S.C. Chetal

LESSONS LEARNED FROM SODIUM-COOLED FAST REACTOR OPERATION AND THEIR RAMIFICATIONS FOR FUTURE REACTORS WITH RESPECT TO ENHANCED SAFETY AND RELIABILITY

Abstract Eighteen sodium-cooled fast reactors (SFRs), a number that includes reactors in operation or shut down, have provided 388 reactor-years of operating experience to date. This paper summarizes the important incidents related to fast reactor sodium components and systems. The solutions incorporated, based on experience, analysis, experimental tests, and research and development for past and current SFRs, are described. The paper also describes lessons learned for future SFRs.

Influence of mis-match of weld and base material creep properties on elevated temperature design of pressure vessels and piping

The stress and strain concentrations developed at the weldments during the long time operation of pressure vessels and piping at high temperature due to the mis-match in the creep properties of weldment constituents (weld, heat affected zone and base metal) are estimated using detailed finite element analysis. Three materials, viz. 2.25Cr 1Mo, SS 316 LN and modified 9Cr 1 Mo which are the most commonly used materials in the nuclear and thermal power plants are considered. A longitudinal seam weld with single and double V (X) configurations are analysed. Parametric studies have been done on weld angle and stresses. Based on the analysis, critical locations and the maximum stress concentration factors in the weldments for the above materials are identified. The weld design procedures of the currently used pressure vessel and piping codes are commented. The importance of ductility based failure criteria is emphasised.

Design, Development, Testing and Qualification of Diverse Safety Rod and Its Drive Mechanism for a Prototype Fast Breeder Reactor

Prototype fast breeder reactor is U-PuO 2 fueled sodium cooled pool type fast reactor and it is currently under construction at Kalpakkam, India. Prototype fast breeder reactor is equipped with two independent fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms, and neutron absorbing rods. The two shutdown systems of prototype fast breeder reactor are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of prototype fast breeder reactor are called as diverse safety rods (DSRs) and their drive mechanisms are called as diverse safety rod drive mechanisms (DSRDMs). DSRs are normally parked above active core by DSRDMs. On receiving scram signal, the electromagnet of DSRDM is de-energized and it facilitates fast shutdown of the reactor by dropping the DSR into the active core. For the development of prototypes of DSR and DSRDM, three phases of testing, namely, individual component testing, integrated functional testing in room temperature, and endurance testing at high temperature sodium, were done. The electromagnet of DSRDM has been separately tested at room temperature, in furnace, and in sodium. Specimens simulating the contact conditions between electromagnet and armature of DSR have been tested to rule out self-welding possibility. The prototype of DSR has been tested in flowing water to determine the pressure drop and drop time. The functional testing of the integrated prototype DSRDM and DSR in aligned and misaligned conditions in air/water has been completed. The performance testing of the integrated system in sodium has been done in three campaigns. During the third campaign of sodium testing, the performance of the system has been verified with 30 mm misalignment at various temperatures. The third campaign has qualified the system for 10 years of operation in reactor. This paper presents design, development, testing, and qualification of the prototype DSR and DSRDM. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual and detailed design features are explained with the help of figures. Details on material of construction are given at appropriate places. Test plans and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are briefed. Brief explanation of test setups and typical test results are also given.

Structural integrity assessment of reactor assembly components of a pool-type sodium fast reactor in a core disruptive accident - II: Analysis for a 500-MW(electric) prototype fast breeder reactor

Abstract A core disruptive accident, considered a beyond-design-basis accident, for the 500-MW(electric) capacity Prototype Fast Breeder Reactor (PFBR) is analyzed using the FUSTIN in-house computer code. In order to have a good understanding of the complicated loading mechanisms and sequences, the analysis studies the effects of introducing internals in the main vessel. Further, the structural integrity of heat exchangers—which are important for decay heat removal during postaccident conditions - was demonstrated with tests that were conducted on a 1/13th scaled-down mock-up; a suitable low-density explosive was developed and characterized to simulate nuclear energy release characteristics. The tests have indicated relatively smaller displacements and strains in the vessel, compared to numerical predictions, and the structural integrity of the decay heat exchangers including tubes was demonstrated. Thus, the reactor assembly components meet the safety criteria specified for PFBR with comfortable margins for the specified mechanical energy release of 100 MJ.

