The largest LOCA considered (2 x 50 mm) is the break of the line between the vessel and the boiler of the pressurizer, Figure XIX-3. At the beginning, the power is removed by the break and by the steam generator. As for blackout, the RRPa reach their full power in one thousand seconds. After a stabilisation at the pressure of the secondary safety valve, the primary pressure reaches the threshold pressure of the safety injection system at 4 000 s.

Steam generator tube rupture

During the first thousands seconds, the residual power is removed by the SG and by the RRP system. To prevent steam release to the atmosphere, the steam is condensed in a dedicated pool. With 4 RRP loops and 5 tubes ruptured, the mass of released steam is about 40 to 50 tons according to this heat sink of the RRP; with 8 RRPa, the released steam is 20 tons. After six thousand seconds, the RRP system is enough to cool the reactor and the released steam by SG is stopped.

Conclusion of the design basis accident management

The intrinsic behaviour of SCOR is considerably improved comparing a standard PWR. This is due to the large thermal inertia, the elimination of the large LOCAs and the suppression of soluble boron. Without any action (i. e. no RRP and no safety injection), the delay before core dewatering is one and half-hour after the beginning of the most penalizing transient.

Safety calculations showed that all the transients were correctly managed in a passive way (in the vessel, in the RRP loop, and in the heat sink) with only 4 out of 16 RRP, whatever the heat sink: pool or air-cooling tower. This represents a redundancy of 16 times 25%. The RRP operation is compatible with an active or passive way whatever the primary pressure or the temperature. The in-vessel heat exchangers of the RRP loop being located very close to the core, and thanks to the flow bypass of the venturi, the RRPs are operational in two-phase flow mode in case of small primary water inventory. Long term cooling may be ensured in total passive condition thanks to the RRP with air-cooling tower. Only a safety injection at 20 bars with a small flow rate is needed one hour after the beginning of the most penalizing LOCA, that is a double break of the pressurizer line (2 x 50 mm). In the event of SGTR, the steam released from the safety valves of the secondary circuit is condensed in a dedicated pool. There’s no released steam in the atmosphere.

The comparison of the typical design basis events between standard PWRs and SCOR is summarized in Table XIX-1.

During normal reactor shut down core decay heat is removed by passive means utilizing Isolation condensers (ICs) immersed in a gravity driven water pool (GDWP) located above the steam drum. Core decay heat, in the form of steam enters the IC tube bundles. The steam condenses inside the tubes and heat is transferred to the surrounding water pool. The condensate returns by gravity to the steam drum. The water inventory in the GDWP is adequate to cool the core for more than 3 days without any operator intervention and without boiling of GDWP water. Fig. III-2 depicts the core decay heat removal system comprising isolation condensers. A separate GDWP cooling system is provided to cool the GDWP inventory in case the temperature of GDWP inventory rises above a set value.

X — 1. Introduction

The CANDU®-SCWR is a pressure tube SCWR reactor concept under development by AECL as part of the Canadian Generation IV program. The main mission of this reactor is expected to be electricity production but other non-electricity applications such as hydrogen production and process heat are also being investigated. Figure X-1 shows a possible layout of the plant. Preliminary design parameters are shown in Table X-1 [2].

The coolant enters the reactor core into the individual pressure tubes at 25MPa and a subcritical temperature of 350oC. The temperature of the coolant rises above the critical point along the fuel channels and exits at about 625oC. The coolant enters the supercritical turbine, which is compact and can be placed inside the containment. Depending on the mission of the reactor, the high — energy stream from the supercritical turbine outlet can be used to generate electricity using conventional turbines or can be used for non-electricity applications.

Heat for co-generation or intermediate/low pressure turbines Drinking water

FIG. X-1. CANDUSCWR schematic.

® CANDU is a registered trademark of Atomic Energy of Canada Ltd (AECL).

The CANDU-SCWCR is similar to a typical CANDU reactor but with the following major differences:

• The lattice pitch is tighter to reduce heavy water cost,

• The coolant is light water at supercritical conditions,

• Uses a modified fuel channel design with internal insulation to accommodate the higher coolant pressure and temperature.

The IRIS reactor vessel (RV) [1] houses not only the nuclear fuel and control rods, but also all the major reactor coolant system (RCS) components (see Fig. XVI-1): eight small, spool type, reactor coolant pumps (RCPs); eight modular, helical coil, once through steam generators (SGs); a pressurizer located in the RV upper head; the control rod drive mechanisms (CRDMs); and, a steel reflector which surrounds the core and improves neutron economy, as well as it provides additional internal shielding. This integral RV arrangement eliminates the individual component pressure vessels and large connecting loop piping between them, resulting in a more compact configuration and in the elimination of the large loss-of-coolant accident as a design basis event.

