The emergency core cooling subsystem has been designed to provide long term residual heat removal in case of a primary leak accident concurrent with a station blackout. The hydroaccumulators with nitrogen under pressure will provide the coolant injection during the first stage of such an accident. The active subsystem is connected before the hydroaccumulators are emptied.

XII — 4. Description of HA-2

The emergency core cooling subsystem has been designed to possibly provide a long term residual heat removal in case of primary LOCA concurrent with a station blackout.

The passive core flooding subsystem includes four groups of hydroaccumulators under atmospheric pressure (2nd stage HA) which are coupled with the pipelines connecting the ECCS hydroaccumulators and the reactor. The hydroaccumulators of the passive core flooding system are connected to the primary system at 1.5 MPa. The hydrostatic pressure of the water column will flood the core removing the residual heat at the last stage of a LOCA for at least 24 hours.

Among the design basis conditions, the main accidental transients (blackout, steam line rupture, steam generator tube rupture, and LOCA) were studied with the CATHARE code. All calculations were performed with 4 out of 16 available RRP loops. These four RRPs are automatically actuated with the shutdown signal. In case of failure of the actuation, a control system will actuate other RRP loops.

Blackout

For this transient, the power is firstly removed by the SG and later by the RRP system. The RRP system reaches its full operation at about one thousand seconds. After one hour and half, the removed power by RRP is enough to cool the reactor.

Cold shock

In case of steam line rupture, the magnitude of cold shock is 22°C at core inlet. The number of the control rods is sufficient to stop the reactivity increase until the cold shut down. After the trip, the residual power is firstly removed by the released steam of the SG, and secondly it is stored in the large thermal inertia of the primary circuit. The RRP loops reach their full power in one thousand seconds. One hour after the beginning of the transient, the RRP remove all residual power. In this transient there’s no released water at the safety valve of the pressurizer.

In AHWR, natural circulation is used to remove heat from the reactor core under normal as well as shutdown conditions. Fig. III-2 shows the main heat transport (MHT) system and the passive decay heat removal system of AHWR. The two-phase steam water mixture generated in the core flows through the tail pipes to the steam drum, where steam gets separated from water. The separated water mixes with the subcooled feed water and flows down the downcomers to the reactor inlet header. From the header it flows back to the core through inlet feeders.

REACTOR BUILDING

GRAVITY DFHVEM

PASSIVE CWTAINWENT

CQOUNE SYSTEM

COOLING

WATER SEAL

TURBINE BUILDING

FIG. 111-1. Simplified flow sheet of AHWR.

FIG. III-2. MHT and decay heat removal system.

Larger density differences between hot and cold legs are possible to be achieved in two-phase flow systems compared to single-phase natural circulation flow systems. The absence of pumps not only reduces operating cost, but also eliminates all postulated transients and accidents involving failure of pumps and pump power supply.

Steady state flow prevails in a natural circulation loop when the driving buoyancy force is balanced by the retarding frictional forces. However, the driving force in a natural circulation system is much lower compared to a forced circulation system. With a low driving force, measures are needed to reduce the frictional losses. The methods adopted to reduce frictional losses include, elimination of mechanical separators in the steam drum and the use of large diameter piping. The larger pipes increase the amount of coolant needed in the primary system.

Elimination of mechanical separators makes the system dependent on natural gravity separation at the surface in the steam drum, which may increase carryover and carryunder. Carryover is the fraction of the liquid entrained by the steam, whereas carryunder is the fraction of vapour that is carried by the liquid flowing into the downcomer. Excessive carryover can damage the turbine blades due to erosion, whereas carryunder can significantly reduce the driving buoyancy force and hence the natural circulation flow rate. The steam drum size is chosen to keep carryunder and carryover within acceptable limits.

A rational start-up procedure of the AHWR has been worked out for low pressure and temperature conditions. For this, after the MHT is filled with water to a desired level in the steam drum, the MHT system is pressurized to an initial desired pressure by using steam generated from an external boiler. Subsequently, the control rods are partially withdrawn and coolant heating up continues at about 2% full power. Core boiling will start only after the steam drum pressure reaches 70 bar and the coolant temperature attains 285oC. The reactor power is increased gradually with controlled subcooling at the inlet of the reactor core until full power is reached.

Emergency core cooling system (ECCS) is designed to remove the core heat by passive means in case of a postulated loss of coolant accident (LOCA). In the event of rupture in the primary coolant pressure boundary, the cooling is initially achieved by a large flow of cold water from high pressure accumulators. Later, cooling of the core is achieved for three days by low pressure injection of cold water from gravity driven water pool (GDWP) located near the top of the reactor building. Fig. Ill-1 shows the emergency core cooling system.

VIII — 1. Introduction

The General Electric (GE) Nuclear Energy has developed a boiling water reactor called the simplified boiling water reactor (SBWR). Major differences between the current boiling water reactors (BWR) and the SBWR are in the simplification of the coolant circulation system and the implementation of a passive emergency cooling system. There are no recirculation pumps to drive the coolant in the vessel of the SBWR. The emergency core cooling and containment cooling systems do not have active pump injected flows.

