[en] The activities are aimed at achieving continual enhancement of Nuclear and Radiation Safety, Development/Validation of New Reactor Systems, Reliable Operation and Reduction in Operational and Construction costs of Nuclear Power Plants, Indigenization and Construction time reduction

[en] Directorate of Technology Development in NPCIL has been tasked with developmental activities related to the four verticals of R and D facilities, Remote Tooling, Indigenization and Construction time minimization. The presentation brings out an overview of these activities with greater emphasis on post Fukushima related improvements and activities related to design verification of first of a kind process systems for the under construction 700 MWe units. These include experimentation : On containment spray to arrive at design finalization and design verification; On Passive Decay Heat Removal System (PDHRS); For performance testing and qualification of Passive Catalytic Recombiner Device (PCRDs); For design and performance verification of Containment Filtered Vent Systems (CFVS) to be provided in 700 MWe Units as well as back fitting in other operating NPPs. Other areas include: Equipment qualification activities in the LOCA chamber as well as other setups; The remote tooling related activities include salient challenges which are overcome in the operating units as well as channel inspection and steam generator inspection campaigns; The indigenization activities would cover the various specific items which have been developed to strengthen the supply chain and build up requisite capabilities for the future growth of Nuclear Program. Construction time minimization related activities would cover the significant modularization being pursued along with other methodologies which can facilitate quicker construction of NPPs with enhanced quality and safety. (author)

[en] Two methods of condensation namely reflux and spray are employed in the bleed condenser (BCD) which is an equipment of the primary heat transport (PHT) of the Pressurized Heavy Water Reactor (PHWR) being constructed at NAPP. The design of the NAPP BCD with respect to these two modes of heat transfer is presented. (author). 9 refs., 3 figs

[en] A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

[en] Full text: In a pressurised heavy water reactor (PHWR) nuclear power plant, those transients where large thermal shrinkage in the primary heat transport (PHT) volume takes place can result in the lowering of PHT system pressure. During some of the reactor trip incidents as reported from the operating units, the minimum PHT pressure obtained has resulted in the unwanted line up of emergency core cooling system (ECCS). This line up is undesirable from the consideration of thermal cycling and hot/cold inventory transfers. A reduction of the dip in the PHT system pressure can be achieved by minimizing the PHT temperature fall and consequent system shrinkage during such transients. To minimize the heat transfer from the primary coolant to the secondary coolant in the steam generator (SG), the SGPC (steam generator pressure controller) set point has to be increased during such incidences. This will lead to a higher steam pressure and temperature in the SG thus reducing the temperature difference and heat transfer between primary and secondary coolant. Another option is to increase the net feed to the PHT volume i.e. by increasing the holdup inventory in a fixed boundary. This can be achieved by having a pressuriser volume riding over the pressure system boundary. In this paper, it is proposed to study the effect of a small size pressuriser of 10 and 6 M³ in the low pressure transients. This would also serve the important design objective of minimizing the reactor trips on PHT low pressure. This transient study has been carried out using the integrated system process dynamics analysis code. This code incorporates the mathematical models for the primary and secondary side heat transport systems along with associated controls. These models are based on coupled solutions of unsteady state mass, momentum and energy conservation equations. The present paper deals with the details of the different schemes considered along with the results obtained with discussion

[en] Full text: The pressurized heavy water reactor (PHWR) consists of horizontal calandria vessel containing a large number of pressure tubes (fuel channels) connected to the reactor inlet and outlet headers by individual feeders. Till date the PHWRs in India have single phase flow. For these reactors one of the design objective is to achieve uniform outlet temperatures. This is achieved by the design process known as feeder sizing. This basically consists of accounting for the individual channel power and centre line geometry of individual feeder and iteratively adjusting the feeder hydraulic resistances within the design constraints such as limiting flow velocities. The above process has been followed for 220 MWe and 540 MWe units to achieve the desired design intent. Recently, design work on further uprating the 540 MWe units to 680 MWe units by use of concept of limited boiling (about 3% quality) at the channel exit is taken up. The sizing for such a unit has required a somewhat different approach with the design objective of achieving uniform steam quality. The approach requires minimizing the hydraulic resistance of the outlet feeders to minimize the effect of two phase on overall circuit hydraulics. This paper discusses the feeder sizing work which has been taken up for 680 MWe units with boiling in the channels

