IV — 1. Introduction

The AP600 and AP1000 are pressurized light water reactors designed by the Westinghouse Electric Corporation to produce 600 MW and 1100 MW of electric power, respectively. Both designs employ passive safety systems that rely on gravity, compressed gas, natural circulation, and evaporation to provide for long term cooling in the event of an accident.

Figure V-1 shows the overall layout of the plant. Figure V-2 is a schematic that illustrates the primary system components. The primary loop consists of the reactor vessel, which contains the nuclear fuel assemblies; two hot legs, which connect the reactor vessel to the steam generators; two steam generators; a pressurizer; four canned motor pumps; and four cold legs.

FIG. V-2. Schematic ofprimary loop of the AP600 and AP1000 plants.

During normal power operation, heat is generated in the reactor fuel. This heat is transported by conduction through the fuel and its cladding and transferred by convection to the water. Since the entire system operates at 15.5 MPa (2250 psia), bulk boiling of the water does not occur. The heated water is transported through the hot legs to the U tubes inside the steam generators. The energy of the primary coolant inside the tubes is transferred to the water on the secondary side by forced convection inside the tubes, conduction through the tube walls and boiling on the outside surface of the U tubes. The cooled water leaving the steam generator is pumped by four canned motor pumps, through four cold legs, back into the reactor vessel where the heating cycle is repeated. Primary system pressure is maintained constant by the pressurizer.

The fundamentally new concept for accident control incorporated into the SWR 1000 includes equipment which, in the event of failure of the active safety equipment, will bring the plant to a safe condition without the need for any instrumentation and control signals or external power. This passive safety equipment (Figures XI-1 and XI-2) includes the following:

Emergency condensers

The function of the emergency condenser system is to remove, in the event of an accident, the decay heat still being generated in the core as well as any sensible heat stored in the reactor pressure vessel
to the core flooding pools, without any coolant inventory being lost from the reactor pressure vessel. The system thus replaces the high pressure coolant injection systems used in existing BWR plants. The emergency condenser system also provides a means for reactor pressure relief that is diverse with respect to the safety-relief valves.

Containment cooling condensers

The function of the four containment cooling condensers is to remove — by entirely passive means — decay heat from the containment following accidents that result in the release of steam into the drywell, and in this way to limit the increase of containment pressure. They provide redundancy and diversity with respect to the residual heat removal system.

Core flooding system

The core flooding system is a passive low pressure flooding system for controlling the effects of loss — of-coolant accidents. It is installed at an elevation which ensures that, following automatic depressurization of the reactor, it can passively flood the reactor core by means of gravity flow. The system provides redundancy and diversity with respect to the core flooding function of the residual heat removal system.

Drywell flooding system

A postulated severe accident involving core melt is controlled such that the molten core is retained inside the reactor pressure vessel. For this purpose the section of the drywell surrounding the reactor pressure vessel is flooded with water in order to cool the exterior of the reactor pressure vessel and thus remove heat from the reactor.

Passive pressure pulse transmitters

The passive pressure pulse transmitter is a completely passive switching device which is used to directly initiate the following safety functions (as a minimum), without the need for instrumentation and control equipment: reactor scram, containment isolation at the main steam line penetrations, and automatic depressurization of the reactor pressure vessel. The passive pressure pulse transmitter comes into action as a result of a drop or increase in reactor water level as well as an increase in reactor pressure. For activating the various safety functions, passive pressure pulse transmitters of redundant design are installed at two elevations. The upper passive pressure pulse transmitters, situated at an elevation beneath that of the normal water level of the reactor pressure vessel, are responsible for initiating reactor scram. The lower passive pressure pulse transmitters, arranged at a lower elevation, activate automatic depressurization of the reactor as well as closure of the main steam containment isolation valves. Further passive pressure pulse transmitters installed at appropriate locations respond to a rise in reactor water level above the main steam nozzles and likewise activate containment isolation at the main steam line penetrations.

This section briefly describes the evolution of a small break loss of coolant accident in MASLWR. Because MASLWR is an integrated reactor system, there are very few plausible primary break scenarios. In the event of a small break, the MASLWR passive safety systems would respond to the accident. The passive safety systems consist of the following components:

• Two, independent, small diameter, steam vent valves (SVV)

• Two, independent, small diameter, automatic depressurization system (ADS) valves

• Two, independent, small diameter, sump recirculation valves (SV)

• A high pressure containment vessel with an internal pressure suppression pool, and

• An external cooling pool that serves as the ultimate heat sink for the high pressure containment and reactor decay heat.

