Pressure drop is the difference in pressure between two points of interest in a fluid system. In general, pressure drop can be caused by resistance to flow, changes in elevation, density, flow area and flow direction. Pressure drops in natural circulation systems play a vital role in their steady state, transient and stability performance.

It is customary to express the total pressure drop in a flowing system as the sum of its individual components such as distributed pressure loss due to friction, local pressure losses due to sudden variations of shape, flow area, direction, etc. and pressure losses (the reversible ones) due to acceleration (induced by flow area variation or by density change in the fluid) and elevation (gravity effect). An important factor affecting the pressure loss is the geometry. In a nuclear reactor, we have to deal with several basic geometrical shapes (circular pipes, annuli, etc.) and a number of special devices like rod bundles, heat exchangers, valves, headers, plenums, pumps, large pools, etc. Other factors are concerned with the fluid status (single or two phase/one component, two-component or multi-component), the flow nature (laminar or turbulent), the flow pattern (bubbly, slug, annular, etc.), the flow direction (vertical upflow, downflow, inclined flow, horizontal flow, countercurrent flow, etc.), flow type (separated and mixed), flow paths (one-dimensional or multi-dimensional, open or closed paths, distributor or collector), and the operating conditions (steady state or transient).

An important focus of this phenomenon is the geometric conditions that hinder the establishment of fully developed flow specially when the fluid in question is a mixture of steam, air and water. This complex thermo-fluid dynamic phenomenon warrants special attention. However, it is worth mentioning here that though in many systems like the primary system of a nuclear power plant, flow is mostly not fully developed, pressure drop relationships used in these systems are invariably those obtained for developed flow. This practice is also experimentally proved to be more than adequate in most of the cases. However, in some specific cases like containment internal geometry, it is necessary to consider thermo fluid dynamics in the developing region.

A final, very important issue, is concerned with the driving force depending on whether the flow is sustained by a density difference in the fluid (natural circulation) or by a pump (forced convection), or whether there will be feedback between the pressure loss and the extracted power or not. Normally the pressure loss inside a device depends on the nature of flow through the device and not on the nature of driving head causing the flow. However, under some circumstances, because of local effects, the pressure loss may get influenced by the nature of driving force.

Some of the safety features included in the ACR-1000, which are discussed in this article, are:

— Two fast-acting, fully capable, diverse, and separate shutdown systems, which are physically and functionally independent of each other and also from the reactor regulating system;

— Emergency core cooling (ECC) system, comprised of:

o Emergency coolant injection (ECI) system to refill the HTS and cool the fuel following a loss of coolant accident (LOCA);

o Long term cooling (LTC) system, a four-quadrant system capable of operating in either ‘shutdown cooling mode’, taking suction from the reactor outlet headers and returning it to the inlet headers via heat exchangers for heat removal, or in post-LOCA ‘recovery mode’, drawing water from dedicated grade level tanks and the reactor building (RB) sumps to cool the fuel in the long term, and;

o Core make-up tanks (CMTs) (see Fig. II-2) that provide make-up to limit the extent and duration of voiding in the intact HTS following a rapid cooldown event that depletes HTS inventory.

— Containment system (steel-lined containment structures with low leakage, containment isolation system, containment heat removal system, etc.);

— Containment cooling system, comprised of local air coolers (LACs) for active air circulation and heat removal, a containment cooling spray for post-accident pressure and temperature suppression, and provisions to interconnect the major volumes of the RB to establish an air flow path for natural circulation;

— Reserve water system (RWS, Fig. II-3), which provides an emergency source of water by gravity to steam generators, moderator system, and HTS, if required;

— Emergency feedwater system (EFW), a four-quadrant system that provides make-up feedwater to the steam generators from reserve feedwater tanks when the main feedwater system is unavailable;

— Primary coolant system (heat transport system), laid out with heat transport pumps and steam generators above the core to promote natural circulation of the primary coolant for accidents when the heat transport pumps are not operating, and;

— Moderator system, a low pressure and temperature heavy water system contained in a Calandria vessel, which moderates nuclear fission and acts as a heat sink following postulated accidents. Natural circulation of the moderator in the Calandria following an accident prevents localized heat-up of the moderator, and can prevent a severe accident even if all normal means of fuel cooling and decay heat removal are not available.

