Isolation condensers (ICs) for removal of decay heat are immersed in the gravity driven water pool (GDWP). The pool is having a capacity of 6000 m3 of water and is divided into 8 symmetry sectors, each containing one IC. In normal operation the pool water circulates through heat exchangers to maintain the pool temperature. Figure III-1 shows the recirculation and cooling system of GDWP water. The decay heat generated in the reactor during shut down is stored in the form of sensible heat of water. However, stratification may influence heat transfer to pool to a great extent and heat storage capacity of the pool in the form of sensible heat is significantly reduced. The GDWP acts as heat sink for passive containment cooling system also (Fig. III-3).

The safety systems that will be employed in the CANDU-SCWR will be based on those employed in the ACRTM design [3]. However, an additional passive safety system that utilizes the separation between moderator and coolant in the CANDU design will also be used with enhancements that are expected to significantly reduce the probability of core damage.

IX — 2.1. ACR Passive safety systems

These systems are described elsewhere in this document for the ACR-1000 design. Differences between the CANDU-SCWR and the ACR-1000 will be taken into consideration when selecting which of these systems will be selected in the reference CANDU-SCWR design. For example, the reserve water system (RWS) will not be used to provide cooling via the steam generators since the CANDU-SCWR uses a direct cycle. Implementation of the following ACR passive safety systems in the CANDU-SCWR is currently under evaluation:

™ ACR™ (Advanced CANDU Reactor™) is a trademark of Atomic Energy of Canada Ltd (AECL).

a) Two independent shutdown systems: these consist of shutoff rods and liquid injection shutdown systems. Both systems are located in the low pressure moderator system and are driven either by gravity (spring-assisted shutoff rods) or pressurized gas (poison liquid injection).

b) Emergency coolant injection (ECI) system: the ECI system consists of accumulator tanks filled with make-up water and pressurized by compressed gas to provide emergency cooling when the core pressure falls below the pressure of the accumulators.

c) Reserve water system (RWS): this system provides gravity-driven cooling water from the reserve water tank (RWT) to several systems such as the containment spray cooling system and the moderator cooling system.

d) Containment cooling system: this system consists of containment cooling spray and is actuated automatically for any event resulting in pressures or temperatures that challenge the integrity of the containment. Once actuated, operation of the sprays relies only on gravity to deliver water from the RWT to the spray headers, which are located at a high elevation in the reactor building. Spray nozzles connected to the spray headers diffuse the cooling water into fine droplets, which fall through the containment atmosphere [3].

The overall approach to safety in IRIS may be represented by the following three-tier approach:

1. The first tier is the safety-by-design™, which aims at eliminating by design the possibility for an accident to occur, rather than dealing with its consequences. By eliminating some accidents, the corresponding safety systems (passive or active) become unnecessary as well.

2. The second tier is provided by simplified passive safety systems, which protect against the still remaining accidents and mitigate their consequences.

3. The third tier is provided by active systems, which are not required to perform safety functions (i. e. are not safety grade) and are not considered in deterministic safety analyses, but do contribute to reducing the core damage frequency (CDF).

XIII — 3.1. First tier

The first tier is embodied in the IRIS ‘safety-by-design’™. Nuclear power plants consider a range of hypothetical accident scenarios. The IRIS ‘safety-by-design’™ philosophy is a systematic approach that aims—by design—at eliminating altogether the possibility for an accident to occur, i. e. to eliminate accident initiators, rather than having to design and implement systems to deal with the consequences of the accident. It should be noted that the integral configuration is inherently more amenable to this approach than a loop-type configuration, thus enabling safety improvements not possible in a loop reactor. To give only the most obvious example, loss of coolant accidents caused by a large break of external primary piping (LBLOCA) are eliminated by design since no large external piping exists in IRIS. Additionally, in cases where it is not possible or practical to completely eliminate potential initiators of an accident, safety-by-design™ aims at reducing the severity of the accident’s consequences and the probability of its occurrence. As a result of this systematic approach, the eight Class IV design basis events (potentially leading to most severe accidents) that are usually considered in LWRs, are reduced to only one in IRIS, with the remaining seven either completely eliminated by design, or their consequences (as well as probability) reduced to a degree that they are no longer considered Class IV events.

