Safety features of the PFBR18 September 2013
India’s Prototype Fast Breeder Reactor project is now moving toward integrated commissioning. Systems for sodium coolant leak prevention, and safety systems to protect the reactor from extreme events, such as tsunami surges and 1000-year cyclone rainfall, are reviewed. By Dr. Prabhat Kumar and Aravinda Pai
Prototype Fast Breeder Reactor (PFBR) is a 500 MWe pool type, sodium-cooled nuclear reactor, which is presently in an advanced stage of construction on the southeastern Indian coastline at Kalpakkam. India has chosen the closed fuel cycle option in view of its phased expansion of nuclear power generation extending through the latter stages of its programme, whereby the full energy potential of uranium and thorium can be exploited efficiently. In fast breeder reactors (FBRs), fission of plutonium-239 enables sizable and gainful conversion of around 70% fertile uranium-238 to fissile plutonium-239 and conversion of fertile thorium-232 to fissile uranium-233 in the radial blanket, which can provide adequate fissionable material to sustain the third stage Indian Nuclear Power Programme. Even though the venture is a prototype, the approach to the project is on par with construction of any commercial nuclear reactor. On completion, this reactor will not only produce 500 MWe power, but also pave the way for abundant power production using indigenous materials and technology and will open up the gateway for the country's future energy security. The primary objective of PFBR is to demonstrate the techno-economic viability of fast breeder reactors at an industrial scale. The fast breeder reactors would make effective utilization of depleted uranium available in the country, and use plutonium as a fuel with significant reduction in radioactive waste.
The positive experience of achievements in the field of new technology is a matter of pride to the nation. Indian industries have accepted the manufacturing challenge by fabrication of critical, over-dimensioned reactor equipment, which are complex to construct. In doing so, industries have significantly improved their own technological capabilities. All the reactor components were manufactured indigenously in India with close tolerances, thereby making PFBR a truly national project. Many new and innovative procedures were developed for installation of critical equipment meeting all the stringent specification requirements. It is heartening that the design features have been correctly translated into manufacturing and erection. Various measures have been taken by the management to protect the site from natural calamities such as earthquakes, tsunami and cyclones. This paper highlights these aspects of the PFBR.
Prototype Fast Breeder Reactor consists of Primary Sodium Circuit (PSC), Secondary Sodium Circuits (SSC), Safety Grade Heat Removal Circuits (SGDHRC) and a steam-water circuit. The primary sodium circuit removes the nuclear heat generated in the core and transfers it to the secondary sodium circuits through intermediate heat exchangers (IHXs). The secondary sodium circuits, in turn, transfer the heat to a steam/water circuit (SWC) through steam generators (SGs).
The primary sodium circuit consists of core, primary sodium pumps, IHXs, primary pipe connecting the pumps and the grid plate, and is contained in a single large-diameter main vessel. The main vessel is welded at the top to the roof slab. The core subassemblies are supported on the grid plate, which in turn is supported on the core support structure. Liquid sodium at 397°C is circulated by two primary sodium pumps through the core and in turn is heated to 547°C.
The non-radioactive secondary sodium is circulated through two independent secondary loops, each having a secondary sodium pump, two IHXs and four SGs. The primary and secondary sodium pumps are vertical, single-stage and single-suction centrifugal type with variable-speed AC drives. The steam generator is a vertical, once-through, shell-and-tube type heat exchanger with liquid sodium flowing in the shell side and water/steam flowing in the tube side.
The boundaries of the sodium systems are designed to have extremely low probabilities of leakage, rapidly propagating failure, and gross rupture under the static and dynamic loads expected during various operating conditions. Therefore, their design includes considerations of degradation of material properties (for example, effect of sodium, temperature and irradiation), transients, residual stresses, flaw size, and so on.
Site safety considerations
PFBR required extensive study of site prior to launch of project. The air, water and earth had to be studied; their parameters were analysed without disturbing them. Sea currents over three seasons, water temperature variations over various depths and ranges, and zoological and botanical features of coastal marine life are some of the features of marine study. Wind speed and direction, humidity, rainfall and particulate matter in the air were studied. Extensive geological study of the site, analysis of strata below the structures to be constructed, and establishment of a network of micro-seismic stations are a few of the many considerations for selection of the PFBR site.