Natural circulation features in ACR are grouped into three generalized categories; cooling of the fuel in the primary heat transport system, cooling of the moderator in the Calandria, and circulation and cooling of the atmosphere inside containment.

Some designs utilize the reactor cavity and other lower containment compartments as a reservoir of coolant for core cooling in the event of a break in the primary system. As such, water lost from the reactor system is collected in the containment sump. Eventually the reactor is completely immersed in water and the isolation valves are opened. Decay heat removal occurs by boiling in the core. The steam generated in the core travels upward through an automatic depressurization system (ADS) valve that vents directly into containment. The density difference established in the situation depicted in Figure 8 between the core region and the pool produces a natural circulation flow that draws water up through the sump screen into the reactor vessel and is adequate in removing the decay heat. In some design cases, natural circulation inside the reactor vessel may be sufficient to remove decay heat without the need of ADS operation. This is a Category D passive safety system.

Annexes I through XX present descriptions of how different variations of these systems work in combination in various advanced water-cooled NPP designs to provide core cooling after a reactor scram.

REACTOR

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The complex set of thermal-hydraulic phenomena that occur in a gravity environment when geometrically or materially distinct heat sinks and heat sources are connected by a fluid can be identified as Natural Circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established.

The above definition includes the situations of a heater immersed into a fluid, of an open flame in the air, of a chimney driven fire, of insurge of hot fluid into a pool of cold liquid, and of a heat source and sink (e. g. heater and cooler) consisting of separated mechanical components connected by piping and situated at different gravity elevations. Natural circulation also drives the occurrence of stratification in horizontal pipes.

Within the scope of this document, this phenomenon involves the following system configurations:

(a) Heat source and sink of primary loop constituted, respectively, by core and steam generator, or boiler, or primary side of heat exchanger, with core located at a lower elevation;

(b) Heat source and sink inside the pressure vessel, constituted, respectively, by core and (typically annular-like region of) vessel downcomer. ‘Steady-state’ NC between core and downcomer occurs owing to continuous cooling of the downcomer fluid by a heat exchanger (boiler or steam generator) or by continuous inlet of feed-water liquid at a temperature lower than core outlet temperature;

(c) Cooling of the containment atmosphere by a closed loop.

In the current generation of nuclear plants, the NC core power removal capability is exploited for accident situations to demonstrate the inherent safety features of the plant (with the noticeable exception of the Dodewaard commercial BWR unit, shutdown in 1997). The natural circulation is also occurring during various phases of the refuelling.

In future generation of nuclear plants, NC is planned to be used for ensuring the nominal operating conditions and for achieving safe cooling following accidents in a wider spectrum than foreseen for current generation reactors.

The HTS is laid out with the heat transport pumps and steam generators above the core to promote natural circulation of the primary coolant for accidents when the heat transport pumps are not operating (see Fig. II-4). This natural circulation flow allows the plant to recover from any trip without relying on the heat transport pumps. Commissioning tests have been done (on current generation of CANDU plants ) to validate the effectiveness of the design feature.

Core make-up tanks (CMTs, see Fig. II-2) are provided to passively limit the extent and duration of voiding in the HTS following events that cause a rapid depletion of the HTS inventory. The CMTs are located above the tops of the steam generator U tube bundles (i. e. at the highest point in the HTS) and are normally maintained at approximately the pressure and temperature of the reactor inlet headers. ‘Flashing’ of the CMT inventory when the HTS depressurizes to below the saturation pressure of the CMTs forces a rapid flow of coolant into the HTS, thus maintaining the HTS at a relatively high pressure and fully filled with water.

Keeping the HTS full and free of void ensures the thermosyphoning capability and allows operation of the LTC pumps (if available) in the shutdown-cooling mode (see Section 2) without the risk of void entrainment and consequential cavitation of the LTC pumps.

Thermosyphoning in the HTS is supported by provision of feedwater to the secondary side of the steam generators for heat removal. There are several feedwater options available in ACR-1000. The main feedwater system (using either the main or start-up feedwater pumps) and the four-quadrant EFW system are capable of supplying feedwater to the SGs at full pressure. If these active sources are not available, passive design feature of supply is provided by gravity from the RWS (see Fig. II-3) after auto-depressurization of the SGs.