The second tier consists of the passive safety systems needed to cope with the still-remaining potential accidents. Notably, the elimination of the possibility for some accidents to occur enables simplifications of IRIS design and passive safety systems, resulting simultaneously in enhanced safety, reliability, as well as economics. In other words, the increased safety and improved economics support each other in the IRIS design.

Figure V-5 illustrates the automatic depressurization system, the passive safety injection, and sump recirculation flow paths and components.

After the lower containment sump and the IRWST liquid levels are equalized, the sump valves are opened to establish a natural circulation path. Primary coolant is boiled in the reactor core by decay heat. This low-density mixture flows upward through the core and steam and liquid is vented out of the ADS-4 lines into containment. Cooler water from the containment sump is drawn in through the sump screens into the sump lines that connect to the DVI lines.

Containment passive heat removal system (C-PHRS) removes heat from the containment in case of a LOCA and is designed to fulfil the following functions: (1) emergency isolation of service lines penetrating the containment and not pertaining to systems intended to cope with the accident; (2) condensation of the steam from the containment atmosphere; (3) retention of radioactive products released into the containment; (4) control of the iodine levels released into the containment atmosphere. The steam from the containment atmosphere condenses on the internal steel wall of the double-containment being cooled from the outside surface by the water stored in the tank.

The system operates due to natural circulation of the containment atmosphere and water storage tank. The design basis of this system is to condense the amount of steam equivalent to decay heat release within 24 hours after reactor trip without water storage tank replenishment. This system does not require electric power supply to operate. The system consists of four independent trains with the redundancy of 4 x 33%. A diagram of the passive heat removal system is shown in Figure XII-2.

FIG. XII-2. Passive heat removal system for LOCAs (1 — cooler of primary containment; 2 — discharge pipeline; 3 — emergency heat removal tank; 4 — make-up pipeline).

The integral SCOR design incorporates PWR technology proven by more than 40 years of experience, and adopts passive safety systems studied at CEA in the 1990s. The safety approach of SCOR strengthens the prevention of abnormal operation and acceptability of failures for a large set of conditions, thereby reducing the expected frequencies of accident initiators and consequences. The design approach reduces the number and complexity of the safety systems and simplifies the required operator actions compared to those of a standard loop type PWRs (Generation II).

Nuclear reactor characteristics described here are quite similar to well-known PWRs (core geometry and materials, reactor control type, etc.). The SCOR design is characterized by the following innovative solutions:

5 See, for example, the detailed description of the effect of ‘safety by design’ given in IAEA-TECDOC-1391 Status of Advanced LWR Designs: 2004 in the description of the IRIS design (p. 591-592).

• Suppression of the large diameter connections on the reactor pressure vessel;

• Passive and integrated emergency core cooling system, based only on natural circulation and using external air as ultimate heat sink;

• Reactor operating with a soluble boron-free core; the control rod mechanisms are integrated in the vessel;

• Choice of a small power density, enabling large operating margin on core power parameters (i. e. to DNBR);

• Easy testability and maintenance of all safety systems.

Consequences of a significant number of accidents are either outright eliminated or reduced by conception, i. e. without any need of active or passive systems. The major safety systems are passive; they require no operator action or off-site assistance for a long time after an accident. Moreover, core and containment cooling is provided for lasting a long time without alternative current power. This approach is generic for different reactor designs.

Although the SLCS performs no design basis safety-related functions, it is classified as safety-related system. The SLC system is designed to provide makeup water to the RPV to mitigate the consequences of a LOCA. The ECCS and the SLC are designed to flood the core during a LOCA to provide sufficient core cooling. By providing core cooling following a LOCA, the ECCS and SLCS, in addition to the containment, limit the release of radioactive materials to the environment. The SLCS provides the reactor with an additional liquid inventory in the event of DPV actuation. This function is accomplished by firing squib-type injection valves to initiate the SLCS. The SLCS is operational at the high pressure. The SLCS is pressurized by utilizing a nitrogen charging subsystem including a liquid nitrogen tank, vaporizer, and high pressure pump for initial accumulator charging.

