D. Mukhopadhyay

Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR

The reactor channel of the horizontal core of pressurized heavy water reactors experiences very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders caused by certain breaks known as critical channel breaks. In this type of accident the reactor trip is delayed causing a gross mismatch of the heat generation and heat removal in the channel, thus leading to rapid temperature rise in the affected channel. A study has been carried out to identify the phenomena and the break size leading to such a situation. Severe fuel damage is predicted in the channel.

Evaluation of operator actions for beyond design basis events for AHWR

Enhanced defence-in-depth is incorporated in the proposed Advanced Heavy Water Reactor (AHWR) as a part of their fundamental safety approach to ensure that the levels of protection in defence-in-depth shall be more independent from each other than in existing installation. Safety is enhanced by incorporating into their designs, increased emphasis on inherently safe characteristics and passive systems as a part of their fundamental safety approach. It is ensured that the risk from radiation exposures to workers, the public and the environment during construction/commissioning, operation, and decommissioning, shall be comparable to that of other industrial facilities used for similar purposes. This implies that there will be no need for relocation or evacuation measures outside the plant site, apart from those generic emergency measures developed for any industrial facility. It has been demonstrated by analyses that there is no core damage for PIEs with frequencies more than 10-10/year. However some scenarios in residual risk domain are considered to demonstrate that dose at plant boundary is within prescribed acceptable limit. It is also possible to arrest core damage progression at various stages of event progression, by incorporating certain operating procedures, without any release. This paper discusses analyses of such low frequency event with multiple failure under the category of “Decrease in MHT inventory” where plant related symptoms like channel exit temperature, channel component temperatures, moderator level with respect to time are quantified. Further analyses are carried out for these events to demonstrate the effectiveness of action plan like flooding of the cavity of the containment.

Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

Thermomechanical Behavior of Pressure Tube Under Small Break Loss of Coolant Accident for PHWR

Some postulated events for pressurized heavy water reactor (PHWR) small break loss of coolant accident (SBLOCA) may lead to flow stratification in the reactor channels. Such stratified flow causes a circumferential temperature gradient in the fuel bundle as well as in the surrounding pressure tube (PT). The present investigation has been performed to study the thermomechanical behavior of a PT under an asymmetric heat-up condition arising from flow stratification in a 19 pin fuel element simulator. A series of experiments has been carried out at various stratification levels and PT internal pressures. The asymmetrical heat-up creates a temperature difference of 400 °C across the diameter of the PT. At high temperature the internal pressure causes ballooning of the PT. With the stratification, ballooning is found to get initiated at top hot side of PT and further propagates unevenly over its periphery. Axially ballooning is found to get initiated from center and then propagates toward both the ends of the PT. This results in an axial temperature gradient on the PT in addition of circumferential gradient. For a pressure higher than 4.0 MPa, the integrity of PT is found to be lost due to the combined effect of circumferential and axial temperature gradient generated under uneven strain distribution.

Severe accident analysis to evolve insight for severe Accident Management Guidelines for Large Pressurised Heavy Water Reactor

The Pressurised Heavy Water Reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Usually for all these designs the Emergency Operating Procedures (EOPs) are developed in support of detailed accident analysis, which gives an adequate coverage for design basis accidents. Currently the designers are making provisions [1& 2] in design to mitigate progression of accidents arising from multiple failure accidents like Large Break Loss of Coolant Accident with failure of Emergency Core Cooling System and failure of moderator as heat sink. Many designs of Large PHWRs have adopted the approach of symptom based EOPs to handle multiple failure events as currently practiced for Light Water Reactors (LWRs). Severe accident analysis is an important aspect which complements Severe Accident Management Guidelines (SAMG) development process. These analysis provide insight into the accident progression and basis to develop the SAMG. The order of uncertainty in modelling the phenomena is very high. Hence it is emphasized that different computational models be used so that an un-biased “insight” can be evolved which can be used for SAMG development. The paper discusses two categories of severe accident analyses for such large reactors for multiple failure transients involving a high pressure scenario (initiation event like SBO) and low pressure scenario (initiating event like LOCA). The insight evolved from these analysis is being discussed in the paper.

Verification of Severe Accident Management strategies for VVER 1000 (V320) reactor

Severe accident analysis of a reactor is an important aspect in evaluation of source term. This in turn helps in emergency planning and Severe Accident Management (SAM). The use of the severe accident management guideline (SAMG) is required for accident situation which is not handled adequately through the use of Emergency Operating Procedures (EOP), thus leading to a partial or a total core melt. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). Initiation of SAMG for VVER-1000 is considered at two core exit temperatures viz. 650°C as a desirable entry temperature and 980°C as a backup action [1]. Analyses have been carried out for VVER-1000 (V320) for verification of some of the strategies namely water injection in primary and secondary circuit. These strategies are analysed for a high and low pressure Primary Circuit transients. Station Blackout (SBO) is one such high pressure transient for which core heat can be removed by natural circulation of the primary circuit inventory by maintaining the secondary side inventory. This strategy has been verified where the feed water injection to secondary side of SG is considered from external power sources (e.g, mobile DG sets) as suggested in SAM guidelines. The second transient analysed for verification of the core flooding during Large Loss of Coolant Accident (LOCA) along with SBO, a low pressure event. The injection to secondary circuit is initiated with the available safety pumps and mobile DG sets as soon as feed pumps trip. The analysis shows that SG flooding is not adequate to arrest the degradation of the core. In the second strategy for LOCA transient, the injection to primary circuit has been initiated at 650°C core exit temperature. The analyses show that core flooding is not adequate to arrest the degradation of the core for the large LOCA where as for small break LOCA the injections through available safety systems are adequate. The assessments are carried out with integral severe accident computer code ASTEC V1.3.

