After the GDCS injection period is terminated, the core starts to boil again due to the long term decay heat. The long term DW pressure increase is mitigated through PCCS condensers. The PCCS removes the core decay heat energy, released to the containment from the RPV, to outside of the containment. Steam condensate from the PCCS is returned to the RPV via GDCS tanks and noncondensable gases are vented to the SP.

The water inventory of the GDCS pools is slowly replenished with condensate draining from the PCCS. For all design basis events, the closed loop of PCCS condensation and GDCS drainage to the RPV results in long term coverage of the core. Beyond design basis events, the GDCS equalization flow may be necessary where multiple failures are assumed. When the vessel water level reaches 1 m above TAF and at least 30 minutes has passed since the Level 1 signal is confirmed, the GDCS equalization line will open to inject water from the SP to the RPV.

VI-4. Conclusion

The ESBWR program is based on the earlier SBWR program, which was sponsored by the US Department of Energy (DOE). The ESBWR program was started in 1993 to improve the economics of the SBWR. There are some significant differences between the SBWR and ESBWR designs. The major differences are: 1) increased core thermal power, 2) much higher core power density, 3) considerably reduced DW and WW volumes relative to the reactor power, 4) increased GDCS tank volume, and 5) increased component numbers of the PCCS and ICS in the ESBWR compared with those in the SBWR. In addition to these, the number of the main steam lines was increased to four in the ESBWR. A multi-year, four-phase program was defined to complete the technology, develop a detailed design, and secure certification with regulatory bodies. Evaluation of the overall design showed that the plant was considerably simplified and that the overall material quantities were significantly lower than those for the SBWR design and other GE designs.

The ESBWR is equipped with a passive safety system that is basically similar to that of the SBWR. This reactor design features the simplification of the coolant circulation system and implementation of passive safety system. There are several engineered safety systems and safety-grade system in the ESBWR which are directly related to the relevant issues and objectives of the present program. The performance of these safety systems under a LOCA and other important transients is a major concern. Since the ECCS is driven by the gravitational head, interactions between the ADS, GDCS, PCCS and other auxiliary systems are important. The safety systems and various natural circulation phenomena encountered after initial blowdown in the ESBWR are somewhat different from the system and phenomena studied by the nuclear community in the existing commercial nuclear reactors.

Comprehensive integral system and separate effects testing have been conducted to verify the functionality of passive safety system [1]. GE performed tests to assess the GDCS performance in a low pressure full-height GIST facility. Results of this study demonstrated the feasibility of the GDCS concepts. GE also performed tests to assess the PCCS performance in a low pressure, full-height Toshiba GIRAFFE facility in Japan. A PANDA facility in Switzerland, with a low pressure and full — height, was built for testing the PCCS performance and containment phenomena in the SBWR. Later, PANDA was modified to partially simulate the ESBWR configuration.

Purdue University designed and constructed an integral test facility, called PUMA (Purdue University Multi-dimensional integral test Assembly), sponsored by the U. S. NRC. Originally, the PUMA facility was designed to address the functionality of SBWR safety system in 1994. The facility was modified to simulate the safety system in the ESBWR in 2006. The facility contains all of the important safety systems of the ESBWR that are pertinent to the postulated LOCA transient.

The design and technology program for the ESBWR involves several utilities, design organizations and research groups. In mid 2002, the technology base of the ESBWR was submitted to the U. S. NRC for review with the objective of obtaining closure of all technology issues. This was a first and necessary step toward obtaining NRC design certification.

REFERENCES TO ANNEX VI

[1] ISHII, M., et al., Second Scaling and Scientific Design Study for GE ESBWR Relative to PUMA Facility with Volume Ratio of 1/475, Purdue University Report PU-NE-04-04 (2004).

[2] ISHII, M., et al., Scientific Design of Purdue University Multi-Dimensional Integral Test Assembly (PUMA) for GE SBWR,” Purdue University Report PU-NE-94-01, U. S. Nuclear Regulatory Commission Report NUREG/CR-6309 (1996).

[3] GAMBLE, R., ESBWR Technology Program: Test Program, NRC-GE Meeting, Rockville, Maryland, USA (2002).