The marine structures of PFBR (including the jetty, undersea tunnel, shore protection and outfall channel, which are either submerged or in the splash zone) are subjected to harsh conditions. Proper selection of concrete, establishing a concrete cover without cracks, and avoiding the corrosion of reinforcement bars (coated and uncoated) were important parameters that had to be considered, as these structures would face nature's fury during a cyclone, tsunami and/or earthquake.
In fact, the PFBR faced the December 2004 tsunami when the second pour of the nuclear island raft concreting was in progress. At that time, the pit excavated for construction of the nuclear island connected building was completely inundated. The water rise during that tsunami was measured to be 4.1 to 4.8 metres at different places on site.
A seawall on the beach has been created to avoid beach erosion during future tsunami and cyclones. A wall made of stone boulders has been created as a barrier to break the energy of any future tsunami wave. The height of tsunami wall built after 2004 tsunami was 5.1 metres above the mean sea level. Following the Fukushima incident, the height of the wall is being raised to 9 metres above mean sea level. The nuclear island finished floor level is raised to 9.6 metres above the mean sea level.
An important observation made at PFBR site during the December 2004 tsunami was large quantity of seashells found spread on a 10-metre-high hill around 250 metres from the shore. This hillock, formed due to the excavated earth from nuclear island area, was free of shells prior to the tsunami. It was proposed that run-up of water from sea during the 2004 tsunami might have been responsible for transport of the shells. This conclusion was reinforced by the March 2011 accident at Fukushima Daiichi, which indicated that a high splash of water can reach heights greater than a tsunami wave, and reach a greater distance inland. Therefore, as a precautionary measure, the PFBR site will feature leak-tight doors and leak-tight pipe and cable penetrations in Nuclear Island Connected Building (NICB) two to three metres above finished grade level, to avoid splash water entry into NICB.
Storm drains are adequately designed to take care of site rainfall from a 1000-year cyclone. In addition, the drainage system will be able to account for the water overtopping the tsunami wall due to splash, if any. The storm drains from the plant penetrate the tsunami wall and discharge the water to sea. Non-return valves are planned to be provided in the storm drain channels to avoid floodwater entry to plant.
All of the routes by which seawater could enter the plant during future tsunami were analyzed. The seawater intake tunnel can bring water from a surging sea through the pump house. Water may travel to various places inside the power island building through tunnels and trenches which connect the seawater pump house with other buildings. The pipe and cable tunnel and trenches emanating from the seawater pump house will thus be sealed for water tightness.
The Nuclear Island Connected Building, see Figure 1, is in rectangular shape, containing eight different buildings laid on a common raft to reduce the magnitude of structural response under the seismic loads from safety considerations. As a precautionary measure, the turbine building is located such that the missile trajectory of a turbine blade is outside the safety-related buildings and the stack. Various heavy density concretes (3600 kg/cubic metre, 3860 kg/cubic metre, 4100 kg/cubic metre, 4200 kg/cubic metre) were deployed for concreting of reactor walls, top shield and cells that contain radioactive equipment in PFBR for additional radiation shielding. Radiometric tests were carried out using a cobalt-60 source to evaluate the absence of voids and to ensure homogeneity of steel shot in the concrete pours.
Core safety features
A pool-type concept is adopted for the core due to the inherently high thermal inertia of the large mass of sodium in the pool, which eases the removal of decay heat.
The main vessel has no penetrations, leading to high structural integrity and no radiation damage. High operating temperature of various systems causing high stresses are minimized by designing the thin walled vessels/structures. The main vessel is surrounded by the safety vessel, closely following the shape of main vessel, with a nominal gap of 300 mm to permit the robotic and ultrasonic inspection of the vessels and to keep the sodium level above the inlet windows of IHX, ensuring continued cooling of the core in case of a leak in the main vessel.
PFBR has 181 fuel sub-assemblies arranged in a triangular pitch. Alloy D9 in 20% cold worked condition (20CW D9) has been chosen for clad and wrapper tubes of PFBR due to its high resistance for swelling and irradiation creep. All the core subassemblies are supported on the grid plate, which in turn is supported on the core support structure. As a precautionary measure, a core catcher is provided just below the core support structure. The core catcher is designed to accommodate the meltdown of seven subassemblies, to prevent core debris reaching the main vessel, and allow core cooling by natural convection of sodium.