As the IRIS integral vessel contains all the RCS components, it is larger than the RV of a traditional loop-type PWR. It has an inner diameter of 6.21 m and an overall height of 22.2 m including the closure head. Water flows upwards through the core and then through the riser region (defined by the extended core barrel). At the top of the riser, the coolant is directed into the upper part of the annular plenum between the extended core barrel and the RV inside wall, where the suction of the reactor coolant pumps is located. Eight coolant pumps are employed, and the flow from each pump is directed downward through its associated helical coil steam generator module. The primary flow path continues down through the annular downcomer region outside the core to the lower plenum and then back to the core completing the circuit. Additional details of the IRIS integral vessel can be found in IAEA-TECDOC-1451. [2]

The accumulators are similar to those found in conventional PWRs. They are large spherical tanks approximately three-quarters filled with cold borated water and pre-pressurized with nitrogen. The accumulator outlet line is connected to the DVI line. A pair of check valves prevents injection flow during normal operating conditions. When system pressure drops below the accumulator pressure (plus the check valve cracking pressure), the check valves open allowing coolant injection to the reactor downcomer via the DVI line.

V-2.5. In-containment refueling water storage tank (IR WST)

The In-containment refueling water storage tank is a very large concrete pool filled with cold borated water. It serves as the heat sink for the PRHR heat exchanger and a source of water for IRWST injection. The IRWST has two injection lines connected to the reactor vessel DVI lines. These flow paths are normally isolated by two check valves in series. When the primary pressure drops below the head pressure of the water in the IRWST, the flow path is established through the DVI into the reactor vessel downcomer. The IRWST water is sufficient to flood the lower containment compartments to a level above the reactor vessel head and below the outlet of the ADS-4 lines.

XII — 1. Introduction

The design of WWER-640/407 (V-407) was developed by the FSUE SPAEP (St. Petersburg, Russian Federation), FSUE EDO ‘Gidropress’ (Podolsk, Russian Federation) and the Russian National Research Centre ‘Kurchatov Institute’ (Moscow).

The design basis of the V-407 is that the prescribed exposure dose limits and the standards for the release of radioactive substances into the environment should not be exceeded under normal operation, anticipated operational occurrences, and in design and beyond-design-basis accidents, for the life (50-60 years) of the plant.

For this purpose, the new WWER-concept V-407 makes wide use of passive systems and components. A number of relatively innovative passive safety systems are implemented in these designs to ensure the fundamental safety functions: reactivity control, fuel cooling, and the confinement of radioactivity.

One important problem related to the implementation of passive systems is the lack of sufficient operating experience using the passive systems/components under actual plant conditions. Besides, the existing computer codes for transient and accident analysis are not sufficiently validated for the conditions and phenomena which are relevant to the passive system functioning (low pressure, low driving pressure and temperature heads, increased effect of non-condensable, boron transport at low velocities, etc.).

The new design features are envisaged to be verified experimentally at a large-scale test facility (1:27 volume and power scale). The design is developed in accordance with the latest Russian safety regulations for nuclear power plants, which meet modern world requirements.

The passive systems of WWER-V-407 using natural circulation consist of:

• Containment passive heat removal system (C-PHRS)

• Emergency core cooling system with three subsystems

• Steam generator passive heat removal system (SG-PHRS).

The overall configuration of the WWER-640/407 reactor systems is shown in Figure XII-1.

XVIII — 1. Introduction

SCOR (Simple COmpact Reactor) is a 2000 MW*th, integrated pressurized light water reactor (PWR), whose design was developed at the Nuclear Reactor Division of Commissariat a l’Energie Atomique at Cadarache in France.

The plant design, begun in 2000, was mainly defined for electricity production but could be adapted for desalination. The net power is 630 MW(e) (31.5% net efficiency). The SCOR concept respects the European Utility Requirements. The SCOR design is based on a compact reactor vessel that contains all the reactor coolant system components, including the pressurizer, the reactor coolant pumps, the control rod drive mechanism and the dedicated heat exchangers of the passive decay heat removal systems. The single steam generator is located above the reactor vessel. When applicable, the SCOR development adopted the generic notion of « Safety by design »5, sort of ‘design oriented safety approach’ which give the priority to a whole plant inherent behaviour able to meet an increased and well mastered level of safety vis a vis of the different safety functions.

The WW is a large chamber with connection to the DW. During the initial blow down, the WW is directly communicated with the DW through the horizontal vents. For long term phase of the LOCA, the WW is communicated with the DW through the PCCS supply lines, condensers and the PCCS vents. When the WW pressure increases above the DW pressure, the vacuum breaker check valves are open. This action ensures that the WW is depressurized by discharging noncondensable gas from the WW to DW and the PCCS is functional. Approximately one-half of the WW volume is filled with a large volume of water that is called SP. The gas space in the WW acts as a receiver for noncondensable gases during a severe accident.

The SP plays an important role in passive safety performance because it provides: 1) A large heat sink, 2) Quenching of steam, which flows through the horizontal vents during LOCA, and 3) Scrubbing of fission products, which flow through the horizontal vents and the PCCS vents.

The WW is directly connected to the DW through twelve vertical/horizontal vent modules. Each module consists of a vertical flow steel pipe, with three horizontal vent pipes extending into the SP water. Each vent module is built into the vent wall, which separates the DW from the WW. The WW boundary is the annular region between the vent wall and the cylindrical containment wall and is bounded above by the DW diaphragm floor.

