Prototype Fast Breeder Reactor Research Articles

Accident Sequence Modeling Methodology for External Flood Probabilistic Safety Analysis of Prototype Fast Breeder Reactor

This paper compares two different accident sequence methodologies for external flood probabilistic safety analysis (EFPSA). It is shown that the two methodologies lead to identical expression for CDF with an example. Using accident sequences developed for internal events PSA is recommended for detailed external flood probabilistic safety analysis of prototype fast breeder reactor.

An absolute method for determination of misalignment of an immersion ultrasonic transducer

An absolute methodology has been developed for quantification of misalignment of an ultrasonic transducer using a corner-cube retroreflector. The amplitude based and the time of flight (TOF) based C-scans of the reflector are obtained for various misalignments of the transducer. At zero degree orientation of the transducer, the vertical positions of the maximum amplitude and the minimum TOF in the C-scan coincide. At any other orientation of the transducer with the horizontal plane, there is a vertical shift in the position of the maximum amplitude with respect to the minimum TOF. The position of the minimum (TOF) remains the same irrespective of the orientation of the transducer and hence is used as a reference for any misalignment of the transducer. With the measurement of the vertical shift and the horizontal distance between the transducer and the vertex of the reflector, the misalignment of the transducer is quantified. Based on the methodology developed in the present study, retroreflectors are placed in the Indian 500MWe Prototype Fast Breeder Reactor for assessment of the orientation of the ultrasonic transducer prior to the under-sodium ultrasonic scanning for detection of any protrusion of the subassemblies.

Variation in Residual Stresses due to Thermal Cycling Induced on the Hardfaced Grid Plate in PFBR

In sodium cooled fast breeder reactors, the high operating temperature necessitates hard facing of grid plate to avoid galling and to reduce the wear loss of the base material . Nickel-base cobalt-free alloy called Colmonoy-5 has been chosen as the hard facing material for the Prototype Fast Breeder Reactor (PFBR) grid plate. The grid plate (GP) which is a critical component made of 316 LN SS that holds the core subassemblies are hard faced, to prevent galling and also to minimize wear caused by subassembly insertion/removal and erosion due to high sodium velocity at 670 K. Thermal cycling of hard faced circular grid plate made by Plasma Transferred Arc Welding (PTAW) generates residual stresses due to differential shrinkage of the molten deposit and difference in coefficients of thermal expansion between the deposit and substrate material. In this project the effect of thermal cycling of a nickel-base hardfacing alloy deposited on an austenitic stainless steel grid plate was studied. Finite element analysis of hard faced circular grid plate is performed for obtaining residual stress which includes elasto-plastic analysis. Coupled thermo-mechanical analysis is done for thermal cycling of hardfaced circular gridplate to quantitatively estimate the residual stress.

Evaluation of Mechanical Properties of Modified 9Cr-1Mo Welds Produced by Narrow Gap Hot Wire and Cold Wire Gas Tungsten Arc Welding Processes for 500MWe PFBR Steam Generators

Modified 9Cr-1Mo material is selected as principal material of construction of Steam Generators (SG) for India's first of its kind 500MWe Prototype Fast Breeder Reactor (PFBR). The fabrication of PFBR Steam Generators involves welding of 12mm, 30mm and 90mm thick modified 9Cr-1Mo components which were carried out by combination of hot wire and cold wire Gas Tungsten Arc Welding processes. Hot wire Gas Tungsten Arc Welding (GTAW) is a relatively new process and limited machines/facilities are available in India for welding with this technique. The most important benefit from the use of a hot wire GTAW system is high weld deposition rate and reduced dilution & porosities from the weld deposits. Eventhough welds were meeting the specification requirements during welding procedure qualification, limited information available on hot wire GTAW process in open literature motivated authors for characterization of welds produced by this specialized technique. During welding procedure qualification, it was found that impact test results of weld metal produced by cold wire GTAW process is superior than weld metal produced by hot wire GTAW process. In light of this, detailed investigation & micro-structural analysis is carried out to find out the root cause for variation in impact properties. This paper details the systematic evaluation and interesting observations during characterization of microstructure & mechanical properties of modified 9Cr-1Mo welds produced by hot wire and cold wire GTAW processes.

