity test of the tetanus component. The levels of tetanus antitoxin obtained by ToBI test were compared to those obtained using the toxin neutralization (TN test in mice routinely employed to perform the quality control of the tetanus component in adsorbed vaccines. The results ranged from 1.8 to 3.5 IU/ml for dT and 2 to 4 IU/ml for DTP by ToBI test and 1.4 to 3 IU/ml for dT and 1.8 to 3.5 IU/ml for DTP by TN in mice. These results were significantly correlated. From this study, it is concluded that the ToBI test is an alternative to the in vivo neutralization procedure in the immunogenicity test of the tetanus component in adsorbed vaccines. A substantial refinement and a reduction in use of animals can be achieved.

Operating experience of steam generator test facility

International Nuclear Information System (INIS)

Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

2006-01-01

Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

Tolerance Test of Eisenia Fetida for Sodium Chloride

Energy Technology Data Exchange (ETDEWEB)

Kerr, M.; Stewart, A.J.

2003-01-01

Saltwater spills that make soil excessively saline often occur at petroleum exploration and production (E&P) sites and are ecologically damaging. Brine scars appear when produced water from an E&P site is spilled onto surrounding soil, causing loss of vegetation and subsequent soil erosion. Revegetating lands damaged by brine water can be difficult. The research reported here considers earthworms as a bioremedial treatment for increasing the salt mobility in this soil and encouraging plant growth and a healthy balance of soil nutrients. To determine the practical application of earthworms to remediate brine-contaminated soil, a 17-d test was conducted to establish salt tolerance levels for the common compost earthworm (Eisenia fetida) and relate those levels to soil salinity at brine-spill sites. Soil samples were amended with sodium chloride in concentrations ranging from 1 to 15 g/kg, which represent contamination levels at some spill sites. The survival rate of the earthworms was near 90% in all tested concentrations. Also, reproduction was noted in a number of the lower-concentration test replicates but absent above the 3-g/kg concentrations. Information gathered in this investigation can be used as reference in further studies of the tolerance of earthworms to salty soils, as results suggest that E. fetida is a good candidate to enhance remediation at brine-damaged sites.

Data acquisition and control for LMFBR component testing

International Nuclear Information System (INIS)

Turner, G.E.

1983-01-01

Liquid Metal Fast Breeder Reactor components such as pumps, steam generators, and piping components are tested for their ability to withstand expected thermal transients of up to 25 0 F/s (14 0 C/s). The Energy Technology Engineering Center performs these tests in facilities specifically designed for that purpose. Although much of the instrumentation and controls for these test facilities are similar to those used in conventional process plants, the requirement to produce, control, and measure the effects of rapid thermal transients results in some not-so-conventional data acquisition and control system design criteria. This paper describes a typical data-acquisition system used at one of the ETEC test facilities and how the thermal transients are produced and controlled in the pump and steam-generator test facilities

Radioactive sodium waste treatment and conditioning. Review of main aspects

International Nuclear Information System (INIS)

2007-01-01

This publication reviews the main aspects relating to the treatment and conditioning of radioactive sodium waste. This waste arises from the operation of liquid metal fast reactors (LMFRs). In this type of reactor, sodium (Na) or sodium-potassium alloys (NaK) are used as a low-effect neutron moderating coolant medium for extracting and transferring thermal energy from the core and they represent a significant technical and safety challenge during operation and decommissioning. This publication provides the reader with technologically oriented information on the present status of sodium waste management approaches and recent achievements related to treatment and conditioning, with the objective of facilitating planning and preparatory work for the decommissioning of LMFRs. This publication provides a comprehensive review of the hazards associated with sodium waste management. Given the large quantities of sodium waste arising during decommissioning or reactor refurbishment, as well as the challenges and varied techniques associated with removal of 100% of all sodium and NaK bulk quantities and residues during decommissioning, a hazards review and analysis is a critical component in planning the dismantling and waste management activities. Roughly half of this publication focuses on sodium waste generating, handling and treatment processes. This includes draining sodium and NaK from plant systems; in situ treatment of residual sodium; cutting techniques for pumps, valves, piping and other components; cleaning of components; potential reuse of sodium; and removal of selected radionuclides from sodium waste with the objective of reducing the waste classification or converting it to exempt waste. The focus is on proven techniques and technologies, and each discussed method includes a review of the associated principle or theory, practical applications, advantages and disadvantages, limitations, industry experience, and final waste products. A review is provided of final

Development of bellows for sodium valves in Japan

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, S; Mukai, K; Fukada, T; Takahashi, T

1980-02-01

Sodium valves for FBR are required to isolate sodium side from the atmosphere completely throughout its lifetime because of preventing sodium leakage with or without radioactivity. A great number of sodium valves have been used in FBR test facilities at O-arai Engineering Center of PNC and many troubles have occurred through their operational experience. Most of the cause of the troubles are the bellows failure followed by sodium leakage. A research and development program on bellows was started to clarify many uncertain factors of its performance and to establish the feasibility of bellows used in sodium. In this program Small Bellows Test Loop was built to perform low cycle fatigue tests on bellows under high temperature conditions. In this report some examples of the investigation of failed bellows occurred at O-arai Engineering Center of PNC are described. The research and development program on bellows is also explained with the summary of recent test results. (author)

Preliminary conceptual design of the secondary sodium circuit-eliminated JSFR (Japan Sodium Fast Reactor) adopting a supercritical CO2 turbine system (1). Sodium/CO2 heat exchanger

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Sakamoto, Yoshihiko; Kotake, Shoji