Design and Development of Diverse Safety Rod and Its Drive Mechanism for PFBR

Prototype Fast Breeder Reactor (PFBR) is U-PuO2 fuelled sodium cooled Pool type Fast Reactor and it is currently under advanced stage of construction at Kalpakkam, India. The Fast Breeder Test Reactor (FBTR) which is the only fast reactor currently operational in India is having only one shutdown system. However the IAEA and Atomic Energy Regulatory Board (AERB) Guide Lines call for two independent fast acting diverse shutdown systems for the present generation reactors. Hence PFBR is equipped with two independent, fast acting and diverse shutdown systems. A shutdown system comprises of sensors, logic circuits, drive mechanisms and neutron absorbing rods. The two shutdown systems of PFBR are capable of bringing down the reactor to cold shutdown state independent of the other. The absorber rods of the second shutdown system of PFBR are called as Diverse Safety rods (DSR) and their drive mechanisms are called as Diverse Safety Rod Drive Mechanisms (DSRDM). DSR are normally parked above active core by DSRDM. On receiving scram signal, Electromagnet of DSRDM is de-energised and it facilitates fast shutdown of the reactor by dropping the DSR in to the active core. This paper presents chronological design and development of the prototype DSR and DSRDM starting from the design specifications. Salient design specifications for both DSRDM and DSR are listed initially. The conceptual & detailed design features are explained with the help of figures. Various important design options considered in the initial design stage, choice of final design along with brief explanation for the particular choice are also given for some of the important components. Details on material of construction are given at appropriate places. Details on various analysis such as large displacement analysis for buckling, bending analysis for determining reactive forces and friction in the mechanism, thermal stress analysis of electromagnet during scram, flow induced vibration analysis of DSRDM and DSR and hydraulic analysis for estimating the pressure drop and drop time of DSR are also given. Test plans for design verification, manufacturing and shop testing experience of prototype systems, and criteria for endurance testing in sodium for qualification of DSRDM and DSR for operation in reactor are also briefed.Copyright

Comparison of SS 304 LN and 316 LN as choice of structural material for primary pipe in fast breeder reactors

In pool type FBRs primary sodium pumps (PSP), operating in parallel, circulate the sodium through the core to remove the nuclear heat. The pumps suck the sodium from cold pool and supply it to a spherical header at the bottom; subsequently the sodium flows through pipes from the spherical header into the grid plate, before entering the core subassemblies. Under normal operating condition, each pipe is subjected to 0.8 MPa internal pressure of sodium flowing at 670 K. As the pipe is located in the cold pool, it is at an isothermal temperature of 670 K during normal operation. The pipe is subjected to hot shock during two thermal transients Viz. offsite power failure (160 times) and loss of Steam water system (47 times). Both these events lead to Safety Grade Decay Heat Removal (SGDHR) operation and the cold pool reaches 791 K during offsite power failure and 816 K during loss of steam water system. Even though the normal operating temperature for the primary pipes conveying sodium from spherical header to grid plate is below creep regime, creep damage occurs due to the governing hot shocks discussed above. One of the criteria for selecting the material is the creep fatigue damage for the longer design life (40 years for the Prototype Fast Breeder Reactor under construction and 60 years for the future reactors). The material choice for out of core reactor assembly components are generally austenitic stainless steels (SS316LN or SS304LN) except for the top shield. In this paper, the choice between SS 304 LN Vs SS 316 LN for primary sodium pipe with respect to the governing failure mode of creep-fatigue interaction is addressed.