CAREM safety systems are based on passive features that don’t require active actions to mitigate accidents for a long period. They are duplicated to fulfil the redundancy criteria. One of them that relies on natural circulation is the residual heat removal system (RHRS), Figure XIV-2. It has been designed to mitigate a Loss of Heat Sink, reducing the pressure of the primary system to values lower than the values at hot shutdown, by removing the decay heat. The RHRS is a simple and reliable system that operates condensing steam from the reactor dome in the emergency condensers. This establishes a stratified two-phase natural circulation loop with the primary system.

FIG. XIV-2. Safety system.

The emergency condensers are heat exchangers consisting of an arrangement of parallel horizontal U tubes between two common headers. The top header is connected to the reactor vessel steam dome, while the lower header is connected to the reactor vessel at a position below the reactor water level. The condensers are located in a pool filled with cold water inside the containment building. The inlet valves in the steam line are always open, while the outlet valves are normally closed, therefore the tube bundles are filled with condensate. When the system is triggered, the outlet valves open automatically. The water drains from the tubes and steam from the primary system enters the tube bundles and condenses on the cold surface of the tubes. The condensate is returned to the reactor vessel forming a natural circulation circuit. In this way, heat is removed from the reactor coolant. During the condensation process the heat is transferred to the pool water by boiling process. This evaporated water is then condensed in the suppression pool of the containment. The pool of the RHRS has a volume sufficient to provide an autonomy greater than the grace period for the prototype (48 h).

The RHRS main characteristics are listed in Table XIV-1.

TABLE XIV-1. RESIDUAL HEAT REMOVAL SYSTEM-EMERGENCY CONDENSER FOR CAREM PROTOTYPE

In case of a very small LOCA (lower than 3/4’) the RHRS is also demanded by the reactor protection system to depressurize the primary system to allow the emergency injection system to act.

XI — 4. Conclusions

In order to assess the CAREM primary circuit numerical modelling a High Pressure Natural Circulation Rig (CAPCN) was built in the decade of the 90s. The CAPCN resembles CAREM in the primary loop and steam generators, while the secondary loop is designed just to produce adequate boundary conditions for the heat exchanger. The CAPCN rig reproduces most of the dynamics phenomena of the RCCS, except for 3-D effects. Several tests were performed covering thermal hydraulics, reactor control and operating techniques around the nominal operational point.

Indicative values of the main variables corresponding to the CAPCN operation at full power are shown in Table XIV-2.

TABLE XIV-2. CAPCN NOMINAL CONDITIONS

Most of the tests performed consisted of an initial self-steady state in which a pulse-wise perturbation induced a transient. In this case the perturbation is a thermal unbalance as severe as possible for operational transients. The following test groups were performed:

• Preliminary tests to characterize components and equipment;

• Thermal balance test, instrumentation calibration and evaluation of their accuracy;

• Dynamic test around the operational nominal point without control: thermo-hydraulic response evaluation;

• Dynamic test with the control loops (power pressure);

• Dynamic test at low pressure and temperature.

Passive core cooling system using a Steam Generator (SG) is implemented in case of a loss of reactor coolant accident (LOCA) to cool-down and depressurize the primary system using SGs. Figure IV-2 shows a scenario for small break LOCA.

To show the validity of this system, it is important to verify natural circulation flow characteristics of the primary side and heat transfer capability through SGs because supplied water from accumulators includes non-condensable gas (nitrogen). From this point of view, a simulation test had been performed in addition to the simple flow test in a single SG tube and numerical analyses using CANAC3-3D code. This test was performed by Kansai Electric Power, Kyushu Electric Power, The Japan Atomic Power, and Mitsubishi Heavy Industries.

III — 3. Conclusions

In the safety concept of APWR+ against small LOCA event, Steam Generators (SGs) are used for decay heat removal. Therefore, it is important to maintain natural circulation in the primary system during small LOCA events to transfer the decay heat from the core to the SG. Non-condensable gas, which dissolves in water from the accumulators and injected to primary loop on LOCA event, may accumulate in the top of U-bend tube and cause siphon brake for natural circulation at the U-bend region of SG.