2014-09-01

Research and development of the supercritical CO 2 (S-CO 2 ) cycle turbine system is underway in various countries for further improvement of the safety and economy of sodium-cooled fast reactors. The Component Design and Balance-Of-Plant (CD and BOP) of the Generation IV International Nuclear Forum (Gen-IV) has addressed this study, and their analytical and experimental results have been discussed between the relevant countries. JAEA, who is a member of the CD and BOP, has performed a design study of an S-CO 2 gas turbine system applied to the Japan Sodium-cooled Fast Reactor (JSFR). In this study, the S-CO 2 cycle turbine system was directly connected to the primary sodium system of the JSFR to eliminate the secondary sodium circuit, aiming for further economical improvement. This is because there is no risk of sodium-water reaction in the S-CO 2 cycle turbine system of SFRs. The Na/CO 2 heat exchanger is one of the key components for the secondary sodium system eliminated SFR, and this report describes its structure and the safety in case of CO 2 leak. A Printed Circuit Heat Exchanger (PCHE), which has a greater heat transfer performance, is employed to the heat exchanger. Another advantage of the PCHE is to limit the area affected by a leak of CO 2 because of its partitioned flow path structure. A SiC/SiC ceramic composite material is used for the PCHE to prevent crack growth and to reduce thermal stress. The Na/CO 2 heat exchanger has been designed in such a way that a number of small heat transfer modules are combined in the vessel in consideration of manufacture and repair. The primary sodium pump is installed in the center of the heat exchanger vessel. CO 2 leak events in the heat exchanger have been also evaluated, and it revealed that no significant effect has arisen on the core or the primary sodium boundary. (author)

Effect of liquid sodium on long-term properties of austenitic steels

International Nuclear Information System (INIS)

Svoboda, V.; Merta, J.; Slach, J.

The effect is discussed of liquid sodium on the long-term properties of austenitic steels corresponding to the ASI 304 and ASI 316 types, mainly of steel CSN 17348. The choice is described of test specimens and of the experimental sodium test equipment. Testing was carried out using the so-called indirect method, i.e., the liquid sodium effect was assessed using the results of creep tests of two groups of specimens, one exposed to sodium and the other to the inert argon atmosphere. Otherwise the tests proceeded under identical conditions. The sodium stand had been manufactured for exposure of test specimens to liquid sodium. The morphology of specimen surfaces was studied by the JSN-50A electron microscope. The results of testing steel CSN 17348-AKV EXTRA S exposed to liquid sodium containing 10 ppm of oxygen at a temperature of 550 degC showed a significant sodium effect on the basic mechanical properties, on long-term creep strength and on the metallurgical properties. (Oy)

Detection of oscillatory components in noise signals and its application to fast detection of sodium boiling in LMFBR's

International Nuclear Information System (INIS)

Ehrhardt, J.

1975-09-01

In general, the surveillance of technical plants is performed by observating the mean value of measured signals. In this method not all information included in these signals is used. On the other hand - for example in a reactor - disturbances are possible which generate small oscillatory components in the measured signals. In general, these oscillatory components do not influence the mean value of the signals and consequently do not activate the conventional control system; however they can be found by analysis of the signal's noise component. For the detection of these oscillatory signals the observation of the frequency spectra of the noise signals is particularly advantageous because they produce peaks at the oscillation frequencies. In this paper a new detection system for the fast detection of suddenly appearing peaks in the frequency spectra of noise signals is presented. The prototype of a compact detection unit was developed which continuously computes the power spectral density (PSD) of noise signals and simultaneously supervises the PSD for peaks in the relevant frequency range. The detection method is not affected by the frequency dependance of the PSD and is applicable to any noise signal. General criteria were developed to enable the determination of the optimal detection system and its sensitivity. The upper limits of false alarm rate and detection time were taken into account. The detection criteria are applicable to all noise signals with approximately normally distributed amplitudes. Theoretical results were confirmed in a number of experiments; special experimental and theoretical parameter studies were done for the optimal detection of sodium boiling in LMFBR's. Computations based on these results showed that local and integral sodium boiling can be detected in a wide core range of SNR 300 by observing fluctuations of the neutron flux. In this connection it is important to point out that no additional core instrumentation is necessary because the

Testing of ceramic gas turbine components under service-like conditions

Energy Technology Data Exchange (ETDEWEB)

Siebmanns, W [Motoren- und Turbinen-Union G.m.b.H., Muenchen (Germany, F.R.)

1978-08-01

If all gas turbine components which are in contact with hot gas are manufactured from special ceramics (silicon nitride, silicon carbide), cycle and component temperatures can be increased up to 1600/sup 0/K. MTU is developing various components, such as combustor and turbine wheel, step by step until they are ready for service. At present, combustors are surviving comprehensive service-like cyclic tests in hot gas at atmospheric pressure (1000 h, 1000 starts per component) without damage. Tests above atmospheric pressure (5 bar) are underway. At MTU, a rotor wheel variant consisting of a metallic hub with inserted single blades is being constructed. The step to aerodynamically contoured airfoils will follow, as soon as the stress problems encountered in connection with the blade root are fully under control. The program will be completed in 1980 with a test run of a prototype turbine made from ceramic components developed by various companies under the leadership of the DFVLR (Aerospace Research and Testing Institute).

Laboratory performance testing of an extruded bitumen containing a surrogate, sodium nitrate-based, low-level aqueous waste

International Nuclear Information System (INIS)

Mattus, A.J.; Kaczmarsky, M.M.

1986-01-01

Laboratory results of a comprehensive, regulatory performance test program, utilizing an extruded bitumen and a surrogate, sodium nitrate-based waste, have been compiled at the Oak Ridge National Laboratory (ORNL). Using a 53 millimeter, Werner and Pfleiderer extruder, operated by personnel of WasteChem Corporation of Paramus, New Jersey, laboratory-scale, molded samples of type three, air blown bitumen were prepared for laboratory performance testing. A surrogate, low-level, mixed liquid waste, formulated to represent an actual on-site waste at ORNL, containing about 30 wt % sodium nitrate, in addition to eight heavy metals, cold cesium and strontium was utilized. Samples tested contained three levels of waste loading: that is, forty, fifty and sixty wt % salt. Performance test results include the ninety day ANS 16.1 leach test, with leach indices reported for all cations and anions, in addition to the EP Toxicity test, at all levels of waste loading. Additionally, test results presented also include the unconfined compressive strength and surface morphology utilizing scanning electron microscopy. Data presented include correlations between waste form loading and test results, in addition to their relationship to regulatory performance requirements

Corrosion performance of advanced structural materials in sodium.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L. (Nuclear Engineering Division)

2012-05-16

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux

Corrosion performance of advanced structural materials in sodium

International Nuclear Information System (INIS)

Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L.

