Avinash J. Gaikwad

Transient analysis following tripping of a primary circulating pump for 500-MWe PHWR power plant

The 500-MWe Indian pressurized 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 pressurizer (surge tank) in the PHT system along with a feed/bleed system and their safety-related implications, simulation model development and transient analyses are necessary. To mitigate the swell and shrinkage in the PHT system and to avoid high/low PHT system pressure during transients, a 30 m/sup 3/ pressurizer along with feed/bleed system is also included in the PHT pressure control system for the 500-MWe PHWR. At the same time, a reactor stepback is also actuated during the one PCP trip transient after tripping the corresponding (same side) PCP in the other loop. The corresponding PCP in the healthy loop is tripped to avoid unsymmetrical flow and pressure distribution in the two identical loops. A detailed transient analysis is required to study the individual contributions of various systems and provisions such as pressuriser, feed/bleed, and reactor step/set back in mitigating the consequences of the malfunction.

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

For AHWR (Advanced Heavy Water Reactor), a pressure tube type Boiling Water Reactor (BWR) with parallel inter-connected loops, the Steam Drum (SD) level control is closely related to Main Heat Transport (MHT) coolant inventory and sustained heat removal through natural circulation, hence overall safety of the power plant. The MHT configuration with multiple (four) interconnected loops influences the SD level control in a manner which has not been previously addressed. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Event (PIEs) for Loss of Coolant Accident (LOCA), which postulates a double ended break in the four partitioned Emergency Core Cooling System (ECCS) header. A conventional individual three-element SD level controller can not account for the highly coupled and interacting behaviors, of the four SD levels. An innovative three-element SD level control scheme is proposed to overcome this situation. The response obtained for a variety of unsymmetrical disturbances shows that the SD levels do not diverge and quickly settle to the various new set points assigned. The proposed scheme also leads to enhanced safety margins for most of the PIEs considered with a little influence on the 100% full power steady-state design conditions.

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding systems coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Logic modification to avoid ECCS lineup in the primary heat transport system during thermal shrinkage transients of 220 MWe PHWR nuclear power plant

Operational transients involving large thermal shrinkage in the primary heat transport (PHT) volume of a pressurized heavy water reactor nuclear power plant may result in the undesirable lowering of PHT system pressure to a very low value. During some of the reactor trip incidents as reported from the operating units, the minimum PHT pressure obtained have lead to unwanted lineup of emergency core cooling system. A reduction of the dip in the PHT system pressure during such incidence can be achieved by increasing the net feed to the PHT volume, i.e., by increasing the holdup inventory in a fixed PHT boundary. Another option to maintain the PHT system pressure high during such transients is to keep the PHT temperature at a higher value. This can be achieved by removing less heat from the primary coolant. To reduce the heat transfer from the primary coolant to the secondary coolant in the steam generator (SG), the steam generator pressure controller (SGPC) set point has to be increased during such incidences. This will lead to a higher steam pressure and temperature in the SGs, thus reducing the temperature difference and heat transfer between primary and secondary coolant. Analysis has been carried out using the integrated system process dynamics analysis code. This code incorporates the mathematical models for the primary and secondary heat transport systems, 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 mitigation schemes considered along with the results arrived at.

Selection of Steam Drum Level Control Method for Multiple Drum Interacting Loops Pressure Tube-Type BWR

Three element Steam Drum (SD) Level Controller has been conventionally used for most of the boilers, Nuclear power plant steam generator & Boiling Water Reactor (BWRs). Based on the process dynamic studies it was found that this scheme does not work properly for an interacting, interconnected multiple loop boiling water system i.e., Advanced Heavy Water Reactor (AHWR). It is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR). It has 4-inter-connected parallel loops with 113 × 4 = 452 boiling channels in the Main Heat Transport (MHT) system. These multiple (four) interconnected loops influences the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. Also it is not possible to maintain a steady level in all the SDs even without any external disturbance/perturbation with 4 conventional 3-element individual SD level controllers. To overcome these interactions it is proposed to interconnect all the four steam drums in the liquid & vapor regions respectively. This makes the 4 SDs behave like a single entity. The influence of the interconnect configuration & the level controller are studied in detail to find a robust solution. The response obtained for unsymmetrical core power, symmetrical power maneuvering and reactor trip transients shows that the SD levels do not diverge and quickly settle very near to the set points assigned with SD interconnect schemes.