Evaluation of Mechanical Properties of Modified 9Cr-1Mo Welds Produced by Narrow Gap Hot Wire and Cold Wire Gas Tungsten Arc Welding Processes for 500MWe PFBR Steam Generators

Modified 9Cr-1Mo material is selected as principal material of construction of Steam Generators (SG) for India's first of its kind 500MWe Prototype Fast Breeder Reactor (PFBR). The fabrication of PFBR Steam Generators involves welding of 12mm, 30mm and 90mm thick modified 9Cr-1Mo components which were carried out by combination of hot wire and cold wire Gas Tungsten Arc Welding processes. Hot wire Gas Tungsten Arc Welding (GTAW) is a relatively new process and limited machines/facilities are available in India for welding with this technique. The most important benefit from the use of a hot wire GTAW system is high weld deposition rate and reduced dilution & porosities from the weld deposits. Eventhough welds were meeting the specification requirements during welding procedure qualification, limited information available on hot wire GTAW process in open literature motivated authors for characterization of welds produced by this specialized technique. During welding procedure qualification, it was found that impact test results of weld metal produced by cold wire GTAW process is superior than weld metal produced by hot wire GTAW process. In light of this, detailed investigation & micro-structural analysis is carried out to find out the ischolar_main cause for variation in impact properties. This paper details the systematic evaluation and interesting observations during characterization of microstructure & mechanical properties of modified 9Cr-1Mo welds produced by hot wire and cold wire GTAW processes.

Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5mm and 10.4mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

Markov analysis for time dependent success criteria of passive decay heat removal system

Safety systems deployed in nuclear industry are generally required to operate for a particular mission time. Most of such systems employ redundancy to ensure their high availability over the stipulated mission time. However, availability of a system with redundant configuration depends upon success criteria and is application-specific. The Safety Grade Decay Heat Removal System of Indian Prototype Fast Breeder Reactor is required to operate with different success criteria during the specified mission time on account of steady decline in decay heat produced by the reactor core.In this paper, Markov analysis is carried out to evaluate the availability of the system under both continuous and periodic monitoring schemes. The study estimates the upper bound and lower bound for mean unavailability of SGDHR system for the specified mission time. Sensitivity analysis of the system attributable to important parameters is also carried out. The study provides a comprehensive approach to model scenarios with time dependent success criteria and provides an insight on the factors affecting availability of such systems.

Stabilization of magnet assemblies of permanent magnet sodium flowmeters used in fast breeder reactors

Permanent magnet flow meters (PMFMs) are used to measure the sodium flow in sodium cooled Fast Breeder Reactor Circuits. Prototype fast breeder reactor (PFBR) which is under construction at Kalpakkam is a 500MWe, sodium cooled, pool type reactor. Sodium flow measurement in various loops of the reactor is of prime importance from operational and safety point of view. To measure the flow of electrically conducting liquid sodium, in primary and secondary circuit pipe lines of PFBR, permanent magnet flow meters are used. PMFM is a non-invasive device, which works on the principle of generation of motional EMF by magnetic forces exerted on the charges in a moving conductor. Flowmeters of different pipe sizes ranging from 10mm to 200mm pipe diameter are required for PFBR. Long term performance of the flowmeters mainly depends on stability of permanent magnets used in flowmeters to generate constant magnetic field in stainless steel (SS) pipes. This paper describes the effects of time and temperature on permanent magnet assemblies made of ALNICO-V used in PFBR flowmeters. The stabilization methodology for ALNICO-V permanent magnet assemblies is evolved and established. Loss of magnetic field strength with respect to time and temperatures is determined by experiments and found negligible.

Influence of Welding Speed over Dilution for Circular Grid Plate Hardfaced with Colmonoy-5

Colmonoy is a nickel based alloy hardfaced with SS-316 used in Prototype Fast Breeder Reactors (PFBR) to avoid self-welding and galling between nuclear reactor and grid plate. Colmonoy is coated on the base metal –SS 316 by Plasma transfer arc welding. Due to thermal cycling that exist in reactors de-bonding between colmonoy and base metal occurs which may lead to catastrophic failure of the Prototype fast breeder reactor. Hence, estimation of the fatigue life period of hardfaced colmonoy is mandatory. Before that, finding optimum parameters which influences fusion, weld strength, composition and dilution are mandatory. In order to evaluate the dilution characteristics with respect to welding speed, different specimens were prepared by holding other parameters constant. Prepared specimens were tested for hardness, microstructure and dilution to locate the difference between each specimen. Comparative studies were carried out on the specimen and their parameters like travelling speed, microscopic structure, Hardness and dilution percentage were tabulated. Based on the tabulated results, the best specimen is chosen with respect to its individual parameters.

Design, development and performance testing of fast response electronics for eddy current flowmeter in monitoring sodium flow

The paper discusses the design, development and performance tests of a fast response processing electronics for an eddy current flowmeter that has been developed indigenously to measure sodium flow in the primary sodium pump discharge line, pumping liquid sodium to the core consisting of fuel sub-assemblies in the prototype fast breeder reactor at Kalpakkam, India. Liquid sodium is the main coolant in Fast Breeder Nuclear Reactor. Eddy current flowmeters are deployed in liquid metal cooled fast reactors. The measurement of flow rate of the coolant is important to maintain the overall performance of the pump which in turn is the safety requirement of the system. The faster response of the measuring system is required to ensure the protection of the reactor under pump seizer and discharge pipe rupture. This work mainly focuses on the reduction of response time of signal processing electronic delay.

