Sodium Loop Research Articles

Summary and Interpretation of the CABRI REP-Na Program

The CABRI REP-Na program was performed in the sodium loop of the CABRI reactor by the French Institut de Radioprotection et de Sûreté Nucléaire. The objective was to study the behavior of high-burnup UO2 and mixed-oxide (MOX) fuel during a reactivity-initiated accident (RIA) and involved eight tests with UO2 and four with MOX fuel. Failures of some UO2 and MOX fuel rods at enthalpy levels ranging from 125 to 472 J/g (30 to 113 cal/g) demonstrated the need for further development of the present safety criteria pertaining to fuel behavior. Detailed interpretation of the test data led to identifying the deleterious influence of a high clad corrosion level with hydride concentrations on clad failure, the contribution of grain boundary gases on fission gas release, and potential clad loading, mainly in MOX fuel during the first phase of the transient without significant clad temperature increase.Questions still remain concerning the transient fission gas behavior and its impact on clad loading during the entire transient, the rod behavior with high clad temperature and internal pressure, and the postfailure phenomena (fuel ejection, fuel/coolant interaction with finely fragmented solid fuel). These issues will be addressed by the CABRI International Program tests under typical pressurized water reactor conditions.

ICONE15-10213 EDDY CURRENT FLOW SENSOR DEVELOPMENT AND TESTING FOR LMFBR SODIUM PUMPS

500MWe Prototype Fast Breeder Reactor (PFBR), under construction at Kalpakkam has two vertical centrifugal pumps in the primary system, operating in parallel. Eddy Current Flow meter (ECFM) has been chosen to measure pump discharge flow. ECFM is an electromagnetic device in which the sensor consists of three coils wound on an iron former. Flow probe is replaceable and can be inserted vertically from the top in pump suction casing, at a point, upstream of the diffuser. The pump will have two probes; each with two flow sensors fitted one above the other. Flow is established through the sensor to the suction side. Theoretical and experimental hydraulic studies were conducted in water to assess flow through the sensor and the pressure drop characteristics. The sensor was calibrated in a sodium loop for velocities established from hydraulic experiments. The sensor was tested upto 773 K and with electronic temperature compensation circuits, the error limit was established. The sensor essentially gives the head developed by the pump. Correlation with respect to the actual pump flow, head developed at different pump speeds were established. This paper presents the details of the sensor and studies conducted.

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ⩽1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

Therapie von Komplikationen der Leberzirrhose

Complications of liver cirrhosis are usually confined to advanced stages of the disease. Bleeding from esophageal or gastric varices may be prevented by treatment with beta-blockers or by endoscopic band ligation in case of large varices and intolerance for beta-blockers. Treatment of an acute bleeding episode from varices can efficiently be treated by endoscopic procedures, potentially in combination with drug therapy. In case of bleeding uncontrolled by endoscopy, TIPS is an effective alternative in selected patients. Treatment of ascites consists of reduction of sodium intake, aldosterone antagonists, and loop diuretics as needed. TIPS or repeated paracentesis may be necessary in refractory ascites. Spontanous bacterial peritonitis (SBP) must be sought and treated with antibiotics in conjunction with albumin administration in order to reduce mortality. Hepatorenal syndrome is characterized by a poor prognosis. Therefore, liver transplantation should be considered in appropriate patients.

Experience in Operating the BN-600 Unit at the Belyi Yar Nuclear Power Plant

Data on the experience in operating the BN-600 reactor with sodium coolant in the first and second loops are presented. It is shown that the equipment and the systems in the sodium loops, which have operated for more than 23 years, are highly realiable. The average installed capacity utilization factor for this period reached about 74%. The losses due to rupture of the heat-transfer pipes in the steam-generator modules and the sodium leaks from the loops were about 0.3%. Disruptions of normal operation are detected reliably and contained by the safety systems present in the unit. Unique experience in performing maintenance and repair work on the sodium equipment has been gained on the BN-600 reactor.