2012-01-01

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and

Operational experience on sodium deposits in KNK reactor and RSB test facility

Energy Technology Data Exchange (ETDEWEB)

Jansing, W; Kirchner, G; Menck, J [INTERATOM, Bergisch Gladbach (Germany)

1977-01-01

A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

UK position paper on sodium fires

Energy Technology Data Exchange (ETDEWEB)

Vaughan, G J [National Nuclear Corporation Ltd., Risley, Warrington, Cheshire (United Kingdom); Glass, D [United Kingdom Atomic Energy Authority, Dounreay Nuclear Power Development Establishment, Thurso, Caithness (United Kingdom); Newman, R N [Central Electricity Generating Board, Berekely Nuclear Laboratory, Berkeley, Gloucestershire (United Kingdom); Ramsdale, S A [United Kingdom Atomic Energy Authority, Safety and Reliability Directorate, Culcheth, Cheshire (United Kingdom); Snelling, K W [United Kingdom Atomic Energy Authority, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1989-07-01

The UK has over several years developed a philosophy for the prevention, mitigation and extinguishment of sodium fires. The systems which were developed for PFR have been continuously revised and modified and from these considerations systems were proposed for CDFR. The latest phases of this development are described with reference to the CDFR plant. The current analytical and experimental work on fires, aerosols and sodium concrete reactions is also discussed. The UK are developing codes to analyse the effects of a sodium fire in a building and to model aerosol behaviour following a fire. Experimental work on small scale fires, aerosol behaviour, filtration devices and sodium concrete reaction is being carried out on a laboratory scale. Techniques for aerosol measurement and characterisation have also been developed and used both In the laboratory and large scale tests. Larger scale tests of sodium fire extinguishment techniques have also been performed. Currently a programme of tests (SOFA) of large scale fires in the open to investigate the chemical and physical changes in the aerosol and its dispersion in the atmosphere are just beginning. The UK studies are intended to both assist in the development of prevention and mitigation systems for design base and beyond design base accidents in any building which contains sodium (or sodium potassium alloy) and also to provide methods for assessing the risks from such accidents. (author)

UK position paper on sodium fires

International Nuclear Information System (INIS)

Vaughan, G.J.; Glass, D.; Newman, R.N.; Ramsdale, S.A.; Snelling, K.W.

1989-01-01

The UK has over several years developed a philosophy for the prevention, mitigation and extinguishment of sodium fires. The systems which were developed for PFR have been continuously revised and modified and from these considerations systems were proposed for CDFR. The latest phases of this development are described with reference to the CDFR plant. The current analytical and experimental work on fires, aerosols and sodium concrete reactions is also discussed. The UK are developing codes to analyse the effects of a sodium fire in a building and to model aerosol behaviour following a fire. Experimental work on small scale fires, aerosol behaviour, filtration devices and sodium concrete reaction is being carried out on a laboratory scale. Techniques for aerosol measurement and characterisation have also been developed and used both In the laboratory and large scale tests. Larger scale tests of sodium fire extinguishment techniques have also been performed. Currently a programme of tests (SOFA) of large scale fires in the open to investigate the chemical and physical changes in the aerosol and its dispersion in the atmosphere are just beginning. The UK studies are intended to both assist in the development of prevention and mitigation systems for design base and beyond design base accidents in any building which contains sodium (or sodium potassium alloy) and also to provide methods for assessing the risks from such accidents. (author)

Reactor component automatic grapple

International Nuclear Information System (INIS)

Greenaway, P.R.

1982-01-01

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment. (author)

Sodium heat pipe module test for the SAFE-30 reactor prototype

International Nuclear Information System (INIS)

Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

2001-01-01

Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

Summary of sodium removal and decontamination programs in the USA

International Nuclear Information System (INIS)

Steele, O.P.; Brehm, W.F.

1978-01-01

The goals of the United States Department of Energy sodium removal and decontamination programs are: 1) to identify plant requirements and develop safe effective processes for sodium removal, both for experimental facilities and for reactor components; 2) to develop effective decontamination processes for removing deposited radioactivity; 3) to establish and put into use a set of workable criteria for requalification and return to service of components after sodium removal and decontamination; 4) to design, build and operate facilities in which to perform the sodium removal and decontamination operations. This paper gives a summary of progress toward each of these goals; details will be presented in other papers at this meeting. Three basic processes for sodium removal have been investigated extensively: the use of water vapor in an inert carrier gas, the use of an alcohol-type reagent, and evaporation. The process development work on the first two processes has been essentially completed. The evaporative process is still under development, but preliminary parameters have been established

Improvement and test calculation on basic code or sodium-water reaction jet

Energy Technology Data Exchange (ETDEWEB)

Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

1999-03-01

In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

Improvement and test calculation on basic code or sodium-water reaction jet

International Nuclear Information System (INIS)

Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo

1999-03-01

In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

In service inspection and repair of sodium cooled ASTRID prototype

Energy Technology Data Exchange (ETDEWEB)

Baque, F.; Jadot, F. [French Atomic Commission, Cadarache Centre, 13108 Saint Paul lez Durance Cedex, (France); Marlier, R. [AREVA, 10 rue Recamier, 69456 Lyon cedex 06, (France); Saillant, J-F. [AREVA/NDE Solutions, 4 rue Thomas Dumorey, BP 70385, 71109 Chalon sur Saone Cedex, (France); Delalande, V. [EDF R and D, 6, quai Watier, 78400 Chatou, (France)

