Steam Generator Test Facility Research Articles

Cross-scale coupling analysis method for OTSG in SFR and its application

The once-through steam generators (OTSG) in sodium-cooled fast reactor (SFR) are characterized by full-regime flow boiling heat transfer, coupled heat transfer on both sides and complex structure with a large dimensional span. This makes the computational fluid dynamics (CFD) and 3D refinement modeling methods to analyze the 3D temperature field of OTSG containing large-scale heat transfer tube bundles generally face the problems of high computational difficulty, low solution efficiency and poor solution stability. In this paper, based on the full-regime flow boiling heat transfer model and the flow pattern discrimination model, a one-dimensional full-regime flow boiling heat transfer model from subcooled water to superheated steam on the water side (tube side) and its cross-scale coupling solution model with the three-dimensional refined CFD model on the sodium side (shell side) are established. The effectiveness of the coupling analysis model was verified by comparative analysis with experimental data of the Indian 19-tube Steam Generator Test Facility (SGTF). The independent coupling of the water side of each heat transfer tube in the OTSG with the sodium side 3D fluid domain is realized. The results show that the cross-scale coupling method using the one-dimensional full-regime flow boiling heat transfer model for the water side and the three-dimensional CFD model for the sodium side can accurately predict the temperature distribution, the main heat transfer characteristic points on the full height of each heat transfer tube, and can obtain the temperature difference between different heat transfer tubes and tube walls. The maximum error between the calculated results and experimental data for the distribution of sodium temperature along the axial direction of the tubes is less than 3%. The cross-scale coupled solution method avoids the direct adoption of the three-dimensional multiphase flow model, which significantly improves the efficiency and stability of the solution of the three-dimensional temperature field of OTSG. This study provides an efficient 3D refinement numerical analysis method for the design and safety verification of OTSG.

Development and validation of transient Thermal-Hydraulic model for SFR steam generators

The steam generator (SG) is an important component in sodium-cooled fast reactors (SFRs). It transfers the heat from sodium systems to the balance of plant (BOP), producing steam for electricity production. Any disturbances originating from the BOP propagate to the nuclear systems through SG. Hence, it is especially important to model and predict the thermal–hydraulic behavior of SG during such transients. Towards this, a thermal–hydraulic model of SG of a typical medium-sized SFR has been developed using general-purpose system simulation code – FLOWNEX. Usually, the moving boundary method is used to model SGs as it involves phase change and multiple heat transfer regimes. However, in the current model, the fixed mesh approach, generally used for modeling pipelines, single-phase heat exchangers, etc. in plant dynamic codes, has been used to model SGs. In this paper, the modeling details and the solution procedures used in the FLOWNEX are discussed in detail. The steady-state results of the model developed have been compared with the results from in-house code DESOPT. The results were in good agreement at various power levels as well as in off-design conditions. The transient performance of the model has been validated against an in-house experiment carried out in Steam Generator Test Facility (SGTF), IGCAR, India, and a steam generator shutdown experiment carried out in Energy Technology Engineering Center (ETEC), Santa Susana Field Laboratory (SSFL), USA. The transient results were also in good agreement, thereby increasing the confidence level of the predictions. The model developed will be integrated to plant dynamic codes containing other system models and will be used to carry out system-level simulations in the future.

Study on startup characteristics of prototype once-through steam generator for China fast reactor

Once-through steam generator (OTSG) is a vital heat transfer equipment of China Fast Reactor-600 (CFR-600), and its startup characteristics are of great importance to the safe operation of nuclear power plant. In this paper, first of all, the Thermal-hydraulic analysis Code Of Sodium heated once-through Steam generator (TCOSS) has been further optimized, and the code was introduced in the text. Secondly, the Steam Generator Test Facility (SGTF) and India Commercial Fast Breeder Reactor (CFBR) were selected for steady-state verification, and the feedwater flow rate increase condition of the prototype OTSG experiment was selected for transient verification. All the verifications were in good compliance, which proved the reliability of the TCOSS code for subsequent calculation of startup conditions. Finally, taking the prototype OTSG of CFR-600 as the research object, and the transient characteristics of cold startup and hot startup have been predicted and analyzed. The results show that both processes of cold startup and hot startup for prototype OTSG can be realized stably. The simulation results of startup processes showed a good agreement with expectations, which can provide support for the development of CFR-600 steam generator.