Neural network based diagnostic system for accident management in nuclear power plants

An artificial neural network (ANN) based diagnostic system for identification of large break loss of coolant accident (LOCA) transients in nuclear power plants (NPPs) has been developed. This is an operator support system which assists the operator in identifying a transient quickly using ANNs. A large database of transient (LOCA) analyses of reactor process parameters has been generated for reactor core, containment, environmental dispersion and radiological dose to train the ANNs. These data have been generated using various codes e.g., RELAP5, CONTRAN. The trained neural network has been integrated with the diagnostic system for future transient prediction. The main diagnostic system features several important operator support features that are useful in accident management. This paper highlights the important features of diagnostic system. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs.

Ballooning of Pressure Tube Under LOCA in an Indian Pressurised Heavy Water Reactor

A study has been carried out by experimental simulation of the loss of coolant accident (LOCA) in an Indian pressurized heavy water reactor (IPHWR). The experiment has been carried out taking a completely voided fuel channel of Indian PHWR at 40 bar inside pressure as test-section. In order to simulate the rate of heat generation during LOCA, the pressure tube (PT) was electrically heated with a 12VDC/3500A rectifier. Initially the set-up was maintained at 300 °C temperature by resistance heating of PT. After attaining nearly steady state a step input of 21 kW electrical heating was given to the test-section which resulted in the temperature rise of PT with a gross rate of 2.8 °C/s. The ballooning deformation of test-section tube i.e. PT initiated at 575 °C temperature. With the progress of ballooning the rate of temperature rise was reduced due to high heat transmission to CT and subsequently to water in the tank surrounding CT. The pressure tube (PT) and calendria tube (CT) contact established at the average PT temperature of 680 °C. The contact was also confirmed from the average temperature profile of CT.Copyright

Thermal Hydraulic Analysis Due to the Changes in Heat Removal for Advanced Heavy Water Reactor

Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Changes in heat removal by primary heat transport system of a reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to Cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. Causes for events under category of increase in heat removal are mainly malfunctioning of feed water heaters, Isolation Condensers (IC) inlet valves and controllers. These events lead to cooldown of system and addition of positive reactivity addition due to void collapse. Various events considered are Feed Water System malfunctions that result in decrease in feed water temperature, inadvertent opening of IC valve, Failure of PHT Pressure Control System and Decrease in pressure controller set point to 67 bars. Causes for events under category of decrease in heat removal are mainly malfunctioning of controllers, feedwater valves and operating events like turbine trip. Functioning of passive cooling system and different valves play important role for these events. These events lead to increase in system pressure. Various events considered are Loss of normal feed water flow (multiple trains), Turbine trip without bypass without IC, Turbine trip without bypass with IC, Turbine trip with bypass without IC, Increase in PHT pressure controller set point, Decrease in level controller set point, Turbine Trip with setback, Decrease in steam flow and Class IV power failure. Changes in the system voids and pressures as a result of change in the heat removal leads to complex reactivity feedback due to coolant temperatures, void fraction and fuel temperatures. These changes in the reactor power together with void distribution change affect two-phase natural circulation flow. This paper brings out these aspects. It discusses descretisation of the system and brings out various design aspects. In this paper summary of analysis for each event is presented, various modeling complexities are brought out, evaluation of acceptance criteria is made and design implications of each event is discussed.© 2004 ASME

Analysis for the Case of Channel Flow and Reactivity Changes for Advanced Heavy Water Reactor

The proposed Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. Decrease in coolant flow or control rod malfunction can lead to undesirable rise in clad surface temperature depending upon severity and characteristics and response of the reactor and associated systems. In this paper safety assessment of the AHWR is made due to above events of different severity. Cause for events under category of decrease in coolant flow is mainly channel blockage of different severity at different locations. There is no other reason as it is natural circulation based reactor. Effect of flow decrease can be different in different channels and at different axial locations. In this paper channel blockages of different sizes are analysed at core inlet and using slave channel approach. Changes in reactivities can occur due to inadvertent withdrawal of one or more control rods from reactor core. In this analysis one control rod assembly is assumed to be removed from core. The event is simulated by addition of 5 mk reactivity in 120 seconds depending on the speed of withdrawal of assembly. The analysis for the above events are complex due to various complex and wide range of phenomena involved during different PIEs under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, coupled neutronics and thermal hydraulics behaviour, and coupled controller and thermal hydraulics. In this paper summary of analysis for each event is presented. In this paper, various modeling complexities are brought out; evaluation of acceptance criteria is made and design implications of each event are discussed.Copyright