Effect of temperature changes on antioxidant enzymes, and oxidative stress in Nerita oryzarum collected along Atomic Power Plant sites

Effect of heated effluent released from the Tarapur Atomic Power Station, India were investigated by analyzing the level of antioxidant enzyme activities on Nerita oryzarum. Seasonal variation of antioxidant enzyme (LPO, CAT, SOD, GPx and GST) were determined at six location. CAT increases in the pre-monsoon season (range of 0.12 to 2.1 mM of H2O2 consume/min/mg of protein) and LPO activities also increasing trend (range of 0.4 to 2.3 nM of MDA/min/mg) during same pre-monsoon at NIII and SII respectively. Enzyme values encountered high during pre-monsoon, which indicating that increased in temperature is resulting in increased activities. In experimental condition also (300C, 350C, 400C and 250C as control), antioxidant enzyme activities were in increasing trend due to raises of water temperatures. Present study was prima facie work related to physiological response of N. oryzarum related to heated effluent released from atomic power station and useful as baseline information for future research work.

Application of Decision Support System during the Emergency Exercises for Nuclear Emergency Management

In India, two independent decision support system (DSS) systems are developed and implemented at NAPS (Narora Atomic Power Station) and MAPS (Madras Atomic Power Station) sites for emergency response. The simulation results of the DSS systems guides in identification of emergency zones and distances and implementation of the protective actions. This paper describes application / testing of the three decision support systems developed by IGCAR, NPCIL and SRI (AERB) for dose projections and decision on the implementation of the protective action in the public domain during the recently conducted emergency exercise at NPP site. Thus three DSS systems developed in India were applied to perform atmospheric dispersion and dose assessment in public domain around the nuclear power plants (NPPs) during the recent emergency exercise. The hypothetical accident scenario was considered for the exercise. The source term estimated based on plant parameters and pre calculated source term for large number of accident scenarios was used during the exercise. The results of the assessments by using these Decision support systems were presented to the decision makers for recommendation and implementation of protective actions such as evacuation, sheltering, KIO3 distribution, contamination control etc. The Numerical Weather Forecast model was used for the area around the NPP to produce the meteorological parameters that were further used by these DSS system. Predictive assessments of the radiological situation in the vicinity of the NPP site was performed during emergency exercise with various source terms. Taking into account the high uncertainties in the source term estimation and inputs to the simulation models, the simulated results from these three DSS show a reasonable agreement. The study demonstrated the utility of DSS systems for the assessment of the radiological consequences of hypothetical nuclear accidents during the emergency exercises at different NPP sites. The experience gained in using the DSS systems for operational application to the Indian NPPs will be further used by the exercise planners and developers to improve the system continuously and their adaptation to all NPP sites in India.

Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.