2015-07-01

In the frame of the large R and D work which is performed for the future ASTRID sodium cooled prototype, In Service Inspection and Repair (ISI and R) has been identified as a major issue to be taken into account in order to enlarge the plant safety, to consolidate its availability and to protect the associated investment. After the first part of pre-conceptual design phase (2008-2012), the running second part of pre-conceptual phase (2013-2015) allows to increase the ISI and R tool ability for immersed sodium structures of ASTRID, at about 200 deg. C, on the basis of consolidated specifications and thanks to their qualification through more and more realistic laboratory tests and simulation with CIVA code. ISI and R items are being developed and qualified during a pluri-annual program which mainly deals with the reactor block structures, the primary components and circuit, and the Power Conversion System. It ensures a strong connection between the reactor designers and inspection specialists, as the optimization of inspectability and repairability is looked at: this already induced specific rules for design, in order to shorten and ease the ISI and R operations, which have been merged into RCC-MRx rules. In the frame of increasing technology readiness level with corresponding performance demonstration, this paper presents R and D dealing with the ISI and R items: it highlights the sensor development (both ultrasonic and electromagnetic concepts, compatible with sodium at 200 deg. C), then their applications for ASTRID structure control (under sodium telemetry, imaging and NDE). Activity for repair is also presented (a single laser tool for sodium sweeping, machining and welding), and finally the effort for associated robotic (generic program for ASTRID applications, specific technological tools for sodium medium, tight immersed bell). The main results of testing and simulation are given for telemetry, vision, NDE applications, laser process repair and under sodium

Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

International Nuclear Information System (INIS)

Iizawa, K.; Sekiguchi, N.; Atsumo, H.

1976-01-01

The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

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.

Thermal hydraulic studies in steam generator test facility

International Nuclear Information System (INIS)

Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

2005-01-01

Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

Effect of carbohydrate or sodium bicarbonate ingestion on performance during a validated basketball simulation test.

Science.gov (United States)

Afman, Gregg; Garside, Richard M; Dinan, Neal; Gant, Nicholas; Betts, James A; Williams, Clyde

2014-12-01

Current recommendations for nutritional interventions in basketball are largely extrapolated from laboratory-based studies that are not sport-specific. We therefore adapted and validated a basketball simulation test relative to competitive basketball games using well-trained basketball players (n = 10), then employed this test to evaluate the effects of two common preexercise nutritional interventions on basketball-specific physical and skilled performance. Specifically, in a randomized and counterbalanced order, participants ingested solutions providing either 75 g carbohydrate (sucrose) 45 min before exercise (Study A; n = 10) or 2 × 0.2 g · kg(-1) sodium bicarbonate (NaHCO3) 90 and 20 min before exercise (Study B; n = 7), each relative to appropriate placebos (H2O and 2 × 0.14 g · kg(-1) NaCl, respectively). Heart rate, sweat rate, pedometer count, and perceived exertion did not systematically differ between the 60-min basketball simulation test and competitive basketball, with a strong positive correlation in heart rate response (r = .9, p basketball simulation test provides a valid reflection of physiological demands in competitive basketball and is sufficiently sensitive to detect meaningful changes in physical and skilled performance. While there are benefits of preexercise carbohydrate or sodium bicarbonate ingestion, these should be balanced against potential negative side effects.

Sodium components cleaning status in the Italian fast reactor program

Energy Technology Data Exchange (ETDEWEB)

De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

1978-08-01

As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

Testing on air cleaning systems: Testing of the components in-place tests

International Nuclear Information System (INIS)

Billard, F.; Brion, J.

1967-01-01

The reliability of air cleaning systems is dependent on testing they are submitted to. Although in-place tests are the most important as they act as final tests upon achieved plants, component tests are necessary too. They allow detection of defective units before they are installed, partition of unit defects from mounting defects and they are more sensitive. For similar reasons, material teats are most useful. The various tests are described, about aerosol filters for one part, iodine trap for the other. The checked features are: materials nature, units sizes, efficiency, air resistance, flammability, humidity resistance, temperature resistance, adsorbent friability, etc... On iodine trapping systems, small check traps, working by-pass with the main trap are periodically subjected to efficiency test. This control allow to cut down the in-place tests frequency, particularly when poisoning from organic vapours is to be feared. (authors) [fr

Sodium reflux pool-boiler solar receiver on-sun test results

Energy Technology Data Exchange (ETDEWEB)

Andraka, C E; Moreno, J B; Diver, R B; Moss, T A [Oak Ridge National Lab., TN (United States)

1992-06-01

The efficient operation of a Stirling engine requires the application of a high heat flux to the relatively small area occupied by the heater head tubes. Previous attempts to couple solar energy to Stirling engines generally involved directly illuminating the heater head tubes with concentrated sunlight. In this study, operation of a 75-kW{sub t} sodium reflux pool-boiler solar receiver has been demonstrated and its performance characterized on Sandia's nominal 75-kW{sub t} parabolic-dish concentrator, using a cold-water gas-gap calorimeter to simulate Stirling engine operation. The pool boiler (and more generally liquid-metal reflux receivers) supplies heat to the engine in the form of latent heat released from condensation of the metal vapor on the heater head tubes. The advantages of the pool boiler include uniform tube temperature, leading to longer life and higher temperature available to the engine, and decoupling of the design of the solar absorber from the engine heater head. The two-phase system allows high input thermal flux, reducing the receiver size and losses, therefore improving system efficiency. The receiver thermal efficiency was about 90% when operated at full power and 800{degree}C. Stable sodium boiling was promoted by the addition of 35 equally spaced artificial cavities in the wetted absorber surface. High incipient boiling superheats following cloud transients were suppressed passively by the addition of small amounts of xenon gas to the receiver volume. Stable boiling without excessive incipient boiling superheats was observed under all operating conditions. The receiver developed a leak during performance evaluation, terminating the testing after accumulating about 50 hours on sun. The receiver design is reported here along with test results including transient operations, steady-state performance evaluation, operation at various temperatures, infrared thermography, x-ray studies of the boiling behavior, and a postmortem analysis.

PG BN 1600 sodium fire protection system

International Nuclear Information System (INIS)

Bar, J.; Urbancik, L.

1978-12-01

A design was developed of a fire protection system for steam generator of a 1600 MW sodium cooled fast reactor (BN-1600). Chemical reactions are described of liquid sodium with atmospheric components and solid materials coming into contact with sodium in its release from the steam generator, and in safeguarding protection against sodium fires. The requirements for the purity of nitrogen as an atmosphere inert to liquid sodium are given. Characteristics and basic parameters are shown of level and spray fires, elementary terms are explained concerning the properties of aerosols formed during fires, the methods and means of release signalling and fire alarm are described as are fire precautions using fire-fighting equipment, modifying the support tank and the cell bottom and building sewage pits. The design of the system comprises an alarm system for liquid sodium using point and line electric contact sensors and flame photometer based aerosol sensors as well as a fire-fighting system based on the system of channelling liquid sodium into emergency discharge tanks filled with an inert gas, a set of fire extinguishers and other fire fighting material, and measures for the elimination of sodium fire consequences. (J.B.)