Influence of thermal expansion bend and tubesheet geometry on guided wave inspection of steam generator tubes of a fast breeder reactor

Periodic assessment of steam generator tubes of a sodium-cooled nuclear reactor is very crucial for smooth operation of steam generators. To examine the integrity, an in-bore magnetostrictive transducer capable of launching and receiving longitudinal ultrasonic guided waves (L(0,2) mode) from the inner diameter side of a steam generator tube developed in-house is used. Preliminary tests conducted on defective steam generator tubes with thermal expansion bends (three successive bends) of the mockup steam generator test facility yield a good sensitivity of 20% wall thickness deep flaw (0.46-mm deep and 1-mm wide half-circumferential groove) and the location accuracy of 10 mm. In order to remove high noise, wavelet-based denoising using discrete wavelet transform is used which improves the signal-to-noise ratio by 5–10 dB. In addition, cross-correlation technique is also used to denoise and unambiguously identify the defect echoes amid noise and multiple reflections between the defects. Furthermore, influence of the thermal expansion bend and tubesheet–spigot structure on L(0,2) mode is studied using the finite element analysis. It is observed that in the thermal expansion (multiple) bend, axisymmetric L(0,2) mode becomes non-axisymmetric (maximum and minimum amplitudes at extrados and intrados, respectively) and undergoes mode conversion to a weak flexural mode F(1,3). The results are validated experimentally. In the tubesheet–spigot structure, L(0,2) mode is found to have ∼10% reflection from spigot–tubesheet transitions, and it is seen to mode convert to bulk waves in the tubesheet. In conclusion, thicker tubesheets are found to be better from the perspective of inspection.

Validation of the finned sodium–air heat exchanger model in MARS-LMR

A Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), for which the Korea Atomic Energy Research Institute (KAERI) has designed a pool type sodium-cooled fast reactor, has a decay heat removal system (DHRS). The DHRS consists of sodium-to-sodium and sodium-to-air heat exchangers and their connecting pipes. There are four loops, which have two active and two passive type sodium–air heat exchangers, that are called a finned-tube sodium-to-air heat exchanger (FHX) and a helical-tube sodium-to-air heat exchanger (AHX), respectively. Recently, Zukauskaus’s air–sodium heat transfer models have been added in MARS-LMR, which is a safety analysis code for the PGSFR. In this study, to validate the newly added heat transfer models for the FHX, two experiments are selected: one is a performance test for a sodium to air heat exchanger (AHX) in the steam generator test facility (SGTF) for the prototype fast breeder reactor (PFBR), and the other is a start-up test with a dump heat exchanger (DHX) in JOYO. All validation results indicate that Zukauskaus’s correlation slightly over-estimates the heat transfer. One possible reasons is a smaller number of rows in the test bundle, which was already mentioned by Zukauskaus and Karni. The RMS values for the prediction of sodium temperature for the PFBR and the JOYO are 16.25% and 27.5%, respectively. When a correction factor is applied, their RMS values improve to 3.19% and 20.71%, respectively. In addition, the MARS-LMR’s prediction for the JOYO shows a much better accuracy with RMS of 7.45% when corrected sodium flow rates are applied.

Porous Body Model Based Parametric Study for Sodium to Air Heat Exchanger Used in Fast Reactors

Sodium to air heat exchangers (AHX) with finned tubes is used in fast breeder reactors for decay heat removal. The aim of decay heat removal is to maintain the fuel, clad, coolant, and structural temperatures within safety limits. To investigate the thermal hydraulic features of AHX, a robust porous body based computational fluid dynamics (CFD) model has been developed and validated against the experimental data obtained from a model AHX of 2 MW capacity in Steam Generator Test Facility at the Indira Gandhi Centre for Atomic Research, Kalpakkam. In the present paper, the developed porous body model is used to study the sodium and air temperature distribution and the influence of various parameters that affect the heat removal rate and sodium outlet temperature in full-size AHX used in the fast breeder reactors. The parameters include mass flow rates and inlet temperatures of sodium and air. The focus of the study has been to identify conditions that can pose the risk of sodium freezing.