Let us postulate the inadvertent opening of an ADS valve. Figure XVII-3 provides a schematic of the postulated pressure trend for illustration purposes. As shown in the figure, the transient begins with a relatively short blowdown period that consists of a subcooled blowdown into the suppression pool within the stainless steel containment. The suppression pool consists of the annular space bounded by the exterior surface of the reactor vessel and inner surface of the containment walls. It is partially filled with water. This water region is integral to the long term removal of decay heat following system depressurization (blowdown). The rapid rise in containment pressure results in a Safety Injection signal, which automatically opens the steam vent valves, the ADS valves, and the sump recirculation valves. The ADS blowdown period serves to further reduce the reactor vessel pressure well below the saturation pressure corresponding to the hot leg temperature. A major advantage to the small volume, high pressure containment is that the blowdown quickly results in equalizing the containment and reactor vessel pressures, effectively terminating the blowdown. As the pressures become equalized, a natural circulation flow path is established in which the sump fluid enters through the sump recirculation valves, descends through the downcomer region in the lower portion of the reactor vessel, rises through the core and riser, and finally exits through the upper vent valve into the containment as a saturated vapor. From the vent valve, the fluid returns to the sump via condensation on the containment walls and/or water surface, thus completing its circuit.

Primary Side——————

Containment—————-

FIG. XVII-3. Illustration of transient phases for a MASLWR SBLOCA.

Finally, to ensure the long term removal of heat from the containment and thus to moderate the containment pressure, the containment itself is submerged in an outer pool of water which is open to the atmosphere. Within this pool, thermal energy is transferred from the outer containment wall to the atmosphere via natural convection and circulation of the water. The pool is formed in the space between the outer containment wall and the inner wall of the concrete structure in which the containment is placed.

In conclusion, a SBLOCA in MASLWR can be divided into three phase, or modes of operation, a blowdown phase, an ADS operation phase and a long term cooling phase. A more detailed description of the MASLWR design is provided in IAEA-TECDOC-1536.

The GDCS provides emergency core cooling after events that threaten the reactor coolant inventory. Following confirmed RPV water Level 1 signal, the ADS depressurizes the RPV to allow the GDCS injection. Once the reactor is depressurized, the GDCS is capable of injecting large volumes of water into the depressurized RPV to keep the core covered for at least 72 hours following loss of coolant accident (LOCA).

The GDCS requires no external AC electrical power source or operator intervention. The cooling water flows from the GDCS pool to the RPV through simple and passive hydrostatic head. A schematic of the GDCS injection is shown in Figure VI-3. The actual GDCS flow delivered to the RPV is a function of the differential pressure between the reactor and the GDCS injection nozzles, as well as the loss of head due to inventory drained from the GDCS pool. As shown in Figure VI-3, the GDCS can be considered as two separate systems: a short term safety system and long term safety system.

The short term safety system is designed to provide short term water makeup to the reactor vessel for maintaining water level higher than the top of active fuel (TAF). There are four identical GDCS drain lines. Each GDCS drain line consists of a pipe exiting from the GDCS pool and squib valves. For short term cooling requirement, each line takes suction from three independent GDCS pools positioned in the upper elevations of the containment. In the ESBWR, there are two GDCS tanks with 602 m3 and one GDCS tank with 795 m3. Flow through each drain line is controlled by squib valves, which remain open after initial actuation. Each short term GDCS drain line connects to two GDCS injection nozzles on the RPV.