Environmental qualification testing of TFE valve components

International Nuclear Information System (INIS)

Eyvindson, A.; Krasinski, W.; McCutcheon, R.

1997-01-01

Valves containing tetrafluoroethylene (TFE) components are being used in many CANDU Nuclear Generating Stations. However, some concerns remain about the performance of TFE after exposure to high levels of radiation. Stations must therefore ensure that such valves perform reliably after being exposed to postulated accident radiation dose levels. The current Ontario Hydro Environmental Qualification [EQ] program specifies much higher postulated radiation exposure than the original design, to account for conditions following a LOCA. Initial assessments indicated that Teflon components would require replacement. Proof of acceptable performance can remove the need for large scale replacement, avoiding a significant cost penalty and preserving benefits due to the superior performance of TFE-based seals. A test program was undertaken at Chalk River Laboratories (CRL) to investigate the performance of three valves after irradiation to 10 Mrad. Such valves are currently used at the Bruce B Nuclear Generating Station. Each contains TFE packing rings; one also has TFE seats. Two of the valves are used in the ECIS recovery system, while the third is used for instrumentation loop isolation or as drain valves. All are exposed to little or no radiation during normal use. Based on the results of the tests, all the valves tested will still meet functional and performance requirements after the TFE components have been exposed to 10 Mrad of irradiation. (author)

Some aspects of sodium technology issued from the operating experience of RAPSODIE and PHENIX

International Nuclear Information System (INIS)

Kremser, J.; Lacroix, A.

1976-01-01

This paper deals with the experience on sodium technology gained from RAPSODIE and PHENIX operation. Problems encountered with sodium circuits, main components, handling equipment and instrumentation are discussed. Some information related to the contamination by radioactive products of sodium and sodium circuits is given

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

Instrumentation for Sodium Circuits; Instrumentation des Circuits de Sodium

Energy Technology Data Exchange (ETDEWEB)

Cambillard, E. [CEA, Centre d' Etudes Nucleaires de Fontenay-aux-Roses (France); Lions, N. [CEA, Centre d' Etudes Nucleaires de Cadarache (France)

1967-06-15

Electromagnetic flow meters, level gauges and differential pressure gauges are among the main measurement instruments designed and tested at the Commissariat a l'Energie Atomique (CEA) for sodium reactors. The main characteristics of the flow meters used with RAPSODIE are indicated. The instruments used in this connection are of the permanent -magnet or electromagnet type (in the primary circuits). A description is given of the calibration methods employed - use is made of diaphragms or Venturi tubes as standard flow meters - and information is given on the results measured for maximum sodium flows of 400 m{sup 3}/h. Three types of continuous level gauge have been studied. Resistance gauge. Two varieties used for the 1 - and 10-MW test circuits of RAPSODIE are described. In one there is a compensation resistance along the whole height of the measuring element (the continuous gauges used with the RAPSODIE reactor are at present of this type). In the other type of gauge a device is incorporated to heat the measurement element and prevent the formation of conducting deposits (prototype sodium tests have been completed). Induction gauge. This type has two coupled coils and is fitted with a device to compensate for temperature effects. A description is given of a prototype which has been built and the results obtained in the course of sodium tests are described. Ultrasonic gauge. With this type, a transmitter is fitted on top of the outside of the sodium container; there is also a vertical wave guide, the bottom of which is immersed in the liquid metal and possesses a reflector system which returns the ultrasonic beam towards the surface. Fixed reference marks provide a permanent means of calibration and the whole apparatus is welded. This type of gauge is now being constructed. The differential pressure gauges that have been built, and used in particular with Venturi tube flow meters, are modified versions of the devices employed with the 1 - and 10-MW test circuits of

Nuclebras' installations for performance tests of nuclear power plants components

International Nuclear Information System (INIS)

Vasconcelos Paiva, I.P. de; Avelar Esteves, F. de; Horta, J.A.L.; Resende, M.F.R.; Pinheiro, R.B.

1984-01-01

The reasons for Nuclebras' Nuclear Technology Development Center to implement a laboratory for supporting Brazilian manufactures, giving to them the means for performing functional tests of industrial products, are presented. A brief description of facilities under construction: the components Test Loop and Facility for Testing N.P.P. components under Accident conditions, and other already in operation, as well as its objectives and main technical characteristics. Some test results had already obtained are also presented. (Author) [pt

Sodium Carbonate is Saltier Than Sodium Chloride to Sodium-Depleted Rats.

Science.gov (United States)

St John, Steven J; McBrayer, Anya M; Krauskopf, Erin E

2017-10-01

In a series of behavioral experiments in the 1960s, G.R. Morrison identified several unique features of the taste of Na2CO3 to rats; namely, it is 1) considerably more intense than NaCl at isomolar concentrations, 2) avoided at 10 times lower concentrations than NaCl to thirsty rats, 3) preferred at 10 times lower concentrations than NaCl in sodium-depleted rats. He also demonstrated its qualitatively similarity to NaCl. In Experiment 1, we confirmed and extended many of Morrison's observations. Rats were injected with furosemide on 3 occasions to stimulate a sodium appetite. After each depletion, rats were given a brief-access taste test in a lickometer presenting, in random order, water and 7 concentrations of salt. One test used NaCl (0.028-0.89 M, quarter log steps), another used Na2CO3, and the third used Na2CO3, but at a tenfold lower concentration range (0.0028-0.089 M). Rats licked NaCl in an inverted-U shaped concentration-response function peaking at 0.158-0.281 M. As Morrison's results predicted, rats licked Na2CO3 in nearly identical fashion, but at a tenfold lower concentration range (peak at 0.0158-0.028 M). In a second experiment, furosemide-treated rats were repeatedly tested with the lower Na2CO3 range but mixed in the epithelial sodium channel blocker amiloride at various concentrations (3-300 μM, half log steps). Amiloride reduced licking for Na2CO3 and shifted the peak response rightward up to about half a log unit. Thus, this "super-saltiness" of Na2CO3 to rats is at least partly amiloride-dependent. © The Author 2017. Published by Oxford University Press. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Filtration of sodium-fire aerosols

International Nuclear Information System (INIS)

Alexas, A.; Jordan, S.; Lindner, W.