With the HTS full and free of void, and with a continuous supply of feedwater to the secondary side, thermosyphoning can continue indefinitely to remove heat from the fuel.

SUPPRESSION

This section describes the types of advanced reactor passive safety systems for removing the heat from the containment and reducing pressure inside containment subsequent to a loss of coolant accident. The types of passive safety systems being incorporated for this function are:

• Containment pressure suppression pools

• Containment passive heat removal/pressure suppression systems

• Passive containment spray

A brief description of each passive safety system for pressure suppression and containment cooling is provided in the following sections. Combinations of these systems are incorporated into the designs described in the Annexes I to XX.

3.1. Containment pressure suppression pools

Containment pressure suppression pools have been used in BWR designs for many years. Figure 9 presents a generic concept of a suppression pool. Following a LOCA, steam is generated into the drywell (the primary containment) following vaporization of liquid and/or steam expansion, both of these coming from the primary system typically due to a break. From drywell the steam-non condensable mixture is subsequently forced through large vent lines submerged in the water in the suppression pools. The steam condenses, thus mitigating a pressure increase in the containment. This is a Category B and C passive safety system.

Large pools may have a very wide spectrum of geometric configurations. Heat transfer in one very limited zone in terms of volume (e. g. by condensing injected steam or by heat transfer from a passive containment cooler) does not imply homogeneous or nearly homogeneous temperature in the pool. Many containment phenomena require steam-liquid interface. Steam discharge into a suppression pool of boiling water reactor is a good example of this case. After break-up of the originally created bubbles in the suppression pool, the subsequent formation of bubble plumes takes place. Consequently, complete condensation occurs and this induces mixing in the pool, the process is being determined by single and two-phase natural circulation. It is important to understand the break-up and plume-stirring process and mechanisms, because the system pressure ultimately controlled by the pressure in the vapour space above the water surface in the suppression chamber. This pressure is the sum of the partial pressures of steam and gas, the former controlled by the temperature at the pool surface. In turn, the pool surface temperature depends on the efficiency of steam condensation in the pool, and the degree of mixing in the pool.

The following is a listing of the steam-liquid interactions related phenomena:

• Direct contact condensation of steam in pool water

• Bubble formation and break-up and the subsequent formation of bubble plums

• Break-up and plume-stirring process and mechanisms inducing mixing in the pool

As example for steam-liquid interactions can be given passive containment cooling (PCC) venting into the suppression pool of ESBWR and also injection of steam-gas mixture through a downcomer vent line into the suppression pool.

In a CANDU reactor the fuel is contained within pressure tubes that run through a vessel called the Calandria. Low pressure and temperature heavy water contained in a Calandria vessel moderates nuclear fission. The pressure tubes are contained within another set of tubes called calandria tubes, separated from the pressure tubes by an annulus gap filled with carbon dioxide. This keeps the high pressure, high temperature coolant on the primary side separated and thermally isolated from the low pressure, low temperature moderator.

The moderator system consists of a closed heavy water recirculating loop that serves to cool and circulate the heavy water moderator through the calandria (see Fig. II-5). If the moderator pumps fail, natural circulation flows prevent formation of hot spots inside the Calandria vessel.

For accidents when all means of normal fuel cooling and long term decay heat removal are not available, the fuel and fuel channels will heat up until the pressure tubes contact the Calandria tubes, resulting in direct conduction of heat to the moderator. If the forced circulation moderator cooling system not available to remove heat, natural circulation inside the Calandria prevents further damage to the fuel channels and keeps the accident from progressing to a severe accident.

Without cooling, the heat transferred to the moderator ultimately exceeds the latent heat of vaporization, and steam is released to the RB atmosphere. Passive make-up water supply from the RWS (by gravity) will keep the Calandria full (see Fig. II-3). Adequate natural circulation flow will be maintained indefinitely for long term decay heat removal.

The LTC system design allows the system to recover and cool water from the RB sumps and pump it back to the RWS. Alternatively, the RWS inventory can be made up by supply from the firewater system. The replenished RWS inventory can then be directed to the moderator. This permits thermosyphoning in the moderator to be sustained, thereby preventing severe core damage, for an unlimited period of time.