In case of a pipe break of GDCS lines at lower RPV elevations, large amounts of GDCS water can be lost and water inventory can be reduced close to the TAF. To compensate aforementioned loss of coolant in the GDCS lines, the long term cooling water is provided by a second GDCS subsystem fed by the SP. This is called GDCS equalization system. This GDCS equalization lines are opened after a prescribed time delay so that the short term GDCS pools have time to drain to the RPV and the signal of the initial RPV water inventory reduction as a result of the blowdown does not make the equalizing lines open. For long term event, the GDCS equalization lines are open when the RPV coolant level decreases to 1 m above the TAF. The squib valves are actuated in each of four GDCS equalizing lines, which connect to one RPV injection nozzle per line. The open equalizing lines leading from the SP to the RPV make long term coolant makeup possible. As shown in Fig. VI-3, GDCS equalization line nozzles are placed at a lower elevation on the RPV than those of the short term GDCS, so the GDCS equalization lines make it possible to prevent the core uncovery even though the short term GDCS injection fails. This long term GDCS equalization system also functions through purely passive hydrostatic head differences.

The GDCS pools are placed above the RPV with their air space connected to the DW. This connection effectively increases the DW air space and provides a larger volume for the released gases produced during a severe accident. After the GDCS pools are drained, the total volume of the GDCS pools are added to the volume of the DW air space.

XII — 1. Introduction

The design of WWER-1000/392 (V-392) was developed by FSUE ‘Atomenergoproject’ (Moscow, Russian Federation), FSUE EDO ‘Gidropress’ (Podolsk, Russian Federation) and the Russian National Research Centre ‘Kurchatov Institute’ (Moscow).

The primary purpose of the V-392 is to ensure the safety of the personnel, the public, and the environment against radiation effects exceeding the specified (prescribed) radiation doses. This principle also addresses the standards for releases of radioactive substances and their content in the environment under normal operation conditions, anticipated operational occurrences, design, and beyond-design-basis accidents during the plants life. The objective of the reactor plant design and nuclear plant process systems is to achieve estimated probability of severe core damage not above 1.0E-5 per reactor-year and the probability of accidental radioactive releases not above 1.0E-7 per reactor-year. These values are specified in Russian safety standards.

The design of NPP with WWER-1000/392 improves technical and economic parameters. Wide application of passive safety means, using natural physical processes, along with the traditional active systems is a specific feature of this design. Each plant designer must solve problems caused by implementing passive safety systems. The passive systems have their own advantages and drawbacks in comparison with the active systems both in the area of plant safety and economics. Therefore, a reasonable balance of active systems and new passive means is adopted in V-392 design to improve safety and public acceptability of nuclear energy.

The passive systems of WWER-1000/392 [1,2] are:

• Passive quick boron supply system,

• Passive subsystem for reactor flooding HA-1 (hydroaccumulators of first stage),

• Passive subsystem for reactor flooding HA-2 (hydroaccumulators of second stage),

• Passive system to maintain low inter-containment gap (annulus) atmosphere pressure,

• Passive residual heat removal system via steam generator (PHRS),

• Passive core catcher.

The reactor main coolant pump flywheel inertia provides the initial boron injection to the primary
system. Nitrogen gas at high pressure injects borated water from the hyrdoaccumulators in two stages

to the core. Gravity provides the driving force in some of the passive safety systems. The last three systems operate in a natural circulation loop providing decay heat removal from the core.

The overall configuration of the safety systems based on passive principles mentoned above is shown in Figure XTTT-1.

The overall functions of the safety systems based on passive principles mentioned above in comparison with active safety systems are shown in Figure XTTT-2.

The maximum removed power by each RRP loop is about 5 to 7 MW according to the operating conditions. This small removed power, whatever the reactor power, makes it possible to test the heat removal system, with the reactor in operation, without significantly disturbing operating conditions. This procedure of test constitutes a significant element with respect to the reliability of these systems.

The RRPp are safety grade chilled water loop and pumps excepted.

VII — 1. Introduction

The Reduced-Moderation Water Reactor (RMWR) is a light-water cooled high-conversion reactor that is being developed by the Japan Atomic Energy Research Institute (JAERI) with collaboration from the Japanese industries. The design study conducted so far has indicated that the RMWR can realize the favourable characteristics of high conversion ratio of more than one, high burn-up, long operation cycle, and multiple recycling of plutonium. The design is characterized by the use of the ‘double flat core’ which consists of two flat core parts and three blanket parts in the vertical direction. This geometry is adopted to increase the neutron leakage from the core to make the void reactivity coefficient negative. The fuel assembly consists of the MOX fuel rods tightly arranged in the triangular lattice with the gap width of typically 1.3 mm to increase the fuel-to-coolant volume ratio.