1979-01-01

Different filter devices have been developed and tested with respect to their use in the off-gas system of liquid-metal fast breeder reactors to prevent the escape of sodium-fire aerosols that might be formed in case of an accident. The testing results have shown that the use of a multilayer sand bed filter is still the best method to filter limited amounts of sodium-fire aerosols over a long operating time. Efficiencies on the order of 99.98 and 98.8% were reached for loading capacities of 500 and 1000 g/m 2 , respectively. Unlimited amounts of sodium-fire aerosols can be filtered by wet scrubbers with an efficiency of 70% per scrubber stage. Fiberglas filters connot be used for the filtration of sodium-fire aerosols over a long operating time because the filter material can be destroyed after several days of operating

Testing Header Component of Electricity Power Industry Boiler

International Nuclear Information System (INIS)

Soedardjo, S.A; Andryansyah, B; Artahari, Dewi; Natsir, Muhammad; Triyadi, Ari; Farokhi

2000-01-01

Testing of header component of Suralaya Unit II electricity power by replication method has been carried out. That header component is cross over pipe which interconnection between Primary and Superheater Outlet Header Secondary Superheater Outlet Header with the operation time over 14 years. The main composition of cross over pipe is 2 1/4 Cr 1 Mo or frequently specified as ferritique steel. The replication testing shown that the damage classification on those cross over pipe in A class based on failure classification from Neubauer and Wedel. Simple calculation in favor of cross over pipe remaining lifetime is about 16.5 years moreover

The design and testing of sodium pumps in the UK to meet the CDFR cavitation criteria

International Nuclear Information System (INIS)

Preece, G.E.; Macleod, I.D.; Wilkinson, D.

2002-01-01

A primary objective in the design of the sodium pumps for the UK Commercial Demonstration Fast Reactor has been to avoid cavitation during normal operation. This requirement arises from the need to avoid blade erosion and also, in the case of the Primary Sodium Pumps (PSP), the generation of cavitation noise which might otherwise interfere with instrumentation installed to detect noise of boiling in the core. This paper outlines the approach adopted to achieve a pump design with good cavitation performance and the programme of model testing carried out in a water loop to establish the cavitation boundaries for incipient cavitation of selected designs using both visual and acoustic techniques

Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

International Nuclear Information System (INIS)

Bolt, P.R.

1976-01-01

Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

Built-In Data-Flow Integration Testing in Large-Scale Component-Based Systems

Science.gov (United States)

Piel, Éric; Gonzalez-Sanchez, Alberto; Gross, Hans-Gerhard

Modern large-scale component-based applications and service ecosystems are built following a number of different component models and architectural styles, such as the data-flow architectural style. In this style, each building block receives data from a previous one in the flow and sends output data to other components. This organisation expresses information flows adequately, and also favours decoupling between the components, leading to easier maintenance and quicker evolution of the system. Integration testing is a major means to ensure the quality of large systems. Their size and complexity, together with the fact that they are developed and maintained by several stake holders, make Built-In Testing (BIT) an attractive approach to manage their integration testing. However, so far no technique has been proposed that combines BIT and data-flow integration testing. We have introduced the notion of a virtual component in order to realize such a combination. It permits to define the behaviour of several components assembled to process a flow of data, using BIT. Test-cases are defined in a way that they are simple to write and flexible to adapt. We present two implementations of our proposed virtual component integration testing technique, and we extend our previous proposal to detect and handle errors in the definition by the user. The evaluation of the virtual component testing approach suggests that more issues can be detected in systems with data-flows than through other integration testing approaches.

Friction characteristics of hardfacing materials in high temperature sodium

International Nuclear Information System (INIS)

Mizobuchi, Syotaro; Kano, Shigeki; Nakayama, Kohichi; Atsumo, Hideo

1980-01-01

Friction and self-welding test were conducted on several materials used for the contacting and sliding components of a sodium cooled fast breeder reactor. In the present study, the friction and self-welding characteristics of each material were evaluated through measuring the kinetic and breakaway friction coefficients. The influence of oscillating rotation and vertical reciprocating motion on the friction mode was also investigated. The results obtained are as follows: (1) Colmonoy No.6, the nickel base hardfacing alloy, indicated the lowest kinetic friction coefficient of all the materials in the present study. Also, Cr 3 C 2 /Ni-Cr material prepared by a detonation gun showed the most stable friction behavior. (2) The breakaway friction coefficient of each material was dependent upon dwelling time in a sodium environment. (3) The friction behavior of Cr 3 C 2 /Ni-Cr material was obviously related with the finishing roughness of the friction surface. It was anticipated that nichrome material as the binder of the chrome carbide diffused and exuded to the friction surface by sliding in sodium. (4) The friction coefficient in sliding mode of vertical reciprocating was lower than that of oscillating rotation. (author)

The sodium fire tests performed in the FAUNA facility on up to 12m2 fire areas

International Nuclear Information System (INIS)

Cherdron, W.; Jordan, S.