[en] The 500 MWe Indian pressurised heavy water (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like reactor addition of a pressuriser (surge tank) in the PHT system along with feed/bleed system and their safety related implications, simulation model development and transient analysis studies are necessary. The paper deals with the details of the mathematical model for pressuriser and parametric study on steam bleed valve (s) stuck open transient analysis. The studies were carried out after including the proposed new SGPC program similar to 220 MWe PHWR, which gives a 48 kg/cm² maximum SG pressure setpoint at zero power without changing the 100 % power SG set pressure

[en] The 500 MWe Indian pressurised heavy water (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like reactor stepback (which reduces the reactor power rapidly) and their safety related implications, simulation model development and transient analysis studies are necessary. The paper deals with the details of the mathematical model for SGs and parametric study on turbine trip transient analysis. The studies were carried out after including the proposed new SGPC program similar to 220 MWe PHWR, which gives a 48 kg/ cm² maximum SG pressure setpoint at zero power without changing the 100% power set pressure

[en] The 500 MWe Indian pressurised heavy water (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like primary heat transport (PHT) system configuration with two loops, four primary circulating pumps (PCPs) and four passes through core, addition of a pressuriser (surge tank) in the PHT system along with feed/bleed system and their safety related implications, simulation model development and transient analysis studies are necessary. The paper deals with the details of the mathematical model for PHT system and parametric study on one PCP trip transient analysis with set/step back. The studies were carried out after including the proposed new SGPC program similar to 220 MWe PHWR, which gives a 48 kg/cm² maximum SG pressure setpoint at zero power, without changing the 100% power set pressure

[en] The 540 MWe Indian Pressurised Heavy Water Reactor (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like Primary Heat Transport (PHT) system configuration with two loops, four Primary Circulating Pumps (PCPs) and four passes through core, addition of a Pressurizer (surge Tank) in the PHT system along with Feed/Bleed system and their safety related implications, simulation model have been developed. A reactor step-back is proposed following one PCP trip. The corresponding PCP in the healthy loop is tripped to avoid asymmetrical flow and pressure distribution in the two identical loops. In spite of such elaborate provisions, the margins from high/low PHT pressure are small following tripping of one PCP. Mathematical models for all the major components and sub-systems of the proposed 540 MWe PHWR were developed based on the conservation equations of mass, momentum, energy and equation of state. All the associated control systems are also modeled. The PHT system includes the reactor core with nuclear fuel, PCP, PHT system pressure controller with feed/bleed system and Pressurizer (Surge Tank). The secondary system includes mainly the Steam Generators (SGs), the SG level and pressure controllers, apart from the various steam cycle components. All these models are integrated together to form the Plant Transient Analysis Computer Code Dyna540. The scenario following one PCP trips leads to different states (high/low pressure in Reactor Outlet Header (ROH)) depending upon the banks in which the PCP trips. The pressurizer is connected to two ROHs on one side of the reactor. The system pressure is controlled based on average of four ROHs pressure. In the case of asymmetrical pump operation, this logic leads to a situation where individual ROH pressure goes very near the low/high PHT system pressure trip set point, even though the controlled average pressure is very close to the set pressure. The PHT high/low trips are based on individual ROH pressures. New logic was proposed for controlling the PHT system pressure that controls the pressure of the ROH with the least margin to the high or low-pressure trip set points. Investigations were performed to demonstrate this approach is helpful in over coming header pressure imbalance and will improve margins to trip set points. To overcome such situation, parametric studies were carried out to evaluate new logic to select pressure of the appropriate ROH for controlling purpose which can avoid high or low pressure trip during these transients. These logics help in improvement of margin from high/low PHT pressure trip set points. (authors)