The SLCS is manually initiated for its shutdown function. In addition, the SLCS is automatically initiated for events beyond the safety design basis, such as a LOCA event. The SLCS contains two identical and separate trains. Each train provides 50% injection capacity. In addition to providing water inventory to the RPV, the SLCS removes the remaining reactivity by injecting boron solution into the RPV after the reactor shutdown.

The passive core catcher increases the safety barrier and confinement by preventing radioactive material releases from the primary circuit and the reactor vessel. The passive core catcher provides receiving and subsequent cooling of liquid and hard corium fractions released from the damaged reactor vessel. It is installed in a concrete pit below the reactor vessel. This device comprises four variable parts located (top-to-bottom) in the direction of corium movement from the reactor vessel to the concrete pit basement. The variable parts are the lower plate, vent header, barrel with the filler, and heat exchanger.

The lower plate is designed as a guiding structure, similar to a funnel, in which the corium from damaged reactor vessel flows into the passive core catcher.

The vent header is installed under the lower plate and is designed as a thermal shield protecting the thermal insulating structures. It also protects the reactor vessel from corium outflow installed on the concrete cantilever and in the foundation of lower plate at the stage of corium outflow from the reactor vessel.

This allows to increase operating period of the specially installed thermal shields and to lower the intensity of their damage in the course of radiation heat exchange with corium and aerosol atmosphere in the concrete pit sub-reactor room.

The vent header increases the operating time and reduces the radiation intensity to the thermal shields.

The barrel with filler functions as a corium diluent and thermal absorber for the surrounding peripheral structures (the core catcher). The optimum content of ferric, aluminium oxides, and structural steel in sacrifice material allows to lower volumetric power density in the corium melt, release of gas and radionuclide masses into the confinement, melt temperature.

The evaporation of water causes a period of corium direct cooling through a heat exchanger. The heat exchanger is gravity fed from the inspection well and begins after the corium chemical reaction with the filler and melting of barrel steel structures. Heat exchanger passive make-up and corium direct cooling by the water supplied through concrete cantilever pipelines allows us to avoid human intervention for 24 hours (the moment of reactor vessel damage with corium).

Korea Atomic Energy Research Institute, Republic of Korea

XVIII — 1. Introduction

The SMART (System-Integrated Modular Advanced ReacTor) is an advanced pressurized light water reactor that is being continuously studied at KAERI (Korea Atomic Energy Research Institute) with a rated thermal power of 330 MW. The reactor is proposed to be utilized as an energy source for sea water desalination as well as for small scale power generation. Advanced technologies such as inherent and passive safety features are incorporated in establishing the design concepts to achieve inherent safety, enhanced operational flexibility, and good economy. The SMART is designed to supply 40,000 tons of fresh water per day and 90MW of electricity to an area with an approximate population of 100,000 or an industrialized complex. In order to demonstrate the relevant technologies incorporated in the SMART design, the SMART-P (i. e. a Pilot plant of the SMART) project is currently underway at KAERI.

The prominent design feature of SMART is the adoption of integral arrangement. The major components of the NSSS such as the core, steam generators, main coolant pumps, and pressurizer are integrated into a reactor vessel without any pipe connections between those components. The schematic diagram of the SMART NSSS is shown in Figure XX-1.

The SMART core is currently being designed with the fuel design based on existing Korea Optimized Fuel Assembly (KOFA) which is in 17 x 17 rectangular rod arrays. The SMART core design is characterized by an ultra long operation cycle with a single or modified single batch reload scheme, low core power density, soluble boron-free operation, enhanced safety with a large negative Moderator Temperature Coefficient (MTC) at any time during the fuel cycle, a large thermal margin, inherently free from xenon oscillation instability, and minimum rod motion for the load follow with coolant temperature control. Due to soluble boron-free operation, an important design requirement for the SMART CEDM is a fine maneuvering capability to control the excess core reactivity. A linear step motor type CEDM is employed for easy maintenance. The minimum step length is 4mm that is short enough for the fine reactivity control. Forty-nine CEDMs are installed in the fifty-seven fuel assemblies of the SMART core.