1983-08-01

The FAUNA test facility started operation in 1979. It serves to investigate large area sodium fires in closed containments and to study the generation, behaviour and removal of sodium fire aerosols. In this report, the experimental results of the 6 sodium pool fires are described which were performed with up to 500 kg of sodium in fire pans of 2 m 2 , 5 m 2 and 12 m 2 surface area, respectively. Both, the thermodynamic data and the data of the reaction kinetics of the fires were determined. In addition, the behaviour of the released aerosols during and after the fire was studied. On the basis of measurements of the temperature profiles at various levels above the fire areas it was shown that the convective flows above fire areas of different sizes in closed containments differ markedly and, obviously, exert an influence on the development of the fire and the release of particles. Whilst in rather small fires the gas above the pan rises as in a chimney and flows back on the walls, no chimney effect can be observed in a large pool fire. In rather large fires higher burning rates and aerosol release rates were observed. Some meters above the fire area temperatures around 300-400 0 C, temporarily even up to 700 0 C, were measured. The tests F5 and F6 were performed above all to observe the fire behaviour in terms of thermodynamics and reaction kinetics in a fully closed containment. (orig./RW) [de

PASLINK and dynamic outdoor testing of building components

NARCIS (Netherlands)

Baker, P.H.; Dijk, H.A.L. van

2008-01-01

The PASLINK test facilities and analysis procedures aim to obtain the thermal and solar characteristics of building components under real dynamic outdoor conditions. Both the analysis and the test methodology have evolved since the start of the PASSYS Project in 1985. A programme of upgrading the

Liquid sodium oxygenmeter

International Nuclear Information System (INIS)

Jakes, D.; Fresl, M.; Svoboda, V.

1979-02-01

The results of test runs of two design varieties of liquid sodium oxygenmeter in sodium loops are described. The accuracy and sensitivity are discussed reached using this instrument within 1 and 10 p.p.m. of oxygen concentration. A change in the used reference system is proposed based on practical experiences and thermochemical calculations. Ceramic electrolyte corrosion is analysed and the possible interpretation of the corrosion effect on the galvanic cell electromotive force is suggested. (author)

On the applicability of dye penetrant tests on vacuum components: Allowed or forbidden?

Energy Technology Data Exchange (ETDEWEB)

Schröder, Michael, E-mail: Michael.schroeder@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstr. 1, 17491 Greifswald (Germany); Biedermann, Christoph; Vilbrandt, Reinhard [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstr. 1, 17491 Greifswald (Germany)

2013-10-15

Highlights: The study aims to clarify the applicability of dye penetrant tests on components exposed to high-vacuum. The results show, that the PT application on components for use under vacuum conditions can in general be allowed. The test surface should have a simple geometry. No gaps or holes. An efficient cleaning after PT is necessary. If PT is foreseen TIG should used as the welding procedure. PT tested components should be baked out after the cleaning in a vacuum chamber at min 150 °C. -- Abstract: The penetrant testing (PT) is a common non-destructive procedure for the testing of components and in particular of welds. With PT it is possible to detect surface imperfections (e.g. cracks) which have a special potential to lead to the failure of the component or of the weld. PT is substantially more sensitive than a purely visual examination. Because the complicated geometries of fusion experiments make the accessibility for repairs during the operation extremely difficult, very high efforts on testing with sensitive procedures, for instance with dye penetrant testing during assembly is required. In contrast to this desire for widespread penetrant testing, however, is the general fear that dye penetrant tested components or welds, which are used in the vacuum, are contaminated by the dye in such a way that they do not fulfill the cleanliness requirements for vacuum components. Therefore dye penetrant testing of such vacuum components is usually considered problematic. This study aims to clarify the applicability of dye penetrant tests on components exposed to high-vacuum. Recommendations are formulated concerning the PT procedure of vacuum components and the cleaning procedures for penetrant tested areas under vacuum necessary after a dye penetrant test.

Stakeholder discussion to reduce population-wide sodium intake and decrease sodium in the food supply: a conference report from the American Heart Association Sodium Conference 2013 Planning Group.

Science.gov (United States)

Antman, Elliott M; Appel, Lawrence J; Balentine, Douglas; Johnson, Rachel K; Steffen, Lyn M; Miller, Emily Ann; Pappas, Antigoni; Stitzel, Kimberly F; Vafiadis, Dorothea K; Whitsel, Laurie

2014-06-24

A 2-day interactive forum was convened to discuss the current status and future implications of reducing sodium in the food supply and to identify opportunities for stakeholder collaboration. Participants included 128 stakeholders engaged in food research and development, food manufacturing and retail, restaurant and food service operations, regulatory and legislative activities, public health initiatives, healthcare, academia and scientific research, and data monitoring and surveillance. Presentation topics included scientific evidence for sodium reduction and public health policy recommendations; consumer sodium intakes, attitudes, and behaviors; food technologies and solutions for sodium reduction and sensory implications; experiences of the food and dining industries; and translation and implementation of sodium intake recommendations. Facilitated breakout sessions were conducted to allow for sharing of current practices, insights, and expertise. A well-established body of scientific research shows that there is a strong relationship between excess sodium intake and high blood pressure and other adverse health outcomes. With Americans getting >75% of their sodium from processed and restaurant food, this evidence creates mounting pressure for less sodium in the food supply. The reduction of sodium in the food supply is a complex issue that involves multiple stakeholders. The success of new technological approaches for reducing sodium will depend on product availability, health effects (both intended and unintended), research and development investments, quality and taste of reformulated foods, supply chain management, operational modifications, consumer acceptance, and cost. The conference facilitated an exchange of ideas and set the stage for potential collaboration opportunities among stakeholders with mutual interest in reducing sodium in the food supply and in Americans' diets. Population-wide sodium reduction remains a critically important component of

Components and system tests on the RFX toroidal power supply

International Nuclear Information System (INIS)

Toigo, V.; Zanotto, L.; Gaio, E.; Perna, M.; Bordignon, P.; Coffetti, A.; Novaro, R.; Bertolotto, P.; Rinaldi, E.; Villa, G.

2005-01-01

The paper deals with the component and system tests performed on the new toroidal power supply system of the RFX experiment. The high technological innovation of the system required a deep experimental characterization and validation campaign; special factory tests were performed on prototypes of single components aimed at verifying the most critical design aspects. Consequently an articulated series of tests were performed, based on a step-by-step approach to achieve the desired coordinate operation of the whole system. The test procedures and the most significant results are described in the paper

SAS3A analysis of natural convection boiling behavior in the Sodium Boiling Test Facility

International Nuclear Information System (INIS)

Klein, G.A.

1979-01-01

An analysis of natural convection boiling behavior in the Sodium Boiling Test (SBT) Facility has been performed using the SAS3A computer code. The predictions from this analysis indicate that stable boiling can be achieved for extensive periods of time for channel powers less than 1.4 kW and indicate intermittent dryout at higher powers up to at least 1.7 kW. The results of this anaysis are in reasonable agreement with the SBT Facility test results

Development of sodium technology

International Nuclear Information System (INIS)

Hwang, Sung Tai; Nam, H. Y.; Choi, Y. D.