Although several types of the RMWR systems have been investigated, the current study focuses mainly on the BWR type due to advantages regarding the core performances and the system simplification. Among the BWR type RMWRs, two system designs have been developed for different core powers. The larger one is the 1300 MW(e) RMWR based on the ABWR design (IAEA — TECDOC-1391). Table VIII-l summarizes the major characteristics of this system, which has 900 fuel assemblies, each of which consists of 217 fuel rods with the outer diameter of 13.7 mm arranged in the triangular lattice in gap width of 1.3 mm (see Figures VIII-1 and VIII-2). The double flat core consists of the lower and upper core parts with the height of 205 and 195 mm, and the lower, medium, and upper blanket parts with the height of 190, 295, and 220 mm as shown in Figure VIII-3. The fissile plutonium-enrichment of the MOX fuel is 18% for the reload fuel at the equilibrium core. The blanket material is depleted UO2.

900 fuel assemblies & 283 control rods

228 mm

Number of fuel rods 217

FIG. VIII-1. Core configuration for the 1300 MW(e) RMWR.

7,600 m m

FIG. VIII-3. Schematic of axial core
configuration for the 1300 MW(e) RMWR.

The smaller one is the 330 MW(e) RMWR system designed to seek benefits of the low initial capital cost and the flexibility in the plant installation corresponding to the power demand (IWAMURA, T. et al., 2002). To overcome the economic disadvantage for small reactors, the system was simplified by adopting natural circulation core cooling and passive safety features. Since the RMWR utilizes the flat and short length core and operates under lower core flow conditions, the steam velocity in the chimney region is smaller than for the conventional BWR. This characteristic allows further simplification: the steam/water separator and the steam dryer may be eliminated since the gravitational steam/water separation is expected to be possible at the free surface. The major characteristics of the reactor are listed in Table VIII-2. A breeding ratio of 1.01, negative void coefficient and natural circulation cooling of the core were realized under the discharged burn-up of 60GWd/t. The core consists of 282 of hexagonal fuel bundles: each has 217 fuel rods arranged in triangular lattice of 1.3 mm in gap width as listed in Table VIII-2. The core height and outer diameter are 1300 and 4140 mm, respectively, as shown in Figure VIII-4. Figure VIII-5 summarizes the plant system concept. The passive safety features include the accumulator injection system, the isolation condenser, and the passive containment cooling system.

VII — 2. Description of natural circulation core cooling for the 330 MW(e) RMWR

The natural circulation loop consists of the core, the divided chimney, the downcomer, and the lower plenum, which is similar to the other natural circulation cooling BWR concepts such as the SBWR. There are, however, several differences in the characteristics comparing to the SBWR, which includes:

1) One-order smaller absolute value for the void reactivity coefficient, which makes the neutronics — thermal-hydraulic coupling to be almost negligible,

2) Smaller frictional pressure drop across the core because of the shorter core length and the lower core flow rate despite the smaller core flow area

3) Hexagonal cross-sectional shape of the divided chimney

4) Elimination of the conventional separator.

The first characteristic makes the evaluation of the instability problem much simpler. Thus, the characteristic can be regarded as a benefit in view of the thermal-hydraulic analysis. The second one is also considered as a benefit for the realization of the natural circulation system. The third one causes the prediction of the void fraction in the chimney to be difficult because the previous studies are almost nonexistent for this geometry. However, the thermal-hydraulic relationships among the pressure, vapor and liquid flow rates, and void fractions are supposed to be basically the same as those for the circular geometry. So this characteristic is not an essential problem but just requires confirmation tests using the actual geometry of the system to get the relationships for the design finalization. Since the free-surface separation was utilized for the old type natural circulation BWR, the fourth characteristic does not create a new problem. Careful consideation, however, is necessary because the conditions are not completely the same between the old BWRs and the RMWR. The mockup test will be required to confirm especially the phase separation characteristics on the free surface.

Figure XV-1 shows a cross sectional view of IMR reactor. As shown in the figure, IMR employs several key concepts for the reactor design. The first one is the integrated primary system concept. Fuel assemblies, control rods, steam generators (SGs), and control rod drive mechanisms (CRDMs) are all installed inside the reactor vessel (RV), and there is no main coolant piping and no large penetration of RV.

