IRIS implements a passive emergency heat removal system made of four independent subsystems, each of which has a horizontal, U tube heat exchanger connected to a separate SG feed/steam line. These heat exchangers are immersed in the refuelling water storage tank (RWST) located outside the containment structure. The RWST water provides the heat sink to the environment for the EHRS heat exchangers. The EHRS is sized so that a single subsystem can provide core decay heat removal in the case of a loss of secondary system heat removal capability. The EHRS operates in natural circulation, removing heat from the primary system through the steam generators heat transfer surface, condensing the steam produced in the EHRS heat exchanger, transferring the heat to the RWST water, and returning the condensate back to the SG. The EHRS provides both the main post-LOCA depressurization (depressurization without loss of mass) of the primary system and the core cooling functions. It performs these functions by condensing the steam produced by the core directly inside the reactor vessel. This minimizes the break flow and actually reverses it for a portion of the LOCA response, while transferring the decay heat to the environment.

XVI-4.2. Emergency boration tanks (EBT)

IRIS has two full-system pressure emergency boration tanks (EBTs) to provide a diverse means of reactor shutdown by delivering borated water to the RV through the direct vessel injection (DVI) lines. By their operation these tanks also provide a limited gravity feed makeup water to the primary system.

The most effective means of describing the function of each of these passive safety systems is to relate their operation in response to a small break loss of coolant accident (SBLOCA). The first phase of a SBLOCA is the subcooled blowdown phase. During this phase, high pressure subcooled liquid is venting from the break under choked flow conditions. The primary system pressure and primary liquid inventory will be decreasing. When low pressure or low liquid level is sensed in the pressurizer, a safety signal is issued resulting in the following automatic actions:

• Scram reactor

• Open the PRHR inlet and outlet valves

• Open the CMT outlet valves

• Isolate Steam Generators (feedwater and main steam)

• Trip reactor coolant pumps (coastdown).

Natural circulation is established in the PRHR loop and the CMT loops. Boiling occurs on the PRHR tubes and hot water begins to fill the top of the CMTs. If the plant continues to depressurize, eventually the primary system reaches the saturation pressure corresponding to the hot leg temperature. Depending on the break size, the system pressure will reach a plateau during which the loop will experience a period of two-phase natural circulation.

If primary coolant inventory continues to decrease, eventually the CMTs will begin to drain. At a predetermined CMT level, the ADS-1 valves will open followed by the ADS 2-3 valves. System pressure will drop very quickly as a result of the ADS 1-3 venting steam into the IRWST. The primary system pressure soon drops below the accumulator tank pressure; and significant quantities of cold borated water are injected from the accumulator into the reactor vessel.

If the CMT liquid level continues to decrease, the ADS-4 actuation setpoint will be reached. The ADS-4 valves open, dropping the primary system pressure below the head pressure of the IRWST liquid. The IRWST drains by gravity into the reactor vessel, out the break and ADS 4 valves into the containment sump. Eventually the IRWST and containment sump liquid levels equalize and the sump valves are opened, establishing long term sump recirculation cooling.

Steam vented through the ADS-4 valves condense on the inside surfaces of the containment vessel. The containment vessel is externally cooled by air and water as needed. The condensate inside the containment is returned to the containment sump and IRWST, where it is available for sump recirculation.

ECCS accumulator system is intended to supply cold water at the core inlet and retain the primary coolant inventory when the primary pressure reduces below 4 MPa(when leaks occur). The system consists of four independent trains with redundancy of 4 x 50%. Each train includes an accumulator tank with boron solution under pressure; the pressure is created by a nitrogen cushion. The accumulators are connected to the reactor downcomer. The check valves are installed on the connecting pipelines to isolate the accumulators from the reactor when the primary pressure exceeds 4.0 MPa. The system operates in accordance with the passive principle, and no signal is required to activate the system.