Thermal Hydraulic Analysis of Severe Accident in PFBR

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.© 2010 ASME

Thermal Hydraulic Analysis Towards a Robust Design of Leak Collection Tray for Pool Type Sodium Cooled Fast Reactors

To protect the sodium cooled FBR plant against the hazardous effects of sodium leak into the ambient, one of the passive protection devices used is the Leak Collection Trays (LCT) below the secondary sodium carrying pipelines in the Steam Generator Building (SGB). The design of LCT is based on immediate channeling of burning liquid sodium on the funnel shaped ‘sloping cover tray’ to the bottom ‘sodium hold-up vessel’ in which self-extinction of the fire occurs due to oxygen starvation. In the secondary heat transfer circuits of FBRs, leakage of liquid sodium from the pipelines is postulated as one of the design basis accidents with probability of occurrence at 10−2 per reactor year. LCT collect the leaked sodium in a hold up vessel, suppress the sodium fire due to oxygen starvation and guide the sodium to an inerted ‘sodium transfer tank’ located at the bottom most elevation of the SGB. The procedure of draining the leaked sodium into the transfer tank has been envisaged as a defense in depth measure against the handling of un-burnt sodium and to guard against larger leak rates than that can be handled by the LCT effectively. Towards this, a network of carbon steel pipelines are laid out connecting all the LCT and the transfer tank through headers in strategic locations, each having a fusible plug. The fusible plug separates the air environment in LCT and argon environment in sodium transfer tank. Woods metal is the preliminary choice for the fusible plug. It is an alloy of 50% Bi, 25% Pb, 12.5% Sn and 12.5% Cd with a melting point of 72°C. The transfer tank is filled with argon at ∼ 0.03 bars-g pressure. Both the header and the tank are at room temperature during normal conditions. Leaked sodium by virtue of its high temperature has to heat up the fusible plug to melt the same and drain into the transfer tank. Transient thermal hydraulic investigations have been carried out to predict the fusing characteristics of woods metal plug. The numerical results have been validated against analytical solutions for idealized conditions. Detailed parametric studies have been carried out with plug thickness as a parameter. It is established that effective melting of the plug and trouble free draining of the leaked sodium is possible for a 3 mm thick fusible plug.Copyright

ICONE15-10859 EXPERIMENTAL EVALUATION OF INTEGRITY OF FBR CORE UNDER SEISMIC EVENTS

The core of Prototype Fast Breeder Reactor (PFBR) is designed to produce 1250 MWt at full power. PFBR is under construction at Kalpakkam, India. In PFBR, the core is of free standing type and one of the major safety criteria for the design of core subassemblies is that the integrity of the core subassemblies should not be impaired and they should not be lifted up from the grid plate even during seismic condition. The net downward force acting on the grid plate is less than the weight of the subassembly due to the hydraulic lifting forces acting on it. Experimental analysis has been carried out to ensure that the subassembly does not get lifted off due to vertical seismic excitation. This paper gives the details of the methodology adopted for the experimental seismic analysis carried out on a core subassembly and the upward displacement of the subassembly under the combined effect of upward fluid force and vertical seismic excitations.

Design and Qualification of Control and Safety Rod and Its Drive Mechanism of Fast Breeder Reactor

Prototype Fast Breeder Reactor (PFBR) has two shutdown systems. The absorber rod of the first system is called Control & Safety Rod (CSR). Control & Safety Rod Drive Mechanism (CSRDM) facilitates start-up & controlled shut-down of reactor and control of reactor power by raising and lowering of CSR and shutdown of the reactor on abnormal conditions by rapid insertion of CSR into the core, i.e., by scram action. After the detailed design and analysis of CSR and CSRDM, they were qualified in two stages. In the first stage, the critical assemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals, were tested individually simulating the operating conditions of the reactor and the design parameters were fine-tuned. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts. In the second stage, prototype CSRDM and CSR were manufactured and subjected to various functional tests in air, in hot argon and subsequently in sodium simulating the operating conditions of the reactor. Tests were carried out keeping CSRDM and CSR at aligned condition and with the possible misalignment between them. The performance was checked and recorded maintaining the temperature of sodium starting from 473 K to 823 K. Then the system was subjected to endurance tests. The results show that the performance of CSRDM and CSR is satisfactory and there is no significant change in the performance during endurance testing.© 2006 ASME