Mutagenesis Studies toward Understanding the Mechanism of Differential Reactivity of Factor Xa with the Native and Heparin-Activated Antithrombin

A unique pentasaccharide fragment of high-affinity heparin activates antithrombin (AT) to enhance its rate of complex formation with factor Xa (FXa) by 200-300-fold. Recent results have indicated that the activation of AT is associated with the exposure of a cryptic exosite on the serpin that is an interactive site for FXa in the complex. Previously, we identified Arg(150) on the autolysis loop of FXa as a candidate residue that may specifically interact with the heparin-activated AT. Three other surface loops on FXa including 39, 60, and the sodium-binding 220 loops have been implicated to be critical for the protease interaction with the activated AT. To determine the extent of the contribution of these loops to the specificity of the FXa interaction with activated AT, several loop mutants of the protease were prepared and their reactivity with AT was studied in both the absence and presence of pentasaccharide. Analysis of the inhibition kinetic data suggests that the residues of both 39 and 60 loop make a minor contribution to the recognition of AT in both the native and activated conformation of the serpin. On the other hand, the reactivity of AT with the sodium loop mutants of FXa in the absence of the cofactor was severely impaired. However, the extent of the rate-accelerating effect of pentasaccharide in the AT inhibition of the mutants was not affected. These results suggest that all three loops play a role in the specificity of the FXa-AT interaction; however, neither loop specifically interacts with the activated conformation of the serpin.

Safety management procedures and practices at Indira Gandhi Centre for Atomic Research

The Indira Gandhi Centre for Atomic Research (IGCAR) operates FBTR (Fast Breeder Test Reactor), KAMINI (neutron source reactor), radiometallurgical laboratory, radiochemical laboratory, reprocessing plant, industrial scale sodium loops, advanced research laboratories, workshops, etc. Codified safety management procedures with systematic surveillance are essential for safe and reliable operations and these are described under the classifications of radiation safety, industrial safety and reactor operations with special emphasis on the human factor. Health physics teams, independent of the plant facility, supervise the radioactive facilities of the centre. Industrial safety standards are maintained by another independent section. Safety management for the reactors include a clear organisational structure, adequate documentation, compulsory training and licencing, safe working methods taking into account human factors and review by independent safety authorities.

Compatibility of steel No. 1.4970 with liquid sodium at high temperatures

The sodium corrosion and the influence of liquid sodium at 873 and 973 K on the creep strength of cladding tubes of steel No. 1.4970 was studied in a sodium loop. The corrosion and creep tests with cladding tube specimens were extended up to 10 000 h. The mass losses showed an effect of the oxygen content in the sodium loop which was controlled by means of a cold trap. The exposure to sodium caused chemical changes as losses of chromium, nickel and manganese and the carbon concentrations were increased in the outer zones of the material. The boron content of the steel was significantly reduced. Despite the changes of the chemical composition of the material, the time to the creep deformation of 1% at 973 K was not influenced by the sodium corrosion. This might be due to the carburization of the steel. All corrosion effects were only marginal at 873 K.

Analysis of CREVONA sodium loop material

Stainless steel specimens equivalent to AISI type 304 taken from the CREVONA sodium loop (Forschungszentrum Karlsruhe, Germany), which was operated for more than 80 000 h, were analysed for microstructures and changes in chemical composition of depleted layers using SEM/EDAX. SEM micrographs were obtained in the cross-section of the specimens to reveal the thickness of the corroded layer. EDX analysis confirms depletion of Ni and Cr in the corroded layer. The leaching rates of chromium and nickel are obtained from the depleted layer width. These results are compared with the degraded layer and corrosion resistant node formation in sodium-exposed AISI type 316 SS specimens. The corroded layer widths of the specimens taken from the CREVONA loop determined using known models for life prediction like those proposed by Thorley and Tyzack are found to be much less than the actual layer widths observed experimentally after sodium exposure. The materials were exposed to flowing sodium for about 10 years.

An electrochemical hydrogen meter for measuring hydrogen in sodium using a ternary electrolyte mixture

An electrochemical sensor for measuring hydrogen concentration in liquid sodium that is based on a ternary mixture of LiCl, CaCl 2 and CaHCl as the electrolyte has been developed. DSC experiments showed the eutectic temperature of this ternary system to be ∼725 K. Impedance spectroscopic analysis of the electrolyte indicated ionic conduction through a molten phase at ∼725 K. Two electrochemical hydrogen sensors were constructed using the ternary electrolyte of composition 70 mol% LiCl:16 mol% CaHCl:14 mol% CaCl 2 and tested at 723 K in a mini sodium loop and at hydrogen levels of 60–250 ppb in sodium. The sensors show linear response in this concentration range and are capable of detecting a change of 10 ppb hydrogen in sodium over a background level of 60 ppb. Identification of this electrolyte system and its use in a sensor for measuring hydrogen in sodium are described in this paper.