Twelve identical SG cassettes are located in the annulus formed by the RPV and the core support barrel. Each SG cassette is a once-through type with helically coiled tubes wound around the inner shell. The primary coolant flows downward in the shell side of the SG tubes, while the secondary feedwater flows upward in the tube side. Therefore, the tubes are under compressive loads from the greater primary pressure, reducing the stress corrosion cracking and thus reducing the probability of tube rupture. The 40°C superheated steam at the exit of the helically coiled tubes eliminates the necessity of a steam separator during normal operations. The twelve SGs are divided into four sections. Each section consists of the neighboring three steam generator cassettes which are connected together with the steam and feedwater pipes. If there is a leakage in one or more of the tubes, the relevant section is isolated and SMART can be operated with reduced power until the scheduled shutdown.

The SMART adopted an in-vessel self-controlled pressurizer (PZR) located in the upper space of the RPV. The volume of the PZR is filled with water, steam, and nitrogen gas. The self-pressurizing design eliminates the active mechanisms such as spray and heater. The system pressure is determined by a sum of the steam and nitrogen partial pressures. In order to minimize the contribution of the steam partial pressure, a PZR cooler is installed for maintaining the low PZR temperature, and wet thermal insulator is installed to reduce the heat transfer from the primary coolant. The coolant temperature of the core outlet is controlled during a power maneuvering so as to minimize the system pressure variation by counterbalancing the increase of the coolant volume of the hot part with the decrease of the coolant volume of the cold part.

The SMART MCP is a canned motor type pump that eliminates the problems connected with conventional seals and associated systems. Four MCPs are installed vertically on the RPV annular cover. MCP is an integral unit consisting of a canned asynchronous 3-phase motor and an axial flow single-stage pump. The motor and pump are connected through a common shaft rotating on three radial and one axial thrust bearings. The impeller draws the coolant from above and discharges downward directly to the SG. This design minimizes the pressure loss of the flow.

There are many inherent safety features in the SMART design. Those include a large negative moderator temperature coefficient due to the boron-free operation, a low core power density, and the reduced xenon oscillations. Furthermore, enhanced safety of the SMART is accomplished with highly reliable engineered safety systems. The engineered safety systems consist of a reactor shutdown system, passive residual heat removal system, emergency core cooling system, safety vessel, reactor overpressure protection system and containment overpressure protection system. As the result of the probabilistic safety assessments for 10 internal events, the core meltdown frequency is predicted as 8.56 x 10’7.

Containment is a key component of the mitigation part of the defence in depth philosophy, since it is the last barrier designed to prevent large radioactive release to the environment.

In advanced heavy water reactor (AHWR), a passive containment cooling system (PCCS) is envisaged which can remove long term heat from containment following loss of coolant accident (LOCA). Immediately following LOCA, steam released is condensed in water pool by vapor suppression system. For subsequent long term cooling, PCCS is provided. PCCS, by definition is able to carry out its function with no reliance on external source of energy.

As mentioned earlier, gravity driven water pool acts as the heat sink for a number of passive heat removal systems including the PCCS. The passive external condensers (PECs) of the PCCS are connected to the pool as shown in Fig. III-3. The containment steam condenses on the outer surface of the tubes of PEC. The water inside the tubes takes up heat from air/vapor mixture and gets heated up. Due to the heating up of water, the natural circulation of water from the pool to PEC and from PEC to pool is established.

One important aspect of PCCS functioning is the potential degradation of heat transfer on PEC outer surface due to the presence of noncondensable gases in the containment. The presence of noncondensable (NC) gases in vapor can greatly inhibit the condensation process. Extensive R&D work is in progress to address this issue. Another aspect of PCCS functioning is the blockage of passive external condenser by noncondensable gas due to the stratification of noncondensable gas/vapor in the containment. In case of AHWR, the noncondensable gas is likely to escape through the central opening provided in the GDWP to the dome region. Experiments are planned to confirm this.