Experimental evaluation of structural integrity of scram release electromagnet

Prototype fast breeder reactor (PFBR), under construction at Kalpakkam, India, plays an important role in the commercialisation of fast breeder reactors (FBR) in India. It consists of two independent, fast acting and diverse shutdown systems. An electromagnet (EM) immersed in sodium acts as scram release device for the second shutdown system of prototype fast breeder reactor.The inside of EM is sealed from the sodium to achieve the required response time and to prevent the exposure of EM coil to sodium. As the EM response time is an important parameter for reactor safety, the integrity of EM is to be maintained under all anticipated loadings. The EM experiences thermal shocks and thermal stresses during reactor transients such as scram.The dissimilar weld joint present in EM is more susceptible to fatigue failure due to these thermal stresses. Failure of weld joint results in the entry of sodium into the EM, increasing its response time with associated safety implications. In this connection, the structural integrity of EM against thermal shocks was experimentally evaluated in Thermal Shock Test Facility. The EM was subjected to 1000 cycles of thermal shocks, which constitutes 29% of total number of shocks required to qualify the EM for 40 years of reactor operation, thus qualifying it for 11.6yr of reactor operation.The testing has enhanced the confidence level for safe and reliable operation of EM of DSRDM in PFBR. The testing not only qualified the EM for use in reactor but also provided input for licensing the erection of DSRDM on reactor pile. Moreover, it provided a direction for qualifying similar components in future.

Biaxial creep deformation behavior of Fe–14Cr–15Ni–Ti modified austenitic stainless steel fuel cladding tube for sodium cooled fast reactor

Twenty percent cold worked Fe–14Cr–15Ni–Ti modified austenitic stainless steel is used as the cladding tube material for the fuel pins of the Prototype Fast Breeder Reactor in India. Biaxial creep properties of the tubes have been studied at 973K by carrying out creep tests by internally pressurizing the tubes. Hoop and axial components of creep strain were measured and found to be significantly different. For a given gas pressure, steady state hoop rate was higher than the axial rate. Steady state hoop and axial creep rates followed Norton's power law with the same stress exponent n=7. Steady state hoop rates determined from biaxial creep tests agreed with the steady state creep rates determined from uniaxial creep tests. For a thin walled closed tube under internal pressure, significant axial deformation along with hoop deformation is indicative of anisotropic deformation of the material.

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.

Determination of total gas content and its composition in Indian PFBR blanket pellets

Total gas content and its composition are important specifications for sintered nuclear fuel pellets particularly in the case of fast breeder reactor fuels. Most commonly, total gas content and its composition is determined by hot vacuum extraction-quadrupole mass spectrometry (HVE-QMS). A number of parameters in this methodology such as temperature, duration of heating for quantitative extraction of evolved gases, total volume of the system, gas analysis conditions etc. need to be optimized for reliable measurements. In addition, sensitivity factors for various gases like H2, CH4, N2, CO, O2 and CO2 in quadrupole mass spectrometry required for quantification of results have been determined and validated employing reference gas mixtures of known composition. Employing these optimized conditions total gas content and its composition in blanket pellets (uranium oxide pellets) of Indian prototype fast breeder reactor was determined employing HVE-QMS. The relative expanded uncertainty (at a coverage factor k = 2) in the measurement of total gas content excluding hydrogen was estimated as per ISO guidelines and it was found to be 9.2 %.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

The 500MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Dynamic modeling of steam water system of prototype fast breeder reactor using RELAP code

The safety of a fast breeder reactor based nuclear power plant with an intermediate coolant loop between the primary and tertiary circuit, depends on the correct functioning of actions initiated by the balance-of-plant (tertiary) systems. Comfortable time will be available for such actions to ensure reactor safety. Perturbations in the Balance-of-Plant (BoP) influence the transient sequence of safety-relevant parameters of the plant in a benign manner. However, for complete and realistic prediction of transient behavior of the whole plant, dynamic models for BoP systems are required to be developed. This paper describes modeling of BoP system of PFBR using RELAP5/MOD 3.4 code. Some of the important transients in the BoP system, which are managed by reactor power setback procedure have been analysed using this code to verify the effectiveness of the procedure adopted.

Fault tolerant distributed real time computer systems for I&C of prototype fast breeder reactor

Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation & control (I&C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I&C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network.