1935 Advanced IHX の基礎研究 : 中間熱交換媒体鉛ビスマスの強制対流熱伝達

Secondary sodium loop elimination is proposed for the loop type LMFBR with using Advanced Intermediate Heat Exchanger (AIHX) for reduction in size and cost. This heat exchanger contains primary sodium tubes, ant tertiary water tubes in a tank filled with intermediate media. Pb-Bi eutectic alloy is one of a promising candidate of an intermediate medium. The result of Pb-Bi experiment in natural and near forced convection is reported.

Numerical Simulation of Sodium-Water Reaction Products Transport in Steam Generator of Liquid Metal Fast Breeder Reactor on Small Water/Steam Leak

The paper presents a three-dimensional model of sodium-water reaction product transport in steam generators with a small water/steam leak. The model is used to study the effects of thermodynamic parameters of steam generators of Liquid Metal Fast Breeder Reactor (LMFBR) on the leak response of a small water leak detection system and to optimize the leak detection system design. A hydrogen gas line density concept is used to develop the differential equation of hydrogen ion transport. The model can be applied to analyze the leak response of hydrogen meters, oxygen meters and other concentration-type leak detection systems. The model was verified with water injections into sodium with different temperatures in a simple sodium loop with a hydrogen meter.

Compatibility of Structural Materials with Sodium According to Data from the Experience Gained in 40 Years of Operation of the BR-5/BR-10 Reactor

The results of investigations of the corrosion characteristics of samples of structural materials cut from the main sodium loops of a BR-5/BR-10 sodium-cooled research reactor during operation over the period 1959–1999 are presented. Analysis of the results of the materials-engineering investigations show that the sodium coolant has good corrosion characteristics and that the operating lifetime of sodium equipment in commercial fast reactors can be increased to 60 yr and longer.

Numerical Simulation of Sodium-Water Reaction Products Transport in Steam Generator of Liquid Metal Fast Breeder Reactor on Small Water/Steam Leak.

The paper presents a three-dimensional model of sodium-water reaction product transport in steam generators with a small water/steam leak. The model is used to study the effects of thermodynamic parameters of steam generators of Liquid Metal Fast Breeder Reactor (LMFBR) on the leak response of a small water leak detection system and to optimize the leak detection system design. A hydrogen gas line density concept is used to develop the differential equation of hydrogen ion transport. The model can be applied to analyze the leak response of hydrogen meters, oxygen meters and other concentration-type leak detection systems. The model was verified with water injections into sodium with different temperatures in a simple sodium loop with a hydrogen meter.

Lithium cooled blanket of RF DEMO reactor

Two blanket options are considered for RF DEMO-S reactor (fusion power ∼2.4 GW) with steady-state operation: lithium (Li) self-cooled and ceramic breeder helium-cooled ones. Lithium-cooled blanket, which uses Li as coolant and tritium breeder and V–Cr–Ti (vanadium–chrome–titanium) alloy as structural material, is described here. Poloidal type of Li flow is chosen under the assumption that electroinsulating coating is developed on the interface of Li/structure material. Results of preliminary neutronic, thermal-hydraulic and stress analysis are presented. The steam turbine cycle converting the thermal to electric energy is recommended with intermediate sodium loop on the basis of the FBR experience.

The chemical state of defective uranium-plutonium oxide fuel pins irradiated in sodium cooled reactors

Steady-state irradiation experiments were conducted in the sodium loop of the Siloe reactor on artificially failed mixed oxide pins that had been pre-irradiated in fast reactors up to 11.5% burnup. The formation of the predominant reaction product Na 3(U,Pu)O 4 starts on the fuel surface and is terminated when a lower O/(U + Pu) threshold of the fuel is attained. The axial extent of the reaction product depends on the size of the initial cladding defect. The occurrence of secondary cracks is possible. Na(U,Pu)O 3 forms at higher fuel temperatures. The existence of Na 3U 1− x Pu x O 4 is shown in pre-irradiated blanket pins after artificial defect formation. Caesium in the oxocompounds is reduced to the metallic state and is dissolved in the coolant. Evidence of a very low chemical potential of oxygen in defective fuel pins is sustained by the occurrence of actinide-platinum metal phases formed by coupled reduction of hypostoichiometric fuel with ε-(Mo,Tc,Ru,Rh,Pd) precipitates. Continued operation of defective pins is not hazardous by easy precautions.

