H. G. Lele

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.

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.

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.

Modelling of thermal and flow stratification for reactor pressure vessel pressurised thermal shock

A Pressurised Thermal Shock (PTS) occurs when the hot vessel wall of a pressurised water reactor is exposed to low-temperature, high-pressure fluid during a transient such as Emergency Coolant Injection (ECI) following a Loss of Coolant Accident (LOCA). Overcooling events have the potential to threat safety because the resulting high thermal stress, along with irradiation induced loss of ductility and stress-concentrating faults, or macro-cracks, can result in a through-the-wall crack propagation and reactor pressure vessel failure. The problem of PTS therefore has been studied extensively. This paper presents the development and application of an analytical thermal hydraulic model for predicting the growth of PTS in the downcomer in case of ECI. A preliminary assessment of pressure vessel integrity is also presented and discussed.

Tarapur atomic power station: analysis of station blackout scenario

Abstract India is currently operating two BWR built by General Electric Company. The design features of these reactors are similar to the Fukushimas BWR except some better containment features in Indian BWR. This paper discusses the enveloping scenario of station blackout of infinite duration with no operator action and no component failure. The paper describes the details of modelling the TAPS-BWR plant model including SCDAP modelling of reactor core in system code RELAP5 and further thermal hydraulic safety assessment of station blackout scenario. The analysis brought out effectively the response of the plant to this high-pressure severe accident scenario. The time line of the severe accident progression will give details of various stages of accident progression along with hydrogen generation, which will be useful in evolving suitable severe accident management guidelines.

Steam drum level control studies of a natural circulation multi loop reactor

Abstract The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory.

Best estimate plus uncertainty analysis of LBLOCA for Indian PHWR

Abstract Deterministic safety analysis is an important tool for confirming the adequacy of provisions within the defense-in-depth concept for the safety of nuclear power plants. One of the important design basis events is considered to be a complete double-ended guillotine rupture i.e. Loss of Coolant Accident (LOCA) of largest and coldest pipe in the primary coolant circuit. The present work deals with this scenario for an Indian PHWR. It highlights the identification of critical break size leading to the maximum clad temperature using the best estimate code RELAP5. Further, important parameters affecting the clad temperature are described along with results of initial sensitivity studies to select dominant uncertain parameters. For the uncertainty propagation, Latin Hypercube Sampling (LHS) is used instead of simple random sampling for Monte-Carlo simulation. The inherent characteristic of LHS is to reduce the required runs for Monte-Carlo simulation to manageable order for the current computing capability. The 95th percentile value of peak cladding temperature (PCT) is obtained by the method and compared with acceptance criteria.

Simulation of load following mode of operation for a natural circulation pressure tube type BWR

Abstract A new nuclear power reactor under design study is a vertical pressure tube type boiling light water cooled and heavy water moderated. One of the passive design features of this reactor is the heat removal through natural circulation of primary coolant at all power level with no primary coolant pumps. Nuclear plants are mainly base load units, but the proposed plant with various advance features has to operate in load following mode i.e. Reactor follows Turbine (in a limited range). In this mode, any alteration in turbine load results in the steam pressure change. The steam pressure error is fed to the Reactor Regulating System (RRS), which changes the reactor power to control the system pressure. To study this mode of plant operation, a plant simulation model with the feedbacks from various controllers has been developed using the RELAP5 code. This integrated plant model has been used for simulating the load-varying scenario for a change in plant load. All the process dynamics, modeling, design verification and performance issues are discussed in this paper.

An Experimental Investigation on the Behaviour of Pressure Tube Under Symmetrical and Asymmetrical Heating Conditions in an Indian PHWR

Thermal behavior of fuel channel under loss of coolant accident (LOCA) is a major concern for nuclear reactor safety. LOCA along with failure of emergency cooling water system (ECC) may leads to mechanical deformations like sagging, ballooning or even release of containment in open atmosphere due to breaching of pressure tube (PT) under certain depressurization and voiding rates. In order to understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of Indian Pressurized Heavy Water Reactor (IPHWR) under symmetrical and asymmetrical heat-up conditions. For simulating the fully voided scenario, symmetrical heating of pressure was carried out by injecting 13.2 KW (2% of nominal power) to all the 19 pins and the temperatures of pressure tube, calandria tube and clad tubes were measured. During symmetrical heating the sagging of fuel channel was initiated at 460 °C and the highest temperature attained by PT was 650 °C. The decay heat from clad tubes was dissipated to moderator mainly by radiation and natural convection. The highest temperature of 680 °C was observed over the outer ring of clad tubes of fuel simulator. Again, to simulate partially voided condition, asymmetrical heating of pressure was carried out by supplying 8.0 kW power to upper 8 pins of fuel simulator and temperature profiles were measured. Along the circumference of pressure tube (PT) the highest temperature difference of 320 °C was observed, which highlights the magnitude of thermal stresses and their role in breaching of pressure tube under partially voided conditions. However, the integrity of pressure tube was intact during both symmetrical and asymmetrical heat-up conditions.Copyright

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.