The second one is the hybrid heat transport system (HHTS), which is a kind of two-phase natural circulation system operating under the self-pressurized saturation condition of the primary coolant. The coolant starts boiling at the upper part of the core, two-phase coolant flows up in the riser, and is condensed and sub-cooled by SGs. In order to control the amount of boiling and the system pressure, IMR has two kinds of SGs, i. e. SG in liquid region (SGL) and SG in vapor plenum (SGV). Table XV — 1 shows the major specifications and operating condition of the IMR primary system.

6m

IMR(350MW(e))

FIG. XV-1. Configuration of the IMR primary systems.

This concept allows eliminating a pressurizer and reactor coolant pumps, which simplify the reactor design and is beneficial to reduce maintenance work. This concept also increases the driving force of the coolant flow. The average void fraction in the riser (20%) is optimized to minimize the required RV height while keeping the appropriate thermal margin of the core. A small amount of boiling is allowed in the core but the core characteristics are still very similar to that of conventional PWRs, because the void fraction is relatively low and most of the core operates under sub-cooled condition. From the safety point of view, this concept realizes a reactor design free from large-scale break of the primary boundary (i. e. LOCA), control rod ejection accident, and loss of flow accident. As the results, no safety injection system and containment spray system is required for IMR. Additionally, such downsizing and safety feature allows applying very small dry containment, which greatly reduces construction cost.

The feasibility of HHTS is one of the most important subjects for IMR because of less knowledge of two-phase flow behavior under such high temperature (345deg-C) and pressure (15.5MPa). In order to verify the feasibility, two series of experiments have been performed. The first one is an air-water scale experiment and the second one is a natural circulation experiment under the actual temperature, pressure, and axial dimension.

With respect to thermal hydraulic phenomena, normal full-power operation is typical of most pressurized water reactor (PWR) systems. Under shutdown cooling conditions, a key feature of the AP600 and AP1000 designs is that it uses core decay heat to drive the core cooling process by natural circulation. In fact, the AP600/AP1000 designs use core decay heat to drive the following six natural circulation processes:

• Primary system natural circulation (2 x 4 loop)

• PRHR loop circulation (1 loop)

• CMT loop circulation (2 loops)

• Lower containment sump recirculation (2 loops)

• Containment internal circulation (steam)

• Containment external circulation (air)

Figure V-3 presents a schematic that describes the connections of the primary system passive safety systems.

AP600 Passive Core С00І11Ш System

DtirffttuflXKtan

Vulvtt

Loop

_ ОПР VtlTMfTt

FIG. V-3. passive safety systems used in the AP600/AP1000 designs.

The AP600/AP1000 passive safety systems consist of:

• A passive residual heat removal (PRHR) system

• Two core make-up tanks (CMTs)

• A four stage automatic depressurization system (ADS)

• Two accumulator tanks (ACC)

• An in-containment refueling water storage tank (IRWST)

• A lower containment sump (CS)

• Passive containment cooling system (PCS)

The SWR-1000 is equipped with 4 emergency condensers having a maximum power of about 60 MW at a pressure of 7 MPa. The emergency condensers are horizontal tubes arranged between two collectors (see Fig. XI-3). Each bundle consists of 104 tubes arranged in four vertical levels. The tubes have a length of 10 m, an inner diameter of 38.7 mm, a wall thickness of 2.9 mm and are made of austenitic steel (design state 1998). The upper collector is connected via the feed line and the lower connector via the back line connected to the reactor pressure vessel. Both lines can not be blocked. The back line is equipped with an anti circulation loop to avoid the establishment of a stratified counter current flow during the normal operation.

Under normal operation conditions the water level in the core is above the level in the emergency condenser bundles. The water inside the tubes is entirely single phase and no heat transfer to the flooding pool takes place (Fig. XI-3 left side).

During an accident the water level in the core sinks below the bundle level of the condenser. The condenser tubes are filled with steam and the emergency condenser acts as a heat sink (Fig. XI-3 right side). In addition, owing to the pools’ elevation, the water in the core flooding pools is used for passive flooding of the reactor core following reactor pressure vessel depressurization in the event of a loss of coolant accident. For this function, spring-assisted check valves open the core flooding lines automatically. Passive core flooding serves as a diverse means of providing reactor pressure vessel coolant makeup which supplements the active core cooling systems.