Formation of NaCrO 2 in sodium systems of fast reactors and its consequence on the carbon potential

The most important corrosion product in sodium loops is NaCrO 2. In this study, the threshold oxygen level in sodium, for the formation of NaCrO 2 when in contact with austenitic (AISI 304 SS, 316 SS and D9 alloy) and ferritic (Fe-9Cr-1Mo and Fe- 2 1 4 Cr-1Mo) steels at temperatures between 573 and 1023 K were evaluated. By a weighted regression analysis of the reliable measurements reported in the literature, the Gibbs energy of formation of NaCrO 2, Δ f G° = −913762.8 + 237.24 T J/mol, was obtained. Using this relation, the threshold oxygen levels ( C O ) were evaluated and fitted by regression analysis into expressions of the form ln C O (ppm) = A + B T . Based on the nature of the carbides formed in AISI 304 SS and 316 SS, the effect of depletion of chromium on the carbon activity of these alloys was analysed by a thermodynamic model.

J. Materials Science: R.G. O'Dennell et al (Materials Research Laboratories, Australia) Vol 27, No 23, 1992, 6490–6494

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type, mixed oxide (MOX) fueled reactor. Sodium flow measurement in various loops of the reactor is of prime importance from the operational and safety aspects. To measure the flow of electrically conducting liquid sodium in secondary circuits, permanent magnet flowmeters (PMFMs) made up of ALNICO-5 are used in Fast Breeder Test Reactor (FBTR), PFBR and existing sodium loops in Indira Gandhi Centre for Atomic Research (IGCAR). PMFM made up of ALNICO-5 are bulky and heavy, hence to reduce size of flowmeter a PMFM made up of Samarium Cobalt (SmCo) is designed and calibrated in existing sodium loop. Calibration of flowmeter requires dedicated sodium rig with pipe sizes of same dimension as flowmeter pipe, which needs modification in existing system. As an alternate method of sodium calibration, three dimensional Finite Element Modeling of SmCo PMFM is done for its voltage signal prediction. In simulation magnetic field distribution across the flowmeter pipe and effect of sodium flow on magnetic field are analyzed. Effect of sodium temperature on flowmeter voltage signal and its change with change in sodium flow rate is predicted with modeling and compared with experimental tests conducted for use of PMFM.

J. Magnetism and Magnetic Materials: P.G. McCormick et al (University of Western Australia, Australia) Vol 116, No 3, 1992, L320–L324

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type, mixed oxide (MOX) fueled reactor. Sodium flow measurement in various loops of the reactor is of prime importance from the operational and safety aspects. To measure the flow of electrically conducting liquid sodium in secondary circuits, permanent magnet flowmeters (PMFMs) made up of ALNICO-5 are used in Fast Breeder Test Reactor (FBTR), PFBR and existing sodium loops in Indira Gandhi Centre for Atomic Research (IGCAR). PMFM made up of ALNICO-5 are bulky and heavy, hence to reduce size of flowmeter a PMFM made up of Samarium Cobalt (SmCo) is designed and calibrated in existing sodium loop. Calibration of flowmeter requires dedicated sodium rig with pipe sizes of same dimension as flowmeter pipe, which needs modification in existing system. As an alternate method of sodium calibration, three dimensional Finite Element Modeling of SmCo PMFM is done for its voltage signal prediction. In simulation magnetic field distribution across the flowmeter pipe and effect of sodium flow on magnetic field are analyzed. Effect of sodium temperature on flowmeter voltage signal and its change with change in sodium flow rate is predicted with modeling and compared with experimental tests conducted for use of PMFM.

Liquid sodium—The heat transport medium in fast breeder reactors

Liquid sodium is used as a coolant in fast breeder reactors on account of its excellent heat transfer properties. It must, however, be in the pure form to be compatible with structural materials. Techniques for its purification to nuclear grade and its characterization had to be developed in our laboratory before we could embark on an R&D programme.It is essential to monitor hydrogen at ppb levels in the sodium circuits of the fast reactor in order to detect water leaks in the steam generator in a timely manner. Similarly it is useful to make on-line measurements of oxygen and carbon at trace levels. Electrochemical sensors have been developed in our laboratory for this purpose. These compact sensors work on the principle of concentration cells and are far superior to devices used elsewhere for this purpose.Corrosion of structural materials in the sodium environment depends on the oxygen content of sodium. In order to understand the mechanism of this corrosion, one must have a good grasp of the thermochemistry of the ternary systems, Na-M-O, where M stands for the alloying constitutents of stainless steels. The phase diagrams of most of these systems were established in our laboratory. A specially designed sodium loop is used in the study of corrosion, activity transport and kinetics of sodium-water reaction.