Small LOCA test was performed using the ‘EOS’ simulation loop of a PWR to verify natural circulation and siphon break in SGs, and we concluded that natural circulation with non-condensable gas was maintained during LOCA event and heat removal by the SGs was continuously effective.

REFERENCES TO ANNEX IV

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs, p139-157, IAEA-TECDOC-1391, IAEA, Vienna (2004).

[2] SUZUTA, T., et al., Development of Advanced Computer Code CANAC3-3D for Next Generation PWR (Part2), Proc. of the 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety(NTHAS2), (2000).

[3] MAKIHARA, Y., et al., Development of Next Generation PWR (APWR+), Proc. of ICONE-9, Nice, France (2001).

[4] ARITA, S., et al., Safety Evaluation of Next Generation PWR (APWR+), Proc. of ICONE-10, Arlington, Virginia, USA (2002).

[5] TANAKA, T., et al., Examination of Natural Circulation and Heat Removal by Steam Generator, Proc. of the 6th International Conference on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-6),#N6P054, Nara, Japan (2004).

Three main steam lines connecting the reactor pressure vessel to the high pressure turbine section serve to transport the steam generated in the reactor to the turbine. Each main steam line inside the containment is allocated a specific number of safety-relief valves for overpressure protection of the reactor pressure vessel. For automatic depressurization, the safety-relief valves are opened either by solenoid pilot valves or by the passive pressure pulse transmitters and diaphragm pilot valves.

The core of the SWR 1000 was reduced in active height and the fuel assembly was enlarged. A consequence of reducing the active core height is that the core can be positioned lower down inside the reactor pressure vessel. This provides a larger water inventory inside the reactor pressure vessel above the core, a feature which facilitates accident control. The fuel assemblies were enlarged resulting in fewer fuel assemblies in the core. This reduces handling times during refueling. Reducing the number of fuel assemblies also reduces the number of control rods, and hence the number of control rod drives as well. The average power density is approximately 51 kW/L.

The reactor pressure vessel has numerous nozzles for connecting the piping of the main steam, feedwater, emergency condenser, shutdown cooling and vessel head spray systems, as well as for accommodating the internal reactor water recirculation pumps, the control rod drives, the core flux monitoring assemblies and the reactor water level, pressure and temperature measuring instrumentation. The four emergency condensers as well as the four standpipes (which connect the passive pressure pulse transmitters and the condensation pots of the reactor pressure vessel level measuring equipment to the reactor vessel) are regarded as being external extensions to the vessel since they are connected to it via non-isolatable lines.

The main auxiliary systems of the SWR 1000 are:

• the residual heat removal (RHR) system;

• the reactor water cleanup system and fuel pool cleanup system;

• the fuel pool cooling system.

Apart from these systems many other auxiliary systems such as waste processing systems, a chilled water system, and HVAC systems exist for normal operation of the plant. The SWR 1000 design concept includes two active systems for low pressure core injection/flooding and residual heat removal. As in earlier plant designs these systems perform the following tasks:

• Cooling of the reactor core during and after normal plant shutdown;

• Water transfer operations before and after refueling;

• Operational heat removal from the core flooding pools and the pressure suppression pool;

• Heat removal from the containment in the event of loss of the main heat sink by cooling the pressure suppression pool and core flooding pool water;

• Low pressure coolant injection into the reactor pressure vessel and simultaneous heat removal in the event of loss-of-coolant accidents.

FIG. XI-1. SWR 1000 — Active and passive safety systems.

FIG. XI-2. SWR-1000 safety concept.

Because MASLWR uses natural circulation for primary loop flow, reactor coolant pumps are not needed. In this regard, its primary flow loop is quite simple as shown in Figure XVII-2. The long vertical tube in the centre of the reactor vessel is called the riser and functions like a chimney to enhance the driving head of the natural circulation flow. Starting from the bottom of the riser, fluid enters the core, which is located in a shroud connected to the riser entrance. While the fluid travels through the core, it is heated and rises by buoyancy through the riser. Hot fluid in the surrounding annulus, outside the riser is cooled by convective heat transfer to a helical coil steam generator. The fluid inside the tubes is at a lower pressure, hence boiling occurs inside the tubes to generate superheated steam. The steam produced within the tube side of this coil travels on to the turbine generator set where it is used to produce electrical power. The cooled primary fluid in the annulus is negatively buoyant and descends to the bottom of the vessel and the inlet of the core thereby completing its loop.