This type of passive safety system uses an elevated pool as a heat sink. Steam vented in the containment will condense on the containment condenser tube surfaces to provide pressure suppression and containment cooling. Three variations of the concept are presented in Figures 10 to12. In the first variation of the concept, Figure 10, an air heat exchanger (HEX) is connected with a pool located on the top of the containment. Single phase liquid is expected to flow inside the HEX driven by gravity gradient caused by the inclination of the same HEX. Experiments have been performed to prove the validity of the solution. In the second variation of the concept, Figure 11, a closed loop filled with single phase liquid connects an air HEX and a pool-type HEX. Natural circulation and heat removal capability are generated when the air HEX receives heat from the containment: this occurs

through liquid heating and stratification that produces a difference between densities in the rising and descending leg of the pool-type HEX. In the third variation of the concept, Figure 12, two different zones of the containment, typically characterized by different pressures in case of accident (pressure is the same during normal operation), are connected with the rising and the descending side of a pool — type HEX. In this case, the steam-air mixture is the working fluid with condensate in the descending leg. Driving forces may be lower than in the previous cases and working condition may not be stable over a reasonably wide range of conditions. Positive driving forces may be low in all three cases and careful system engineering is needed. These passive safety systems are of Categories B and D.

FIG. 10. Containment pressure reduction and heat removal following a LOCA using steam condensation on condenser tubes.

3.2.
Passive containment spray systems

Figure 13 shows a design that implements a natural draft air cooled containment. Subsequent to a LOCA, steam in contact with the inside surface of the steel containment is condensed. Heat is transferred through the containment wall to the external air. An elevated pool situated on top of the containment provides a gravity driven spray of cold water to provide cooling in a LOCA scenario. The air flow for the cooling annulus, that is generated by a chimney-like type effect, is a Category B passive safety system. The containment vessel sprays are a Category D passive safety system.

FIG. 13. Containment pressure reduction and heat removal following a LOCA using a passive containment spray and natural draft air.

The geometrical and layout characterization of passive safety systems constitutes the key subject for Sections 2 and 3. The thermal-hydraulic characterization for the same systems requires the consideration of phenomena, i. e. passive safety system phenomena.

Therefore the aim of Section 4 is twofold:

• To classify thermal-hydraulic phenomena for passive systems;

• To establish a correlation between systems (as described in Sections 2 and 3)

(bullet above).

Gravity driven cooling provides emergency core cooling water by gravity draining, in events with loss of coolant. This system requires a large volume of water above the core, plus additional depressurization capacity, so that the primary coolant system can be depressurized to allow for gravity flow from the elevated suppression pool. Since there are no large reactor vessel pipes at or below the core elevation, this design ensures that the core will remain covered by water during all design basis accidents. In general, gravity driven cooling concept is mainly based on the depressurization of the reactor pressure vessel to sufficiently low pressures to enable reflood of the core by gravity feed from an elevated pool. When the gravity driven cooling operates, the gravity drain flow rate to the reactor pressure vessel depends on the piping geometry, the state of the fluid, and the pressure conditions in both the water pool and the reactor pressure vessel. Flow entering the reactor pressure vessel during the later stages of blowdown during a postulated loss-of coolant accident (LOCA) must be sufficient to keep the nuclear core flooded. The system which provides gravity driven cooling is a simple and economical safety system.

The following is a listing of the gravity driven cooling related phenomena:

• Depressurization of the reactor pressure vessel by discharging through depressurization valves into the drywell and increase of pressure in the upper part of containment;

• Evaporation in the reactor pressure vessel due to depressurization;

• Friction in the gravity driven cooling system and injection lines including the valves in these lines;

• Large amounts of cold water immediately floods the lower parts of the reactor pressure vessel, causing:

— Collapse of voids

— Condensation of steam

— Suppression of boiling

— Increase of water level inside the reactor pressure vessel;

• Condensing of steam out of the reactor pressure vessel and drywell gas space until air accumulates on the primary sides of the passive containment cooling system, resulting in termination of steam condensation.

As examples for gravity driven cooling system can be given gravity driven cooling system (GDCS) of ESBWR and passive core flooding system (HA-2 hydraulic accumulators of the second stage) of WWER-1000/392 and passive core flooding systems (ECCS tank)of WWER-640/407.