Besides the gradual decrease in power and the increase of moderator temperature, the reactor becomes critical again and again. Subcriticality of the reactor can be achieved safely only by the later boron injection due to operator intervention that is only possible after certain pressure decrease. Such kind of ATWS analyses require 1- or 3-dimensional core calculations taking into account the thermo-hydraulic feedback effects.
The second code module is the neutron kinetic core model that capable to calculate the one-dimensional or three-dimensional power distribution.
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The third module is the hot channel model. Conservative boundary conditions are applied separately to the neutronics and to the hot channel calculations. Conservative neutronics values reactivity coefficients, axial power distributions in the initial states, reactivity worth of the control rods assumed for reference states like full power or hot zero power states. In fact the axial peaking can not be set conservatively by peaking factors for all possible conservative rod positions.
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For the hot channel calculations a conservatively set low inlet mass flow, higher gas gap conductance are applied in order to increase the calculated DNBR. Example 3: Recriticality evaluation of steam line break accidents The assessment of the steam line break accidents has different purposes and conservative assumptions as regards: integrity of the reactor core, evaluation of pressurized thermal shock phenomena, and calculations of the stresses of the containment.
An important aspect of the analysis was the description of uneven cooldown in different loops and consequently in different sectors of the reactor vessel and the reactor core. A three-dimensional description is essential to properly simulate the core behavior. The first used code module is the thermal-hydraulic system code that calculates the primary pressure, the core inlet temperature and flow fields.
The model has to be capable to properly model the thermal mixing in the downcomer. The second code module is the neutron kinetic core model that capable to calculate three-dimensional power distribution in case of asymmetrical cooling and rod positions. Conservative assumptions were made concerning the isothermal recriticality temperature of the core C and the coincidence of a stuck control rod with the sector of cold water from the damaged steam generator.
Certain system failures can intensify the degree of the cooling asymmetry. In the worst cases the main steam isolation valve fails to close and as a consequence the reactor coolant pump of the affected loop remains in operation thus driving the cold water directly into the reactor core. For boration only the HPSI pump is assumed to function and the boron is initially supplied to the opposite side of the core from the stuck rod and the power peak.
These system failures usually have opposite influence on DNBR.
In spite of the conservative assumptions the return to power does not damage the core. Not even DNB is calculated to occur. In general, a modest recriticality from cooldown of the core inlet temperature can easily be compensated by a well-behaved return to power. This is particularly true in cases of low coolant flow rate, when the coolant temperature rise in the core acts as the main feedback mechanism.
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These analyses demonstrate that even at full coolant flow rate the feedback from fuel temperature can keep the power within harmless limits. Lele, S. Gupta, H. Kushawaha and V. It is also useful in evaluation of plant mitigation and plant control capabilities. It also plays important role in providing limiting conditions of operation, deciding safety settings and providing inputs to designers from safety considerations. Safety analysis of Advanced Heavy Water Reactor is complex task.
Complexities arise due to large number of events to be analysed covering wide range, different phenomena to be simulated for such PIEs and complex reactor configuration based on advanced concept. 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. Moreover AHWR also has its own specific features and systems pertaining to its natural circulation characteristics.
In AHWR, as soon as the power generation starts the pump function of maintaining flow into the core starts based on natural circulation principles. This necessitates appropriate power computation with proper reactivity feedbacks. Different thermal hydraulic conditions exist in different parts of the core during both steady state and transient conditions. The power, pressures and inventories can change rapidly during these transient. This requires simulation of single and two phase natural circulation at different power levels, inventories and pressures and also under depleted inventory conditions.
Coupled neutronics and thermal hydraulics behaviour should be simulated appropriately considering various reactivity feedbacks.
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Different channels exhibit dissimilar physics behaviour that must be accounted for. Accumulators are of advanced type in which flow through paths of different resistances need to be simulated with appropriate switching action. To determine cut off level in accumulator vapour pull through phenomena should be taken in to account. In many transients coupled controller and thermal hydraulics behaviour plays a significant role. This paper discusses various aspects of safety analyses like methodology of analysis, different computer codes and their applicability for the analysis, particularly with reference to simulation of coupled phenomena.
In this paper, emphasis is given on slave channel analysis methodology for channels with different physics characteristics. 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. The Primary Heat Transport System of the proposed AHWR consists of the reactor core, core inlet and core outlet extensions, inlet feeders, tailpipes, steam drums, downcomers and inlet header Fig.
The reactor core consists of Pressure Tubes. In each Pressure Tube, a element single fuel bundle is housed. A displacer rod is placed at the centre of the bundle and the fuel elements are arranged in concentric circles. In the first circle, second circle and third circle 12, 18 and 24 fuel elements are arranged respectively. The steam drum has the large interface area, which provides natural steam separation, without using mechanical separators. At the central location between two plates of the steam drum bottom, feed water enters which leads to a subcooled condition after mixing with the saturated water.
This subcooled water leaves the steam drums through 16 downcomers to a single ring Inlet Header. From the Inlet Header, the coolant enters the Pressure Tubes through inlet feeders and core bottom extensions. The reactor has reactor shutdown system for tripping the reactor, relief system for relieving system pressure, isolation condenser for decay heat removal and three stage ECCS.
Coupling between Global analysis having point kinetics neutronics module and channels of dissimilar characteristics is also discussed and analysis is carried out for different channels.
Different models as described below can be used in the multistage analysis. Analysis can be carried out in number of stages and accordingly appropriate computer codes need to be used. In this section different codes used for such multistage analysis are described. These equations are formulated in differential stream tube form with time and one space dimension as independent variables and in terms of time and volume-average dependent variables.
Finite difference formulation is used to advance the heat conduction solutions.
The heat transfer correlation package can be used for heat structure surface connected to hydrodynamic volumes and contains correlations for convective, nucleate boiling, transition boiling and film boiling heat transfer from the wall to water and reverse transfer from water to wall, including condensation. Each fuel pin of the fifty four rod cluster is simulated using a 3-D conduction model with temperature dependent thermal conductivity.
The pressure tube and calandria tube are also separately simulated by the 3-D model. Convective and radiative heat transfer serve as the boundary conditions. For determining radiative heat exchange between different nodes a radiosity approach is followed. The deformation of pressure tube and fuel bundles is simulated thermally. The first part computes releases from different locations in the fuel channels into the PHT system, while the second part involves transportation into PHT system and containment. The one dimensional conservation equation of mass, momentum and energy are written for each subchannel and solved numerically while taking into account the possible interactions between adjacent ones.
These flow interactions involve diversion cross flow caused by pressure differences between the subchannels and turbulent interchange brought about by pressure and flow fluctuation. The basic equations and method of solutions can be obtained from  3. The core channels have been modeled by eight equivalent paths, representing 4 average , , and and 4 hot channels , , , Volume 4 to 8 of each components represent core channels.
Volume 1 and 2 of theses parallel paths represent vertical and horizontal portion of inlet feeders 4 NB sch. Volume 9 of these components simulates core top extension, while volumes, 10, 11 and 12 represent tail pipes.
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Each steam drum is connected to channels. Steam drum 3. Saturated volume of steam drums components , , and are modeled separately using 10 control volumes in each drum. The bottom part of the drum subcooled portion in between two baffles is represented by subcooled volume components , , and The feed water along with recirulation flow enters into the subcooled part of the drums.
Flow from this portion of steam drums enters the downcomers. Feed water conditions are simulated using component Suction header of feed pumps is modeled by component Two feed pumps are modeled by components and Component models 10 NB discharge header.