Performance evaluation of a resonance ionization mass spectrometer developed for the FFDL system of fast reactors

To prevent a fuel failure event from becoming a serious radiation accident, sodium-cooled fast reactors are equipped with a system for failed fuel detection and location (FFDL). The FFDL instrument employed in the prototype fast breeder reactor Monju is based on the gas tagging method, in which precise and accurate measurements of krypton and xenon isotope ratios (78Kr/80Kr, 82Kr/80Kr and 126Xe/129Xe) must be performed in a short time. Burnup measurements also contribute to accurate determination of 82Kr/80Kr. We have developed a highly sensitive resonance ionization mass spectrometer for the isotopic analyses, which uses resonance ionization of Kr and Xe atoms by a pulsed laser at wavelengths of 216.7 and 249.6 nm, respectively. In evaluating the performance of our spectrometer, we find that systematic errors caused by isotope shifts can be reduced to negligible levels, and that statistical errors of 3% at a nuclide concentration of 7 ppt can be achieved with a single measurement time of about 40 minutes for each Kr and Xe isotope ratio. This means that, within one hour, about 200 fuel assemblies can be individually identified with a probability of 99%, verifying the applicability of our spectrometer to the FFDL system of fast reactors.

An Overview of Welding Aspects and Challenges during Manufacture of Intermediate Heat Exchangers for 500mwe Prototype Fast Breeder Reactor

Prototype Fast Breeder Reactor (PFBR) is first of its kind 500MWe pool type, sodium cooled nuclear reactor which is presently in advanced stage of construction at Kalpakkam, India. Intermediate Heat Exchanger (IHX) is a shell and tube type, counter current, sodium to sodium heat exchanger. IHX is a very important, massive, over dimensional (∼42 tones in weight, ∼2 meters in diameter and ∼18 meters in length) and critical component of reactor, as it transfers heat from the radioactive primary circuit sodium to non-radioactive secondary circuit sodium forming the boundary between these two circuits. The principal material of construction of IHX is austenitic stainless steel grade 316LN. Each IHX has 3600 nos. of straight seamless tubes which are rolled and then welded to the either ends of top and bottom tubesheets by autogenous pulsed Gas Tungsten Arc Welding (GTAW) process. The shell welding around the tube bundle is extremely difficult and challenging task due to small gap between the tube bundle and shell inside diameter. Special arrangements were made to avoid arc strike or fusion on the tube during shell welding around the tube bundle.A mechanical hardfaced seal arrangement at the interface of the IHX outer shell and the inner vessel stand pipe is the chosen design concept to ensure leak tightness in the IHX penetrations. Based on radiation dose rate & shielding considerations during maintenance, handling and decommissioning, nickel based ENiCr-B hard facing alloy (Colmonoy-5) was chosen to replace the traditionally used cobalt based stellite alloys to improve the resistance to high temperature wear, especially galling of mating surfaces in sodium environment. The more versatile Plasma Transferred Arc Welding (PTAW) and Gas Tungsten Arc Welding (GTAW) were used for deposition of Colmonoy-5. The hardfacing on the seal ring is really challenging task, as the diameter is too large and thickness & width is too small and which has hardfacing on all mating surfaces. Extensive varieties of trials were conducted on the mockup to optimize the hardfacing process along with heat treatment cycle to obtain minimum distortion deploying special tools and fixtures.Intermediate Heat Exchanger involves 10mm and 28mm thick borated stainless steel components around the tube bundle and bottom portion confirming to ASTM 887, Type 304 B4, Grade B classification to reduce the radiation level in secondary sodium system while flowing through the IHX. Various types of welding trials were conducted on the test coupons for welding procedure qualification of borated stainless steel components using versatile grades of welding consumables. More than hundred nos. of trial coupons welded using 10 varieties of welding consumables have shown fissures and cracks in weld, as these welding consumables found not having good compatibility with SS 304 B4, Grade B borated steel. Special development work was taken-up for the first time in the country by developing a special grade ∼1% borated welding consumable (GRINOX-308BRN electrode) specifically for welding of borated steel components of IHX. This paper highlights on welding aspects and challenges during fabrication of Intermediate Heat Exchangers for 500MWe Prototype Fast Breeder Reactor.

Comparison of Microstructure and Properties of Modified 9Cr-1Mo Welds Produced by Narrow Gap Hot Wire and Cold Wire Gas Tungsten Arc Welding Processes

Modified 9Cr-1Mo material is used as principal material of construction of Steam Generators (SG) for India's first of its kind 500MWe Prototype Fast Breeder Reactor (PFBR).The fabrication of Steam Generators is carried out by combination of hot wire and cold wire Gas Tungsten Arc Welding (GTAW) processes. Eventhough welds were meeting the specification requirements, limited information available on hot wire GTAW process in open literature motivated authors for characterization of welds produced using this specialized technique. The impact properties at 0°Cof modified 9Cr-1Mo welds produced by hot wire GTAW processis found inferior compared to welds produced by cold wire GTAW process. The properties such as tensile strength, hardness, impact toughness at +18°C were found almost comparable for the welds produced by both the welding processes. This paper characterizes the microstructure and various mechanical properties of modified 9Cr-1Mo welds produced by narrow gap hot wire and cold wire GTAWprocesses.