The PFBR also has enhanced negative reactivity in the core, which is an important inherent safety feature of the reactor.
The primary liquid sodium is radioactive. Therefore, radioactive primary sodium is not used directly to produce the steam. The secondary sodium circuit exists in between primary sodium circuit and steam-water circuit to ensure that radioactive primary sodium is always inside the Reactor Containment Building (RCB). In addition, the secondary sodium circuit in between primary sodium and steam-water circuit is envisaged to prevent carryover of hydrogenous materials and reaction products (water, steam, hydrogen, sodium hydroxide) into the core, in case of a sodium-water/steam reaction incident in the steam generators.
A pressure differential is maintained in intermediate heat exchangers between the primary sodium coolant and secondary sodium coolant, so that any leakage would flow from non-radioactive secondary sodium to the radioactive primary sodium, and not vice versa. This feature also contains radioactive primary sodium inside the reactor containment building, and helps to reduce radiation dose rate for the reactor operation personnel and population around the reactor.
PFBR has two independent, reliable, fast-acting, automatic, diverse shutdown systems, comprising sensors, logic circuits, drive mechanisms and neutron absorber rods with boron-carbide (B4C) pellets. PFBR has nine Control & Safety Rod Drive Mechanisms (CSRDM) and three Diverse Safety Rod Drive Mechanisms (DSRDM). The functions of CSRDM are to control of reactor power and facilitate start-up and controlled shut down of the reactor. In addition, the CSRDM is used to shut down the reactor in abnormal conditions. The function of DSRDM is to facilitate shutdown of the reactor in abnormal conditions in case of non-availability of CSRDM.
Materials for reactor components were selected according to factors including operating conditions, availability of design data in nuclear codes, ease of fabrication, international experience and safety. The principal PFBR construction material is austenitic SS316LN/SS304LN. This is a low-carbon stainless steel chosen to ensure freedom from sensitization during welding of the components. This steel has excellent high temperature mechanical properties and good compatibility with liquid sodium coolant. The manufacture of Nuclear Steam Supply System (NSSS) components is carried out in separate nuclear clean hall conditions for quality assurance. Various control and quality assurance measures are essential for critical reactor equipment during each and every stage of design, raw material procurement, welding, fabrication, non-destructive examinations, testing, handling, installation and post-installation preservation to ensure high degree of reliability for the design service life of 40 years. Like any other engineering mega-project, PFBR has extensive civil, mechanical, electrical and instrumentation activities. The quality requirements in every arena of PFBR are far in excess of conventional engineering project.
All sodium pipelines inside the PFBR Reactor Containment Building (RCB) are double-walled and provided with a hot guard pipe and are inerted with nitrogen. Rigorous Foreign Material Exclusion (FME) practices are adopted during each and every stage of fabrication to ensure absence of foreign material inside the piping before helium leak testing. Boroscopic inspection of pipe lines is carried out in the inaccessible areas to ensure FME in the complete piping systems.
During component erection and installation, utmost care and a variety of control measures were taken to ensure specified horizontality, elevation, alignment, co-planarity, axis orientation and co-axiality of centres of various reactor components. Load testing of transportation/handling structures is done promptly for all the equipment, simulating the actual job handling conditions before erection of equipment for purposes of validation.
PFBR has four emergency diesel generator sets, each of 4.5 MVA capacity. Two DGs are located on the east side of Nuclear Island Connected Building (NICB) and two are located on the west side. While the eastern DGs are on the sea side, around 300m from the coastline, the DGs on the west side, in the shadow of NICB, are around 450 metres from the coast. Therefore, the risk of a tsunami surge disabling the west side DG sets is prevented. The two DGs on the west side are capable of ensuring safety during a Station Black Out (SBO) scenario. Yet as an additional precaution, it is planned to have two mobile DGs as well.
Decay heat removal systems
Safety, quality, diversity and reliability are of supreme importance in a nuclear power plant. After reactor shutdown, the residual heat (mainly fission product decay heat) in PFBR is removed through special decay heat removal condensers connected to the steam generators that bypass the turbine-generator. This system is known as the Operational Grade Decay Heat Removal System (OGDHRS). In order to improve the reliability of Decay Heat Removal (DHR) function, a fully-passive Safety Grade Decay Heat Removal System (SGDHRS) consisting of four special thermo-siphon loops in natural convection mode is provided. This SGDHR system uses atmospheric air and does not require any power for removal of decay heat from the reactor. A simple functional diagram of SGDHRS is shown in Figure 2. Each loop consists of a sodium-to-sodium heat exchanger (DHX) dipped in the primary sodium hot pool (completely independent of the IHX) and a sodium-to-air heat exchanger (AHX).