Experimental evaluation of the heat transfer performance of sodium heated once through steam generator

Steam generator is a crucial component in a nuclear power plant because its availability is directly linked to the availability of heat transport system and thus the plant availability. In Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction in India, eight number of steam generators each with a heat transfer capacity of 156MWt transfers 1250MW of heat from secondary sodium to the conventional steam/water system. The sodium heated once through steam generator with 23m long seamless straight tubes produces super heated steam at 17.2MPa pressure and 493°C temperature. A model steam generator of 5.5MWt power was tested in steam generator test facility of Indira Gandhi Center for Atomic research for validating the thermal hydraulic and mechanical design of the steam generator. The testing revealed the adequacy of heat transfer capability of the steam generator to transfer the intended power. From the experimental data it is estimated that the steam generator has 8.3% more tube surface area than the required to produce steam at nominal conditions. This paper gives the details of the model steam generator, heat transfer experiments conducted to validate the thermal design and the method for estimating the additional heat transfer area in once through type steam generator.

Detection of Simulated Steam Leak into Sodium in Steam Generator of PFBR by Argon Injection Using Signal Analysis Techniques

The Prototype Fast Breeder Reactor (PFBR) is pool-type, sodium-cooled reactor. PFBR has eight steam generators (SGs) in its secondary circuit. The availability of SG units in the secondary circuit is of critical importance, as this determines the entire plant availability. Constant exposure to high temperature and high pressure may result in water or steam leaks in the tubes of SGs by which sodium and water reaction will take place. The growth in a leak may lead to tube rupture and large pressure increase in the secondary circuit. So, an important requirement from the viewpoints of safety and economics is to detect leaks at the incipient stage. Hydrogen detection methods, which are currently used to detect the initiation of steam leaks to sodium, involve transport delays, as the hydrogen evolved during the leak has to reach the sensor location. It is possible to detect a leak of 1 g/s within 1 s by acoustic leak detection. Another advantage is that it is possible to locate the position of a leak by installing several acoustic sensors on a SG.Experiments to develop a suitable signal-processing technique were carried out in the Steam Generator Test Facility (SGTF); argon was injected into sodium at argon pressure of 2 to 10 MPa through a 0.5-mm orifice. Signal-processing techniques, an autoregressive noise variance technique, and a wavelet detailed coefficient variance technique were studied and compared for their sensitivity to detect a steam/water leak. The acoustic technique employing wavelet decomposition is found to be promising for detecting a leak at the incipient stage. This paper discusses the details of the experiments carried out and the instrumentation and signal analysis techniques used for leak detection in SGs.

Experimental evaluation of sodium to air heat exchanger performance

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 2MW 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 2.34MW 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, Nueq 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 Nueq with respect to the Reynolds number was studied.

Estimation of hydrogen mass flux from water side to sodium side in a sodium heated Mod 9Cr1Mo steam generator

Heat, produced in a SFR (Sodium cooled Fast Reactor) is exchanged to a sodium circuit, where it is further transferred from sodium to water in a steam generator (SG). SG, typically made of ferritic steel, is a shell and tube heat exchanger with sodium on shell side and water/steam on tube side. Water/steam, flowing through the SG tube, reacts with the tube material, producing magnetite and hydrogen. Most of the hydrogen, formed in the water side, permeates into sodium through the SG tube wall and becomes the major impurity in sodium. The mass flux of hydrogen, permeating into sodium, is particularly important in the design of cold trap (CT) which is used for purifying sodium. Experiments were performed with a model SG made of Mod 9Cr1Mo having 19 tubes, in Steam Generator Test Facility (SGTF) at Indira Gandhi Centre for Atomic Research (IGCAR), to quantify hydrogen flux. Material, structural layout and operating conditions of SG of Prototype Fast Breeder Reactor (PFBR) were simulated. This paper brings out the data and experiences gained through the experiments.