XI — 3.2. ECCS tank subsystem

ECCS tanks (tanks of atmospheric type) are intended to flood the reactor core allowing long term residual heat removal when operating at a pressure below 0.29 MPa. The system consists of four independent trains with redundancy of 4 x 50%. Each train includes the tank filled with boron solution, valves, and pipelines. The ECCS tank is connected to the reactor downcomer by the pipeline with two check valves. The system operates under the hydrostatic force effect.

The large annular space between the core barrel shell and the reactor vessel contains the heat exchanger-pump modules. Each of the sixteen modules of Figure XIX-2 comprises a primary pump and a heat exchanger used to remove residual heat.

As shown in the figure, the submerged pump, is supplied with water exiting the steam generator. The coil-type motor is submerged downstream of the impeller. The pump motor has a stationary cylindrical stator surrounding the rotor that is connected to the pump’s impellers. Electrical power is supplied to each of the 16 pumps by cables through small penetrations in the reactor vessel. The primary water that flows around the outer ring cools the coils; therefore associated piping penetrations through the reactor vessel for cooling water are eliminated. The spool pump needs an additional external motor to provide high inertia during coastdown in order to mitigate the consequences of Loss — Of-Flow Accidents (LOFAs). This additional inertia, provided by an external motor with an adequate flywheel, is linked to the spool type pump by an electrical connection. At the outlet of the pump, water is accelerated by a venturi, passes into a diffuser and then through the decay heat exchanger tube bundles.

The decay heat exchanger of the RRP system (see XIX-4.1.2) consists of bayonet tubes whose outside surface is wetted by the primary fluid. The secondary water flows first in the internal tube and then upwards through the annular space bound by the two tubes. The water box is located in a dead zone, behind the venturi. This type of heat exchanger is of interest, as it does not require a water box at the exit of the heat exchanger. This reduces the primary pressure drop and allows the free expansion of the tubes. Thermal loadings are reduced leading to an increased mechanical resistance and an enhanced reliability.

A flow bypass is installed where the venturi is located, between the core exit and the cold leg. It allows a natural convection of the primary fluid during pump shutdown. During normal operation, high flow velocity at the venturi throat leads to a decrease of the local pressure. The cross section area of the venturi throat is designed to balance the pressure between the hot leg (core exit) and the cold leg (heat exchanger-pump module), in order to prevent bypass flow in normal condition. This primary flow layout with a venturi and an integrated heat removal system is issued from the CEA patent N° 92 05220

The decay heat exchanger-pump module can be easily extracted from the reactor vessel once the steam generator has been removed. The pump power supply and the heat exchanger secondary feed-lines are set in the vessel via a removable opening in the upper part of the reactor vessel.

During a LOCA or transient, the control rod drive system (CRDS) shuts down the reactor. After the reactor shutdown, the main concern to safety is how to remove the decay heat. To depressurize the RPV, the ADS is actuated. As shown in Fig. VI-6, the RPV pressure decreases rapidly as soon as the ADS is actuated. The steam vented through the SRVs is sent to the SP where it is condensed, but the steam vented through DPVs goes to the DW. Large amount of steam and noncondensable gases are vented from the DW to the SP through primary horizontal and vertical vents. During this period, the SP is the primary heat sink to prevent over-pressurization of the containment and the DW pressure is adjusted by venting steam through the horizontal vents into the SP.

Since the ADS is actuated at the Level 1 signal, it is necessary to remove the decay heat in order to prevent the over-pressurization of the RPV. The ICS is designed to remove the decay heat although the RPV is at very high pressure. Similarly, the SLCS is designed to makeup the water into the RPV at high pressure before the GDCS injects water.

National Atomic Energy Commission (CNEA), INVAP, Argentina

XIII — 1. Introduction

CAREM is an Argentine project to achieve the development, design, and construction of an innovative simple and small Nuclear Power Plant (NPP), which is jointly developed by CNEA (National Atomic Energy Commission) and INVAP. This nuclear plant has an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the design, and contributes to a higher safety level. Some of the high level design characteristics of the plant are: integrated primary cooling system, self-pressurised primary system and safety systems relying on passive features.