XV — 1. Introduction

IRIS is a pressurized water reactor that utilizes an integral reactor coolant system layout. The IRIS reactor vessel houses not only the nuclear fuel and control rods, but also all the major reactor coolant system components including pumps, steam generators, pressurizer, control rod drive mechanisms and neutron reflector. The IRIS integral vessel is larger than a traditional PWR pressure vessel, but the size of the IRIS containment is a fraction of the size of corresponding loop reactors, resulting in a significant reduction in the overall size of the reactor plant. IRIS has been primarily focused on achieving design with innovative safety characteristics. The first line of defence in IRIS is to eliminate event initiators that could potentially lead to core damage. In IRIS, this concept is implemented through the ‘safety-by-design’ ™ IRIS philosophy, which can be simply described as ‘design the plant in such a way as to eliminate accidents from occurring, rather than coping with their consequences.’

The core make-up tanks effectively replace the high pressure safety injection systems in conventional PWRs. Each CMT consists of a large volume stainless steel tank with an inlet line that connects one of the cold legs to the top of the CMT and an outlet line that connects the bottom of the CMT to the direct vessel injection (DVI) line. The DVI line is connected to the reactor vessel downcomer. Each CMT is filled with cold borated water. The CMT inlet valve is normally open and hence the CMT is normally at primary system pressure. The CMT outlet valve is normally closed, preventing natural circulation during normal operation. When the outlet valve is open, a natural circulation path is established. Cold borated water flows to the reactor vessel and hot primary fluid flows upward into the top of the CMT.

V-2.3. Automatic depressurization system (ADS)

The automatic depressurization system consists of four stages of valves that provide for the controlled reduction of primary system pressure. The first three stages consist of two trains of valves connected to the top of the pressurizer. The first stage opens on CMT liquid level. ADS stages two and three open shortly thereafter on timers. The ADS 1-3 valves discharge primary system steam into a sparger line that vents into the IRWST. The steam is condensed by direct contact with the highly subcooled water in the IRWST. The fourth stage of the ADS consists of two large valves attached to ADS lines on each hot leg. The ADS-4 valves open on low CMT liquid level and effectively bring primary side pressure down to containment conditions. The ADS-4 valves vent directly into the containment building.

In the event of a core melt accident, pools of water located outside of the reactor pressure vessel will flood the lower part of the reactor cavity outside the pressure vessel. The pools are located above the pressure suppression chamber and are approximately two-thirds full of water. They are physically separated from each other by four equipment compartments containing mechanical components, piping and ventilation equipment. Each pool houses an emergency condenser, a containment cooling condenser (above the water surface), a core flooding line connection, and the SRV discharge pipes with steam quenchers (see Fig. XI-3 to XI-5). In addition, a drywell flooding line leads to the bottom of the drywell for cooling the exterior of the reactor pressure vessel.

XI — 5. Conclusions

In the past, different investigations concerning the emergency cooling were done. Besides the R&D work of FRAMATOME an experimental testing program was conducted 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. During 1996-1998 the European Union supported the BWR R&D Cluster of innovative passive safety systems, where the FZ Julich, Germany, the GRS Cologne, Germany, the NRG Petten, The Netherlands, NRG Arnheim, The Netherlands, PSI, Switzerland, SIET Piacenca, Italy, STORK NUCON, The Netherlands and VTT Espoo, Finland were involved.

Two large test facilities (NOKO, FZ Julich for separate effect tests and PANDA, PSI for integral tests) delivered experimental data. In addition operational data from the Doodewaard NPP were used. Post­test calculations were performed using system codes: ATHLET, RELAP and TRAC; lumped parameter codes such as COCOSYS, RALOC and GOTHIC. The CFD-codes applied were CFX and PHOENIX.

During the single effect tests, the heat transfer capability of the emergency condensers and the containment cooling condensers were investigated thoroughly. Special attention was directed to the simulation of condensation in horizontal tubes and to 3D stratification phenomena in the surrounding pools.

The passive cooling systems of the PSRD shown in Fig. XVIII-3 consist of four natural circulation loops: the first one is to transfer decay heat from the core to the SG; the second one from the SG to the EDRS heat exchanger (HEX); the third one from the EDRS-HEX and the CWCS-HEX inside the CV; and the fourth one from the HEX inside the CV to that outside the CV in the CWCS where the heat is transferred to the atmosphere. The second system called EDRS starts the operation passively after the feedwater pump termination due to the safety signal generation upon an abnormal event occurrence. This passive initiation relies on the hydraulic force valve developed at the JAERI. The hydraulic force
valve has the pressure-imposing line connecting to the feedwater pump to control the valve condition. The valve remains closed only when the pressure in the pressure-imposing line is higher than that inside the valve, i. e. the SG secondary pressure for this application. The feedwater pump termination, therefore, creates the pressure condition that opens the valve. After the hydraulic valves are passively opened, the steam will flow up toward the EDRS-HEX from the main steam line, while the liquid initially existed in the EDRS-HEX will flow down to the feedwater line.