Table XVII-1 lists the steady-state operating conditions for MASLWR. The design provides a 53oK temperature rise from the core inlet to the core outlet. In addition, it is designed to provide superheated steam at the helical coil outlet to eliminate the need for separators and driers. The secondary side pressure was selected so that off-the-shelf low pressure steam turbines could be implemented.

TABLE XVII-1. MASLWR STEADY-STATE OPERATING CONDITIONS

The basic nature of the passive safety systems and accident management strategies for the SBWR and ESBWR are similar. The major engineered safety systems and safety grade systems in the ESBWR are:

FIG. VI-1. Schematic of the natural circulation flow inside the RPV.

Gravity driven cooling system (GDCS),

Automatic depressurization system (ADS), which consists of the depressurization valve (DPV) and safety relief valve (SRV),

Isolation condenser system (ICS),

Standby liquid control system (SLCS),

Passive containment cooling system (PCCS), and Suppression pool (SP).

Figure VI-2 presents a schematic of the ESBWR including passive safety system. The GDCS and PCCS are unique to the ESBWR and SBWR. They represent the passive ECCS and containment cooling systems of currently operating BWRs. The ADS is actuated at a prescribed RPV condition and depressurizes the reactor pressure vessel so that the GDCS can be actuated to supply water into the RPV. The goals of the safety systems are to adequately cool the core by maintaining a water level above the active core and to provide a sufficient heat sink via the PCCS and SP in the wetwell (WW) to keep the containment pressure and temperature below the design criteria. The ICS is an engineered safety system, which participates in the most accident transients. It is functionally similar to that in operating BWR and acts as a passive decay heat removal system.

The initial state of the NPP is the operation at rated power. As a result of break of main coolant pipeline, the discharge of significant mass of coolant of the primary circuit takes place. When the pressure of the primary circuit becomes less than 13.7 MPa, the scram is initiated. Stop valves are closed after trip of the reactor. In this condition, the main coolant pump is switched off and coast down when the difference between the temperature of coolant in a hot leg and saturation temperature under actual pressure becomes bellow 10 °C. The steam generator PHRS starts up by the fact of decrease of the difference between primary and secondary side pressure. The boric acid solution is supplied to the reactor from traditional hydroaccumulators with initial pressure of 4.0 MPa by corresponding primary pressure decrease, without any signal actuation. The containment PHRS will provide condensation of the steam in the containment.

Thus, in the first stage of the accident, primary pressure is decreased due to loss of coolant and operation of the passive heat removal system. Further cooling down and pressure decrease are realized via steam generator PHRS and containment PHRS. When the pressure difference between primary circuit and containment is decreased to 0.6MPa, passive valves of the emergency depressurization system (primary circuit untightening subsystem) open to connect reactor inlet and outlet with the fuel pond space. When the reactor and containment pressure difference is decreased below 0.3 MPa, the ECCS tanks (elevated hydroaccumulators open to the containment) begin to flood the reactor. This sequence results in creating of so called emergency pool where the reactor coolant system is submerged to and in connection of this emergency pool with the spent fuel pond. The natural circulation along the flow path shown in Figure XII-3 (reactor inlet plenum — core — reactor outlet plenum — ‘hot’ depressurization pipe — fuel pond — ‘cold’ depressurization pipe — reactor inlet plenum) provides the long term heat removal from the core in case of the LOCA combined with complete loss of power supply. The water in the emergency pool and spent fuel pool reaches the saturation point in about 10 hours. The steam generated will condense on the internal surface of the steel inner containment wall, and condensate flows back into the emergency pool. This configuration ensures also the heat removal from reactor vessel bottom to keep the corium inside the reactor in case of postulated core melt event.