The CAREM concept was first presented in March 1984 in Lima, Peru, during the IAEA conference on small and medium size reactors. CAREM was, chronologically, one of the first of the present new generation of reactor designs. The first step of this project is the construction of the prototype of about 27 MW(e) (CAREM-25). CAREM has been recognized as an International Near Term Deployment (INTD) reactor by the Generation IV International Forum (GIF).

Several inherent and passive systems have been adopted in Indian innovative reactor, AHWR. Analyses have been performed to prove design concepts of these systems. Experiments and further analyses of these systems are being carried out rigorously. Several major areas of R & D have been identified for detailed study and the required development activities are in progress.

Extensive work has been carried out in the area of natural circulation. A transparent rectangular loop has been installed to study natural circulation. The stability of natural circulation with different heater and cooler orientations has been studied in the loop. Start-up procedure and instability studies are being carried out at high pressure natural circulation loop (HPNCL). Flow pattern transition instability studies using neutron radiography have been conducted at APSARA reactor in flow pattern transition instability loop (FPTIL). Integral behaviour experiments are being conducted in an integral test loop (ITL). The ITL (Fig. III-5) was commissioned in 2005 in BARC. The performance of isolation condensers for reactor decay heat removal is being evaluated in the ITL. A parallel channel experimental facility is set up in BARC to investigate parallel channel instability. Void coefficient of reactivity has been simulated in the loop. The pre-test single channel stability and parallel channel stability analyses by RELAP 5 and other in house codes have been carried out for the ITL and parallel channel loop. Two-phase low flow pressure drop experiments are being conducted in 3 MW boiling
water loop (BWL) across the various components of coolant channel. Earlier single phase and two phase (air-water) pressure drop experiments were performed on simulated full scale fuel bundle of AHWR in flow test facility at low pressure.

Thermal stratification inside a water pool is being investigated. One dimensional theoretical model and computer code for solving two dimensional Navier-Stokes equations have been developed to study the stratification phenomena in the water pool. Other generalized computer codes available are also being used for this purpose.

An experimental set up to study phenomena associated with passive containment isolation is being set

up.

The effect of non-condensable gas on steam condensation inside a vertical tube has been investigated experimentally. A comparison of local heat transfer coefficient determined by theoretical model with experimental data has been carried out. An experimental facility to investigate the effect of non condensable gas on condensation of steam on the outer surfaces of tubes of passive external condensers has been commissioned and experiments are in progress.

XI — 1. Introduction

In 1992, German utilities awarded FRAMATOME (former Siemens) a contract to develop a new BWR nuclear power plant using passive safety systems, and together with the utilities FRAMATOME started development work on a new BWR with a net capacity of 750 MW(e). During the conceptual phase, lasting from February 1992 until September 1993, priority was given to developing passive safety systems to replace or supplement active systems. At the end of the conceptual phase, it was decided that the new requirements for this advanced BWR, especially economic aspects, justified a concept with a reactor thermal output of 2778 MW and a net electrical output of 977 MW. Since 2000 the net electrical output was further increased to 1254 MW.

The four-year basic design phase for the resulting ‘SWR 1000’ plant started in mid-1995. In parallel, an experimental testing program was conducted at FRAMATOME’s own testing facilities and at other German and European research centers to provide verification of the mode of operation and effectiveness of the SWR 1000’s passive safety systems.

In the following sections, after a description of the general design principles and the safety systems, the passive safety components that use natural circulation are described in detail. In particular, descriptions are provided for the emergency condenser, the containment condenser and the pressure vessel cooling after a hypothetical severe accident.

The containment pressure suppression system (PSS), shown in Figure XVI-3, consists of six water tanks and a common tank for non-condensable gas storage. Each suppression water tank is connected to the containment atmosphere through a vent pipe connected to a submerged sparger so that steam released in the containment following a loss of coolant or steam/feed line break accident is condensed. The suppression system limits the peak containment pressure, following the most limiting blowdown event, to less than 1.0 MPa (130 psig), which is much lower than the containment design pressure. The suppression system water tanks also provide an elevated source of water that is available for gravity injection into the reactor vessel through the DVI lines in the event of a LOCA.