Measurement of Steady-State CHF in Horizontal Channels for Low-Pressure, Low-Flow Conditions

Experimental investigations on critical heat flux (CHF) are mostly on vertical channels involving high mass fluxes and high system pressures. Reported studies on CHF in horizontal flow channels under low-pressure, low-flow (LPLF) conditions are limited. Understanding CHF is essential in the design and operation of heat exchangers and heat-generating devices including fuel channels of nuclear reactors. The present work investigates CHF in horizontal tubes for low steady flow at atmospheric pressure conditions. Appearance of a “red hot” spot on the test section is considered to be the occurrence of critical heat flux condition in this study. Present data could not be predicted using the reported method of applying a correction factor for the vertical lookup table data. A correlation using the experimental data is developed incorporating the fluid-to-fluid modeling parameters for the prediction of CHF in horizontal channels under LPLF conditions. Numerical study using thermal hydraulic system code RELAP5 suggests liquid film dryout as the mechanism of CHF occurrence in the present investigations.

Reliability studies of decay heat removal system

Passive systems have been implemented in most of the advanced reactors in order to improve the availability of the plant. In this paper the methodology used for performing the passive system reliability analysis has been discussed. A case study on passive decay heat removal system of large sized PHWRs has also been discussed. Based on the thermal hydraulic analysis failure points have been generated and the failure surface has been developed. Finally by using the Monte Carlo simulation technique failure probability has been estimated.

Statistical Aggregation of Extreme Value Analysis Models

Extreme value analysis is important for designing the engineering structures robust enough to withstand external hazards such as wind load, flood level, earthquake etc. Important use of routinely measured station data is made to obtain year wise maximum data of the required variable specifically related to an external hazard. As per common practice, the year wise extreme data are fitted with generalised extreme value distribution function to make predictions for various return periods. Model uncertainty with respect to the variation of model parameters is also estimated. Multiple models are developed for data from the measuring stations. A methodology for statistical aggregation of multiple models is developed and demonstrated considering data from four measuring stations. In this statistical aggregation method, the statistical property of the GEV model has been preserved.

Uncertainty in Fission Product Transient Release Under Accident Condition

During power irradiation of nuclear fuel pin in a nuclear power plant, fission product radionuclides get accumulated in the fuel matrix. Following an initiating event causing design extension condition without core melt accident, temperature of the fuel increases even after reactor shutdown due to the decay heat. The fractional release of radionuclides from fuel matrix to the pellet-clad gap based on thermal diffusion phenomena is exponential of an exponential function of fuel temperature. The fractional release is found to be very small if fuel temperature is below the 1500 K. Higher fuel temperature persistent duration increases fractional release. An analytical uncertainty estimation methodology for this non-linear function has been developed and applied to demonstrate its application to the radionuclide release phenomena. The safety importance of the fuel temperature and duration from the radionuclide release point of view and associated radiological impact has been assessed.

Comparison of Lumped Parameter and CFD Code Predictions: Sump Evaporation Phenomena

In nuclear reactor containment, Sump, Calandria Vault and Calandria Vessel contain large amounts of water. Condensation on walls and containment spray system actuation also results in accumulation of water in the containment sump during the accident conditions. Evaporation of water takes place during the accident conditions and needs to be accounted in the hydrogen distribution analysis. Lump parameter codes such as ASTEC have built-in models for sump water evaporation. However, CFD codes are being increasingly used for containment hydrogen distribution studies, development of a sump water evaporation model for multi-dimensional calculations is required. The sump model is implemented through mass and energy balance using two different approaches. The main focus of the paper is on the simulation of sump evaporation experiment conducted in TOSQAN facility using the lumped parameter code and its comparison with the CFD results and the available experimental data.