[en] Full text: The pressurized heavy water reactor (PHWR) consists of horizontal calandria vessel containing a large number of pressure tubes (fuel channels) connected to the reactor inlet and outlet headers by individual feeders. Till date the PHWRs in India have single phase flow. For these reactors one of the design objective is to achieve uniform outlet temperatures. This is achieved by the design process known as feeder sizing. This basically consists of accounting for the individual channel power and centre line geometry of individual feeder and iteratively adjusting the feeder hydraulic resistances within the design constraints such as limiting flow velocities. The above process has been followed for 220 MWe and 540 MWe units to achieve the desired design intent. Recently, design work on further uprating the 540 MWe units to 680 MWe units by use of concept of limited boiling (about 3% quality) at the channel exit is taken up. The sizing for such a unit has required a somewhat different approach with the design objective of achieving uniform steam quality. The approach requires minimizing the hydraulic resistance of the outlet feeders to minimize the effect of two phase on overall circuit hydraulics. This paper discusses the feeder sizing work which has been taken up for 680 MWe units with boiling in the channels

[en] The pressurized Heavy Water Reactor (PHWR) consists of horizontal calandria vessel containing a large number of pressure tubes (fuel channels) connected to the reactor inlet and outlet headers by individual feeders. Coolant flow distribution among the pressure tubes play a vital role in extraction of thermal power. For these reactors one of the design objectives is to achieve uniform coolant outlet temperatures by providing coolant flows according to the channel power. This is achieved by the design process known as feeder sizing. This basically consists of accounting for the individual channel power and centre line geometry of individual feeder and iteratively adjusting the feeder hydraulic resistances within the design constraints such as limiting flow velocities, channel flows. Recently, the first unit of 540 MWe i.e Tarapur Atomic Power Project (unit 4) has been commissioned. This paper discusses the verification of the coolant flow distribution in the core with the design estimated flow during different commissioning phases. (author)

[en] In a 540 MWe PHWR reactor at TAPP-4 the pressuriser has been incorporated in the PHT pressure control system to provide the necessary vapour cushion for PHT main circuit to reduce pressure variations due to transients involving swell and shrinkage. Need for the Pressuriser is due to the large inventory in PHT main circuit and relatively large heat source. The incorporation of Pressuriser is one of the modifications while upgrading from prevalent operating 220 MWe reactors to the present 540 MWe at TAPP-4. The sizing, design and selection of the 540 MWe Pressuriser has been done w.r.t swell/shrinkage requirement during normal/transient operation of the reactor. In this paper the commissioning and operating experience of the Pressuriser in conjunction with the PHT pressure control system is presented. During light water commissioning of PHT circuit some major hurdles like failure of pressuriser heaters due to faulty instrumentation were faced. Subsequently the following corrective action was carried out. The cause of instrumentation/logic fault was reviewed and rectified, heaters were reinstalled, level instrumentation behaviour was repeatedly checked with light water, the pressuriser normal operating level and heater cut-off level was also raised, the performance of the Pressuriser with the revised levels was checked by analysis. After heavy water filling various tests carried out on the pressure control system which confirmed the effectiveness of the pressuriser. The ongoing smooth operation of TAPP-4 at 90 % FP also justifies the requirement of Pressuriser in 540 MWe PHWR. (author)

[en] Although the immediate need to purify the Primary Heat Transport coolant during a normal shutdown may not be imminent yet a prolonged shutdown may call for purification due to cumulative build-up of ionic impurities and crud levels in the coolant. Shutdown purification in operating 220 MWe reactors is effected through a separate shutdown Ion Exchanger placed in the circuit across the S/D cooling system. Purification flow through this circuit was quite less and was also incurring high man rem expenditure. Relatively higher S/D purification flow and doing away with the S/D Ion Exchanger formed the basis of the newly incorporated ECN involving addition of just four globe valves and forcing the flow through the dedicated main purification path . This paper brings out the minor changes carried out in 540 MWe reactors, and the ongoing 220 MWe projects, that brought about relatively higher purification flows and reduced man rem expenditure. (author)

[en] In 700 MWe PHWR, during reactor operating at full power, more than two channels are required to be refueled per day. To achieve radial flux flattening for maximum power output from core, a differential burn-up scheme is adopted. This is done by judicious adjustment of the relative refueling rates in different core regions. When the refuelling of channel is undertaken i.e. during 'ON POWER REFUELLING' both fuelling machine (F/Ms) are clamped on both end fitting (E/F) of same channel. During on-power refuelling of PHWRs, cold water enters the channel from Fuelling Machine, which may affect the void fraction and temperature at the outlet of the channel being refuelled. This kind of temperature perturbation which might affect the neighboring channels due to their communication with the refuelling channel through the common ROH. To estimate the temperature drop and void fraction at the exit of channel which has been refuelled, a computer code 'RECHAN' has been developed. (author)

[en] Daily online refueling and positive void reactivity gain are two main characteristics of PHWR. 700 MWe PHWR is under construction. It consists of 392 fuel channels and in each fuel channel there are 12 fuel bundles. It is designed in such a way that there will be voiding (boiling) of coolant from 10"t"h fuel bundle onwards. In the refueling of a fuel channel, 8 burnt fuel bundles will be replaced with fresh NU fuel bundles. Daily refueling of fuel channels will introduce about 0.3 mK positive reactivity and perturbation in the operating neutron flux shape. Being a loosely coupled reactor core, 700 MWe PHWR will be equipped with reactor regulation system to control reactor bulk power as well as zonal powers. The effect of refueling of a fuel channel is studied in 3 different scenarios by coupling the neutronics and thermal hydraulics properties of reactor core in a code and results are presented