FIG. XVI-3. IRIS spherical steel containment arrangement.

The IRIS design includes a specially constructed lower containment volume that collects the liquid break flow, as well as any condensate from the containment, in a cavity where the reactor vessel is located. Following a LOCA, the cavity floods above the core level, creating a gravity head of water sufficient to provide coolant makeup to the reactor vessel through the DVI lines. This cavity also ensures that the lower outside portion of the RV surface is or can be wetted following postulated core damage events.

As in the AP600/AP1000, the IRIS safety system design uses gravitational forces instead of active components such as pumps, fan coolers or sprays and their supporting systems. The safety strategy of IRIS provides a diverse means of core shutdown by makeup of borated water from the EBT in addition to the control rods; also, the EHRS provides a means of core cooling and heat removal to the environment in the event that normally available active systems are not available. In the event of a significant loss of primary-side water inventory, the primary line of defence for IRIS is represented by the large coolant inventory in the reactor vessel and the fact that EHRS operation limits the loss of mass, thus maintaining a sufficient inventory in the primary system and guaranteeing that the core will remain covered for all postulated events. The EBT is capable of providing some primary system injection at high pressure, but this is not necessary, since the IRIS strategy relies on ‘maintaining’ coolant inventory, rather than ‘injecting’ makeup water. This strategy is sufficient to ensure that the core remains covered with water for an extended period of time (days and possibly weeks).

Thus, IRIS does not require and does not have the high capacity, safety grade, and high pressure safety injection system characteristic of loop reactors. Of course, when the reactor vessel is depressurised to near containment pressure, gravity flow from the suppression system and from the flooded reactor cavity will maintain the RV coolant inventory for an unlimited period of time. However, this function would not be strictly necessary for any reasonable recovery period since the core decay heat is removed directly by condensing steam inside the pressure vessel, thus preventing any primary water from leaving the pressure vessel.

The IRIS design also includes a second means of core cooling via containment cooling, since the vessel and containment become thermodynamically coupled once a break occurs. Should cooling via the EHRS be defeated, direct cooling of the containment outer surface is provided and containment pressurization is limited to less than its design pressure. This cooling plus multiple means of providing gravity driven makeup to the core provide a means of preventing core damage and ensuring containment integrity and heat removal to the environment that is diverse from the EHRS operation.

REFERENCES TO ANNEX XVI

[1] COLLADO, J. M., Design of the reactor pressure vessel and internals of the IRIS integrated nuclear system, Advanced Nuclear Power Plants (Proc. Int. Congress Cordoba, Spain, 2003), ICAPP03- ISBN: 0-89448-675-6 (2003).

[2] INTERNATIOAL ATOMIC ENERGY AGENCY, Innovative small and medium sized reactors: Design features, safety approaches and R&D trends: Final report of a technical meeting held in Vienna, 7-11 June 2004, IAEA-TECDOC-1451, Vienna (2005).

The main control room (MCR) is also augmented by passive systems. The main control room habitability system (VES) will pressurize, cool, and provide fresh air to the MCR in the event of an accident. This system is initiated by a high radiation signal in the MCR. It isolates the MCR ventilation system and pressurizes the room to slightly higher than atmospheric pressure, thereby limiting infiltration by airborne contaminants.

V-5. Conclusions

The AP600 was the first passively safe nuclear plant to be certified in the United States. The certification was based on comprehensive integral system and separate effects testing conducted by Westinghouse and the U. S. Department of Energy at the SPES test facility in Italy and at the APEX test facility at Oregon State University. The U. S. Nuclear Regulatory Commission (NRC) conducted confirmatory tests at the ROSA-AP600 test facility in Japan and the APEX Facility at Oregon State University. The AP600 received Final Design Approval from the NRC in September, 1998 and Design Certification in December, 1999. The AP1000 received NRC Final Design Certification in January 2006.