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, τ c = δ·g -0.83 ·10 (3570/T Na -3.34) , in 400-500 deg C of liquid sodium atmosphere. The characteristics of pressure propagation and gas flow, and

Development of sodium technology

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tai; Nam, H Y; Choi, Y D [and others

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, {tau}{sub c} = {delta}{center_dot}g{sup -0.83}{center_dot}10{sup (3570/T{sub Na}-3.34)}, in 400-500 deg C of liquid sodium atmosphere. The characteristics

The Sodium Tail of the Moon

Science.gov (United States)

Matta, M.; Smith, S.; Baumgardner, J.; Wilson, J.; Martinis, C.; Mendillo, M.

2009-01-01

During the few days centered about new Moon, the lunar surface is optically hidden from Earth-based observers. However, the Moon still offers an observable: an extended sodium tail. The lunar sodium tail is the escaping "hot" component of a coma-like exosphere of sodium generated by photon-stimulated desorption, solar wind sputtering and meteoroid impact. Neutral sodium atoms escaping lunar gravity experience solar radiation pressure that drives them into the anti-solar direction forming a comet-like tail. During new Moon time, the geometry of the Sun, Moon and Earth is such that the anti-sunward sodium flux is perturbed by the terrestrial gravitational field resulting in its focusing into a dense core that extends beyond the Earth. An all-sky camera situated at the El Leoncito Observatory (CASLEO) in Argentina has been successfully imaging this tail through a sodium filter at each lunation since April 2006. This paper reports on the results of the brightness of the lunar sodium tail spanning 31 lunations between April 2006 and September 2008. Brightness variability trends are compared with both sporadic and shower meteor activity, solar wind proton energy flux and solar near ultra violet (NUV) patterns for possible correlations. Results suggest minimal variability in the brightness of the observed lunar sodium tail, generally uncorrelated with any single source, yet consistent with a multi-year period of minimal solar activity and non-intense meteoric fluxes.

Air pollution control system testing at the DOE offgas components test facility

International Nuclear Information System (INIS)

Burns, D.B.; Speed, D.; VanPelt, W.; Burns, H.H.

1997-01-01

In 1997, the Department of Energy (DOE) Savannah River Site (SRS) plans to begin operation of the Consolidated Incineration Facility (CIF) to treat solid and liquid RCRA hazardous and mixed wastes. The Savannah River Technology Center (SRTC) leads an extensive technical support program designed to obtain incinerator and air pollution control equipment performance data to support facility start-up and operation. A key component of this technical support program includes the Offgas Components Test Facility (OCTF), a pilot-scale offgas system test bed. The primary goal for this test facility is to demonstrate and evaluate the performance of the planned CIF Air Pollution Control System (APCS). To accomplish this task, the OCTF has been equipped with a 1/10 scale CIF offgas system equipment components and instrumentation. In addition, the OCTF design maximizes the flexibility of APCS operation and facility instrumentation and sampling capabilities permit accurate characterization of all process streams throughout the facility. This allows APCS equipment performance to be evaluated in an integrated system under a wide range of possible operating conditions. This paper summarizes the use of this DOE test facility to successfully demonstrate APCS operability and maintainability, evaluate and optimize equipment and instrument performance, and provide direct CIF start-up support. These types of facilities are needed to permit resolution of technical issues associated with design and operation of systems that treat and dispose combustible hazardous, mixed, and low-level radioactive waste throughout and DOE complex

Mass transfer of steels for FBR in sodium loop

International Nuclear Information System (INIS)

Susukida, Hiroshi; Yonezawa, Toshio; Ueda, Mitsuo; Imazu, Takayuki; Kiyokawa, Teruyuki.

1976-06-01

In order to grasp quantitatively the corrosion and mass transfer of steels for FBR in sodium loop and to establish their allowable stress value and corrosion rate, a special sodium loop for material testing was designed and fabricated and the steels were given 3010 hours exposing test in the sodium loop. This paper gives the outline of the sodium loop and the results of the test. (1) Carburization and a slight increase in weight were observed in the specimens of type 304 stainless steel exposed in the sodium loop for 3010 hours, while decarburization was observed in the specimens of 2 1/4 Cr-1 Mo steel. It is considered that these phenomena were caused by the downstream factor of the sodium loop. (2) A remarkable decrease of Charpy absorbed energy was observed in the specimens of type 304 stainless steel exposed in the sodium loop. It is considered that this resulted from the weakening of the grain boundary due to heat history and mass transfer. (3) The specimens exposed in the sodium loop must be washed by ultrasonic waves in a water bath after washing in alcohol. (auth.)

French research on a general approach to sodium fires

International Nuclear Information System (INIS)

Malet, JC.

1989-01-01

This document gives a general idea of how one may deal with the safety of an installation in case of a sodium leak generating either a sodium pool fire, or a sodium spray fire, or a combined sodium fire, and in case of a sodium leak either with or without fire and/or any aggravating phenomenon such as aerosols or sodium-concrete reactions, for example. This paper describes the means used to reduce fire consequences. These means are either design codes used for dimensioning premises with their ventilation system or for defining a course of action to be taken in case of fire, or equipment and components (fire fighting equipment, detection devices, etc.) used to prevent or to fight fires, or finally, to repair the installation after a fire, the latter operation including the processing of the residues. (author)

Optimal test intervals of standby components based on actual plant-specific data

International Nuclear Information System (INIS)

Jones, R.B.; Bickel, J.H.

1987-01-01

Based on standard reliability analysis techniques, both under testing and over testing affect the availability of standby components. If tests are performed too often, unavailability is increased since the equipment is being used excessively. Conversely if testing is performed too infrequently, the likelihood of component unavailability is also increased due to the formation of rust, heat or radiation damage, dirt infiltration, etc. Thus from a physical perspective, an optimal test interval should exist which minimizes unavailability. This paper illustrates the application of an unavailability model that calculates optimal testing intervals for components with a failure database. (orig./HSCH)