[en] Indian Pressure Heavy Water Reactors generally consists of 300-400 horizontal fuel channels depending on the rating of the plant. It consists of a hot Pressure Tube (PT) and a concentric Calandria tube (CT) with a 9 mm annulus containing CO₂ gas as annulus gas monitoring system. Inside each pressure tube there are 10-12, 0.5 m long fuel bundles, each consisting of 19 or 37 fuel pins with fuel elements containing natural UO₂ pellets and Zircraloy-4 clading. During a postulated loss of coolant accident (LOCA), the coolant supply is lost and designed Emergency Core cooling system (ECCS) will actuate and removes decay heat from fuel. Aim of the present research is to find means of enhancing the heat transfer efficiency between the CT and moderator, such that in the case of any accident, film boiling period is reduced and channel integrity is ensured. One of the ways to enhance heat transfer from CT to moderator is by increasing the critical heat flux (CHF) on the outer surface of the calandria tube. Increasing the CHF for a given subcooling on the outer surface of the calandria tube can be achieved by surface modifications. The ability of the surface modification to increase the CHF on the outside surface of the calandria tube is demonstrated in a series of small scale experiments on a 15.2 mm OD and 0.4 mm thick zircaloy tube. Analysis of the test results indicated that increasing the tube's CHF using a glass-peening or grounded or grit blasted process is a promising option for reducing moderator subcooling requirements

[en] The pressurized Heavy Water Reactor (PHWR) consists of horizontal calandria vessel containing a large number of pressure tubes (fuel channels) connected to the reactor inlet and outlet headers by individual feeders. 700 MWe reactor consists of two figure of eight loops. Each loop is made of one Steam Generator (SG), one Primary Circulating Pump (PCP), Reactor Inlet Header (RIH) and Reactor Outlet Header (ROH) on either side of core. The coolant is circulated from the RIH to the pressure tube through the inlet feeders and comes out and joins ROH through outlet feeders on the other side of the core. In 540 MWe PHWR the core is split into two vertical halves such that each vertical half forms one loop. This design concept has been modified in 700 MWe by implementing interleaving of feeders. In this concept the channels of both loops are uniformly distributed throughout the core. The PHWR have positive void coefficient of reactivity. The main reason for the choice of interleaving design concept was to lower reactivity rise arising from the rupture of header during an anticipated Loss Of Coolant Accident (LOCA). This concept also benefits in reducing the magnitude of the peak power, minimizing flux tilt and improving safety margin. The voiding of the coolant in 540 MWe reactor would affect one vertical half of the core whereas now in 700 MWe reactor the void would be uniformly distributed throughout the core. This concept of Interleaving of feeders has been achieved by increasing the length of the headers so as to cater to the supply of coolant to the channels throughout the core and modification of the header location and feeder layout. This paper discusses the implementation of feeder interleaving and the thermal hydraulic aspect in 700 MWe reactor. (author)

[en] Highlights: • Effect of spray mass flow rate, SMD and nozzle geometry is on the vessel depressurization is studies in a scaled down facility. • Decrease in pressure and iodine concentration with time follow exponential trend, the depressurization rate decreases with the increase in the droplet size. • Depressurization rate Increases with the increase in spray mass flow rate. • An empirical correlation for decay rate constant is proposed. • The experiments are modeled using ASTEC LP code. - Abstract: The containment of a nuclear reactor acts as the ultimate barrier for radionuclides to be leaked in the environment. Severe accidents in nuclear reactors may result in the pressurization of the containment and may provide different potential leak paths. So, to enhance the safety of new designs of Indian pressurized heavy water reactors, an additional safety measure called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents. As a contribution to the safety analysis of condition post severe accident, experiments are carried out to investigate the system performance. The accidental conditions are simulated by injecting the saturated steam into the test vessel filled with air at atmospheric pressure and temperature. The effect of different spray parameters such as nozzle geometry, spray mass flow rate and Sauter mean diameter on vessel pressure and temperature is measured. Three initial vessel pressure 1.5 bar, 2.0 bar, 2.5 bar and five different nozzles with nozzle orifice diameter as 1.65 mm, 2.10 mm, 2.45 mm, 2.75 mm and 3.10 mm are chosen for the experimental investigation. The vessel pressures are chosen based on the post accidental conditions in Indian pressurized heavy water reactor. Experiments are conducted with water at room temperature as the spray medium. The experiments are carried out in a vessel of 500 mm diameter and 1200 mm height. These studies are carried out to optimize the containment spray system configuration for best effectiveness. It is seen that the Sauter mean diameter and nozzle geometry influences the depressurization rate of the vessel. The depressurization rate is inversely proportional to the Sauter mean diameter while it increases with the increase in the spray mass flow rate/ Reynolds Number.

[en] Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization. The pressure and temperature history of the system is measured using high temperature pressure transmitter and the K-type thermocouples. The iodine scrubbing is measured through periodic sampling from the vessel. The influence of the Sauter mean diameter (SMD) is studied on the depressurization rate of the vessel at different vessel pressures. Studies are performed to optimize the containment spray system configuration and to establish the phenomena with respect to Indian pressurized heavy water reactors. In all the experiments, the spray flow rate is kept constant, while the SMD is varied by using different spray nozzles.