In the event of a pipe break within the DW, the increased pressure inside the DW forces a mixture of noncondensable gases, steam and water through either the PCCS or the vertical/horizontal vent pipes and into the SP where the steam is rapidly condensed. The noncondensable gases, which are transported with the steam and water, are contained in the free gas space volume of the WW. Performance of the pressure suppression in condensing steam has been demonstrated by a large number of tests.

There is sufficient water volume in the SP to provide adequate submergence over the top of the upper row of horizontal vents, as well as the PCCS return vent. When water level in the RPV reaches at one meter above the TAF, water is transferred from the pool to the RPV through the GDCS equalization lines. Water inventory of the SP and the GDCS pool is sufficient to flood the RPV to at least one meter above the TAF.

A passive residual heat removal system (see Figure XIII-3) is included in the design to remove heat from the reactor plant. The PHRS is designed in case of a station blackout including the loss of emergency power supply. The removal of residual heat should be provided without damage to the reactor core and the primary system boundary for an unlimited time. The PHRS consists of four independent trains, each of them being connected to the respective loop of the reactor plant via the secondary side of the steam generator. Each train has pipelines for steam supply and removal of condensate, valves, and an air-cooled heat exchanger outside the containment. The steam that is generated in the steam generators due to the heat released in the reactor; condenses and rejects its heat to the ambient air. The condensate is returned back to the steam generator. This occurs by natural circulation.

FIG. XIII-3. Principal diagram of PHRS.

XIX-4.5.1. Severe accidents

Compared to the standard PWR, safety is improved by the elimination of initiating events based on a specific design, as mentioned previously (optimum between safety, economy and human interface): large breaks on the primary circuit, reactivity insertion accident by rod ejection. However, the hypothetical case of a core meltdown is manageable in the following manner:

• Core meltdown: corium cooling should be ensured by reactor vessel pit flooding, because the core power density is small, and the large grace delay before an hypothetical core meltdown reduces the decay heat when the corium enters the lower plenum.

• Hydrogen risk: the Reactor Vessel Compartment atmosphere is inerted (cf. Figure XIX-5) to prevent hydrogen combustion like BWRs.

XIX-4.5.2. Management of the design extension condition

For SCOR, the transients in the design extension conditions are practically eliminated.

• H1 (total loss of the heat sink): SCOR concept is based on several independent decay heat removal loops (RRP) ready to operate in passive mode, having their heat sink either on pools with a limited availability of several hours, or on air cooling tower whose availability is almost infinite.

• H2 (total loss of feed water of the Steam Generator): the decay heat is removed by systems located on the primary circuit with a large redundancy (16 x 25%). There is no need of safety auxiliary feed water system.

• H3 (total loss of the power supplies): natural convection is possible on all the decay heat removal systems with the integrated heat exchangers, from the primary circuit to the heat sink.

• H4 (loss of the containment spray or loss of the low pressure safety injection): SCOR has no containment spray, because the containment is a pressure suppression containment type. The low pressure safety injection has a less significant role than in standard PWRs, because of the large primary inertia, the suppression of the large LOCAs and the strong effectiveness of decay heat removal systems.

• ATWT (Anticipated Transient Without Trip): SCOR has two independent shutdown systems. These transients will be treated individually as for the standard PWRs. The management of this complementary condition is eased owing to the always negative and higher moderator temperature coefficient than that of standard PWRs.

• Multi steam generator tubes rupture and non-isolable containment: the discharge of the Steam Generator is carried out in a dedicated pool.

• Failure of High Pressure Safety Injection: no HPSI is foreseen in SCOR.

XIX-5. Containment

The compactness of the primary circuit of SCOR leads to design of a pressure suppression containment similar to BWRs. The containment building in Figure XIX-5 consists of two physically separate compartments. The lower compartment is the reactor containment. The upper compartment is the mainly building again external hazard to protect the secondary side. The two compartments of the containment building are:

• The compartment of the primary side (or primary containment) is located under the reactor vessel-SG mating surface. It contains the vessel and all its primary pipe connections. Its volume is small. A pressure suppression device, as in BWR, controls the pressure. This compartment is inerted to manage the hydrogen risk.

• The compartment of the secondary side (or secondary containment) houses the steam generator. It is not inerted because it has no contact with the primary circuit when the vessel is closed.

XIX-6. Conclusions

SCOR is a simple compact PWR operating under low pressure. All the components are located inside the vessel, including the decay heat removal systems. The reactor has only one SG acting as the vessel head. The soluble boron free core operates with a low power density.

The configuration offers significant safety improvements over traditional loop-type PWRs (generation II) in achieving safety by design. The SCOR design eliminates the large LOCA. The decay heat removal system located very close to the core is very efficient. Calculations made with the CATHARE code have shown that, for the most penalizing accidents of the Design Basis Conditions, the core remains safely cooled with only four out of sixteen passive decay heat removal systems. For the most penalizing LOCA, a Low Pressure Safety Injection with small flow is required for a short time, one hour after the beginning of the transient.

The compactness of the reactor leads to the use of a pressure suppression containment type. It is inerted to prevent the hydrogen risk and, in case of a hypothetical core melting, the corium is cooled by pit flooding.