Failure Analysis of Leaked Stainless Steel Housing of Electro Chemical Hydrogen Metre

The failure of the stainless steel housing of Electro Chemical Hydrogen Metre (ECHM) in the Steam Generator Test Facility resulting in a minor sodium fire has been investigated through systematic metallographic and fractographic analyses. The root cause of the failure was identified as caustic stress corrosion of the stainless steel flange that forms a part of the ECHM housing.

Steam generator test facility—A test bed for steam generators of Indian sodium cooled fast breeder reactors

Abstract Second stage of India's three stage nuclear power programme consisting of the development and operation of fast breeder reactors (FBRs) commenced with the construction, commissioning and operation of fast breeder test reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. FBTR operating experience and better understanding of the technologies involved has enabled India to design a 500 MWe prototype fast breeder reactor (PFBR). Construction of PFBR is in advanced stage at Kalpakkam. Steam generator (SG) is one of the critical components of a fast reactor power plant. It is a once-through SG housing both water at high pressure and sodium at low pressure separated by a tube wall. Any defect to this barrier resulting in an internal leak allowing sodium and water to react could severely affect the plant availability. A 5.7 MWt capacity steam generator test facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test a 5.5 MWt sodium heated once through SG of 19 tubes similar to the indigenously designed 157 MWt, 547 tube SG of PFBR. The test programme includes the evaluation of thermal hydraulic performance of SG, instability mapping, measurement of tube bundle vibrations etc. Most of the planned experiments have been completed successfully. The heat transfer experiment has proved that the SG has a heat transfer area margin of 12.25%. This experiment has also helped in validating the design codes and reducing the area margin for SGs of future FBRs. Experiment on flow induced vibration of tube bundles has shown that the bending stress due to vibration is much less than the endurance limit of the material. The results of the experiments and the valuable lessons learnt from the operating experience have been successfully applied to PFBR and will be applied to future commercial fast breeder reactors. The facility is also planned as a test bed for SGs of Indian FBRs and to assess improved designs of auxiliary equipment used in FBRs. This paper brings out various experiments conducted and the experience gained during the operation of the facility.

Design, Development and Testing of a Large Capacity Annular Linear Induction Pump

Abstract Electromagnetic Pumps have been used for pumping liquid sodium in auxiliary circuits such as fill and drain and purification circuits of sodium cooled fast breeder reactors. Despite their low efficiency these pumps are used in fast reactors because of their high reliability and low maintenance due to absence of moving parts. Besides, EM Pumps can be used for pumping impure sodium. IGCAR has developed electromagnetic pumps of various capacities and successfully used them in experimental facilities. This paper deals with the design, development and performance testing of a large electromagnetic pump called Annular Linear Induction Pump (ALIP). This 170 m3/h capacity ALIP is for use in PFBR Secondary Sodium Fill and Drain Circuit (SSFDC) and was introduced in the sodium circuit of SGTF for testing its performance. In this type of pump, a linearly traveling magnetic field is generated by means of circular windings placed spatially apart in slots and excited by 3-phase supply. This traveling field induces circulating currents in liquid sodium which generates secondary magnetic field. Interaction of primary magnetic field and secondary magnetic field produces pumping force on liquid sodium. The pump duct is made of SS316L. Pump winding is made up of copper with class H insulation. The pump is designed for 360 V and for a maximum sodium temperature of 450 °C. The pump is a reflux type of pump with both inlet and outlet on the same side. The pump was tested in the cold leg of Steam Generator Test Facility (SGTF) and its performance characteristics were obtained. The efficiency of the pump was also calculated and compared with the theoretical value. The successful testing and operation of the pump in SGTF has indicated sound design and indigenous manufacturing capability. This paper describes the design data of the pump and details of the testing carried out in IGCAR

Numerical and Experimental Studies on the Performance of Thermal Baffles in Sodium Heated Steam Generator

The Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam, India. PFBR has eight numbers of sodium heated once through Steam Generators (SG) each with 157 MWt capacity. Parts like steam generator tube sheet will see high temperature gradients during normal operation as well as in transient conditions. Thermal baffles are provided to protect the steam generator tub e sheets from the high temperature gradients during transient and steady state conditions. In Steam Generator Test Facility (SGTF) of IGCAR a 5.5MWt capacity SG was tested for its thermal hydraulic performance. Effectiveness of the thermal baffles provided in the model SG of PFBR is studied numerically and the performance of the same has been predicted. Further Transient experiments were conducted in Steam Generator Test Facility for evaluating the same. This paper brings out the details of numerical studies and the experimental evaluation of the effectiveness of thermal baffles provided to protect the steam generator tube sheets.