There are two types of AHX and DHX design concepts used in different SGDHR loops to avoid common cause failure: two SGDHR loops have AHX and DHX of one design, and the remaining two loops have AHX and DHX of another design. SGDHRS will operate when there is a failure of OGDHRS due to component failures in the secondary or steam-water circuit, or in the event of Loss of Offsite Power (LOP). Leak Before Break (LBB) criteria is applicable for main sodium piping of safety-grade decay heat removal circuits. Since the DHR function failure frequency shall be less than 10-7/year, highly-reliable DHR systems are provided for PFBR.
AHX casing is provided with two dampers in the inlet and two dampers in the outlet with diversity in damper drives to enhance the reliability of circuit activation. On each side, one damper is motor-operated and the other damper is pneumatically-operated with dedicated class 2 power supply. Provisions are made to operate the dampers manually if automatic and remote manual opening fails. During reactor operation, all four loops are in a poised state with all dampers are in a slightly-open position allowing about 35% nominal air flow in the SGDHR loop, with a heat loss of about 2 MWt. This helps quickly establish natural convection flow when the dampers are fully opened for decay heat removal through SGDHR system. In DHX, the top portion of the shell side is perforated for a length sufficient to permit primary sodium entry in the shell side even in the event of a sodium leak in the main vessel. This arrangement ensures effective decay heat removal through SGDHR system even in the event of a main vessel sodium leak.
Elaborate planning and various mock-ups were done for manufacture, inspection, testing and erection of over-dimensioned and critical PFBR equipment. Despite its first-of-a-kind technological challenges, the construction work has progressed well. Almost all major reactor equipment such as Main Vessel, Safety Vessel, Inner Vessel, Roof Slab, Core Catcher, Core Support Structure, Grid Plate, Steam Generators, Intermediate Heat Exchangers, Decay Heat Exchangers, Large Rotating Plugs and Small Rotating Plugs have been placed in the designated location in the reactor. All the core subassemblies, that is, outer boron carbide, steel shielding, storage location sub-assemblies, dummy fuel sub-assemblies and dummy blanket sub-assemblies that are required to be installed prior to sodium filling in the reactor, have been installed on the Grid Plate. The Primary Tilting Mechanism and Primary Ramp, a part of the Inclined Fuel Transfer Machine, has been fixed to the Grid Plate. Erection of fuel handling equipment at various locations of fuel building is nearing completion. Sodium system piping erection work is in progress.
Erection of turbine equipment is in an advanced stage of completion. Major equipment such as generator stator and rotor, live steam reheater, steam water separator and high and low pressure turbine modules have been erected.
Pre-commissioning activities have been commenced for the completed systems; they are at various stages for the conventional auxiliary systems. The underwater trolley and a few other fuel handling systems have been commissioned. All four of the diesel generators have been commissioned. Commissioning of the 230 KV gas insulated switchyard and 6.6KV switchgear boards has been completed. All six outgoing lines of the transmission system from the switchyard have been commissioned and are kept charged and connected to the southern India grid. Instrumentation works for various systems are in progress.
PFBR project has reached an advanced stage of construction and is moving towards integrated commissioning. The project was 95% physically complete by the end of May 2013. Based on the present progress, all efforts are being made to attain criticality in or before September 2014.
The specification requirements, dimensional tolerances and acceptance criteria for PFBR are far more stringent than ASME or many other international standards. Very high standard quality control and quality assurance during design, civil construction, raw material procurement, welding, fabrication, inspection, testing and erection has given confidence in the possibility of trouble-free service from Prototype Fast Breeder Reactor for the design service life of 40 years. ¦
Dr. Prabhat Kumar, distinguished scientist and chairman & managing director, and Aravinda Pai, scientific officer (long range planning & strategies),
Bharatiya Nabhikiya Vidyut Nigam Limited (BHAVINI), Department of Atomic Energy, Kalpakkam-603 102, India