Since the water is filled in the containment, the liquid-phase natural circulation between the EDRS — HEX and CWCS-HEX starts immediately after the temperature difference is generated between the two HEXs. The CWCS is a heat pipe system utilizing a refrigerant. The operation of the CWCS does not require the valve operation: the valves in the CWCS, if any, are opened for the normal condition. In summary, the decay heat removal system operation of the PSRD relies only on the operation of the hydraulic valve having inherently fail-safe nature, which means the passive cooling system of the PSRD is very reliable.

The capability of these natural circulation systems has been analyzed by using the RELAP5 code, which confirmed the feasibility of the basic concepts of this system (ISHIDA et al., 2003-2). Although the natural circulation behaviour in the containment is difficult to be analyzed by the RELAP5, much more accurate analysis can be performed by using the computational fluid dynamics (CFD) code because the analysis target is the single-phase liquid natural circulation.

XVIII-4. Conclusions

The basic design for the PSRD has been completed. The feasibility of the concepts of the PSRD has been confirmed through the analyses with thermal-hydraulic codes developed for the current — generation LWRs. The basic nature of the load follow characteristics has been characterized with the RETRAN code. The analysis results indicated that the reactor can respond smoothly to a typical load change due to a rather large negative moderator density reactivity coefficient. The functions of the passive safety systems to transfer the decay heat to the atmosphere and maintain the RPV coolant inventory have been confirmed through the LOCA analyses with the RELAP5 code. The current research focuses on the optimization of the system and measures to site in the demand area such as the placement in a deep pit filled with sea-water in a harbor.

REFERENCES TO ANNEX XVIII

[1] ISHIDA, T., et al., Development of In-vessel Type Control Rod Drive Mechanism for Marine Reactor, J. Nucl. Sci. and Tech., vol.38, No.7 (2001) pp.557-570.

[2] ISHIDA, T., et al., Concept Of Passive Safe Small Reactor For Distributed Energy Supply System, Proc. of 11th International Conference on Nuclear Engineering, (2003) ICONE11-36470.

[3] ISHIDA, T., SAW ADA, K., YNOMOTO T., Performance of Safety System of Passive Safety Small Reactor for Distributed Energy Supply System, Proc. of GENES4/ANP2003 (2003).

[4] KUSUNOKI, T., et al., Design Of Advanced Integral-Type Marine Reactor MRX, Nuclear Engineering and Design, Vol. 201 (2000) pp.155-175.

[5] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs 2004, IAEA-TECDOC-1391 (2004) pp.755-769.

The PCCS is a passive system which removes the decay heat released to the containment and maintains the containment within its pressure limits for design basis accidents such as a LOCA. The schematic of the PCCS is shown in Fig. VI-5. The PCC heat exchangers receive a steam-gas mixture from the DW, condense the steam and return the condensate to the RPV via the GDCS pools. The noncondensable gas is vented to the WW gas space through a vent line submerged in the SP. The venting of the noncondensable gas is driven by the differential pressure between the DW and WW. The PCCS condenser, which is open to the containment, receives a steam-gas mixture supply directly from the DW. Therefore, the PCCS operation requires no sensing, control, logic or power actuated devices for operation.

The PCCS consists of six PCCS condensers. Each PCCS condenser is made of two identical modules and each entire PCCS condenser two-module assembly is designed for 11 MWt capacity. The condenser condenses steam on the tube side and transfers heat to the water in the IC/PCC pool. The evaporated steam in the IC/PCC pool is vented to the atmosphere. PCCS condensers are located in the large open IC/PCC pool, which are designed to allow full use of the collective water inventory, independent of the operational status of any given PCCS loop.

To substantially limit the release of fission products beyond the containment; a permanent under pressure is maintained in the inter-containment gap of the V-392 design. This safety function, one of the most important, is fulfilled by two systems: (1) an exhaust ventilation system equipped with a filtering plant with suction from the inter-containment gap and outlet into the stack; (2) a passive system of suction from the inter-containment gap. The first system is intended to control removal of a steam-gas mixture from the inter-containment gap under accidents with the total loss of power. The system is capable to remove at least 240 kg per hour that is equivalent to the inner containment leaks of 1.5% containment volume per 24 hours. The second system consists of the lines connecting the inter-containment gap with PHRS exhaust ducts, which are always in the hot state. This enables permanent removal and purification of inner containment leaks regardless of power supply and operator actions. According to estimations, the under pressure is maintained at any point of the inter­containment gap with inner containment leaks up to 2.8% of containment volume per day (the design basis for the containment is 0.3%). The technical solution described above in combination with the systems for the containment pressure decrease (traditional spray system and new passive heat removal system) allows us to give up the filtered venting system designed for V-392. Even though this system satisfies the current requirements to filtered venting; this should not increase the risk of containment failure. Filtered venting is not required within the short term of a core melt accident.