Steam Generators for Future Fast Breeder Reactors

Abstract The operating experience of fast reactors worldwide has highlighted the need for reliable performance of Steam Generator (SG) as it is one of the most critical components deciding the plant availability. Based on design, development and manufacture of steam generator for 500 MWe Prototype Fast Breeder Reactor (PFBR) and also the reassuring operational feedback from experimental Steam Generator Test Facility (SGTF), the concept of integral once through SG with mod. 9Cr 1Mo as principle material of construction is retained for future FBRs. SG is a vertical counter flow shell & tube heat exchanger with sodium on shell side and water / steam in tubes. For future FBRs, the present design of SG for PFBR is reassessed and optimised based on manufacturing experiences with focus on improved economics and enhanced safety. Various key aspects like selection of number of SG modules per secondary sodium loop, tube length, heat transfer area margin and optimization of tube size are discussed in the paper. The results of CFD studies carried out with a 30 sector model coupled with an in-house code (DESOPT) to investigate on thermal and flow distribution of such longer steam generators at different operating conditions are also highlighted. Through detailed optimisation studies, it is concluded that 3 SG per loop with each having 433 tubes of 30 m long, 12.6 mm ID and 2.4 mm wall thickness for a design life of 60 years provides economical configuration leading to a saving of ∼25% in total cost. Further, enhanced safety is ensured by reduction of ∼40% tube to tube-sheet joints as compared with PFBR SG.

Validation of the MSG (Multi-dimensional Thermal-hydraulics Analysis Code for Steam Generators) for CFD Modelling of Sodium Heated Steam Generators

A CFD code, MSG (Multi-dimensional Thermal-hydraulics Analysis code for Steam Generators), which is optimized for the helical coil type steam generators of the MONJU Fast Breeder Reactor at Tsuruga, Japan, has been developed. The purpose of this study is to validate the MSG code against experimental data recorded at MONJU and at the 50 MW steam generator test facility (SGTF) at O-arai Engineering Center, Japan, and to perform a cross-correlation with results from the superCOPD code. Quantitative all benchmarks produced sodium temperatures within 2% of measured values, whereas steam temperatures were scattered within 5%. Qualitative, discrepancies between the calculational mesh and the actual physical model of the facilities resulted in higher deviations in a fine mesh than in a crude and in the MONJU benchmarks notable surges were observed in the outlet steam and sodium temperatures. The SGTF in-vessel sodium temperature distributions revealed an overestimated rate of heat transfer. Finally, in the superCOPD code the analysis revealed a fixation of the calculated steam outlet temperatures caused by an inadequate treatment of steam superheat. It was recommended that the numerical scheme of the MSG code should be optimized and that the heat transfer correlations should be improved.

Critical Heat Flux Experiments and Correlation in a Long, Sodium-Heated Tube

Critical heat flux (CHF) experiments were performed in the Steam Generator Test Facility (SGTF) at Argonne National Laboratory for application to liquid metal fast breeder reactor steam generators. The test section consisted of a single, straight, vertical, full-scale LMFBR steam generator tube with force-circulated water boiling upwards inside the tube heated by sodium flowing countercurrent in a surrounding annulus. The test section tube parameters were as follows: 10.1 mm i.d., 15.9 mm o.d., material = 2 1/4 Cr–1 Mo steel, and 13.1 m heated length. Experiments were performed in the water pressure range of 7.0 to 15.3 MPa and the water mass flux range of 720 to 3200 kg/m2˙s. The data exhibited two trends: heat flux independent and heat flux dependent. Empirical correlation equations were developed from over 400 CHF tests performed in the SGTF. The data and correlation equations were compared to the results of other CHF investigations.