XII-6. Conclusions

Several passive safety systems based on natural circulation phenomena are used in WWER-640/407 reactors to fulfill the fundamental safety functions of reactivity control, fuel cooling, and radioactivity confinement. Implementation of these systems made it possible to significantly increase power plant safety in terms of the expected severe core damage and excessive radioactivity release frequencies.

REFERENCES TO ANNEX XII

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Natural circulation data and methods for advanced water cooled nuclear power plant designs, IAEA-TECDOC-1281 (IAEA, Vienna, 2000).

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs 2004, IAEA-TECDOC-1391 (IAEA, Vienna, 2004).

In case of an accident, the heat removal device should not release steam in the atmosphere during a SGTR (steam generator tube rupture). In case of over pressure transient, the released steam is condensed in a dedicated pool. The steam generator is not considered as the main system for decay heat removal. It acts as a thermal buffer until the safety systems on the primary side are fully operational.

XIX-4.1.2. System on the primary circuit

The primary system is cooled by means of the heat exchangers located in the downcomer (Figure XIX-4). Each exchanger has a dedicated heat sink. So there are sixteen independent loops, called RRP system (Residual heat Removal on Primary circuit). There are two types of heat sinks:

• Four RRP are cooled by immersed heat exchanger in a pool (RRPp);

• Other twelve are cooled by heat exchanger in air-cooling tower (RRPa).

All RRP are able to operate in natural convection in the loop and in the heat sink.

The design of the circuits is very simplified. The RRP loops are designed to resist the primary pressure. Isolating valves are placed on the circuit to avoid the risk of primary water passage outside the containment in the event of heat exchanger tube rupture. A surge tank for the water dilations from the cold shutdown to the full power operating condition carries out the pressure control of RRP circuit.

There is no control valve on the secondary loop, those are placed on the level of the heat sink: thermal valves or air leaves so that the temperature of the RRP loop remains high when the reactor is in power. Therefore, the RRP is ready to passively operate, just by the opening of the air leaves on RRP air coolers or by the opening of the thermal valves on RRP pools.

Forced convection is only required when the chilled cooling is needed for core refuelling. The twelve RRPa are able to cool the primary system down to cold shutdown. They replace the normal reactor heat removal system.

VI — 1. Introduction

The long operating cycle simplified boiling water reactor (LSBWR) is a modular boiling water reactor (BWR) plant that is designed by Toshiba Corporation. The reactor concept described in this section has a small power output, a capability of long operating cycle, and a simplified BWR configuration with comprehensive safety features. To be economically competitive, simplification of systems and structures, modularization for short construction period, and improvement in availability are included into the LSBWR design. For comprehensive safety features, the aim is to need no evacuation by utilizing highly reliable equipment and systems such as large RPV inventory, bottom located core layout, in-vessel retention (IVR) capability and passive emergency core cooling system (ECCS) and primary containment vessel (PCV) cooling. Figure VII-1 shows conceptual drawing of the LSBWR.

FIG. VII-1. Conceptual drawing of the LSBWR.

Natural circulation core cooling is applied for eliminating recirculation pumps. This results in high reliability in operation. For attaining natural circulation core cooling, the fuel length is shortened to 2.2m from the conventional 3.7m to decrease the pressure drop. Figure VII-2 shows the core and fuel bundle for the LSBWR. Figure VII-3 shows the LSBWR reactor internals and configuration.

Fuel Rod

1.2C lattice bundle Part Length Rod

FIG. VII-2. Core and fuel bundle for the LSBWR.

FIG. VII — 3. LSBWR reactor concept.

The cylindrical type drywell with small diameter can be designed by routing the safety relief valve piping through the spacing between the RPV and the drywell wall and the main vent pipe from the RPV top to the suppression pool. The drywell air space is minimized and contains only SRV and depressurization valve (DPV) components, the gravity driven core cooling system (GDCS) and drywell loading piping. Since MS and FW piping is routed through suppression pool air space, which are protected by the guard pipe, GDCS piping is contained in the access tunnel placed in the lower part of suppression pool, and isolation valves are installed outside PCV etc.

Since the reactor core is placed at the bottom of the RPV, the emergency coolant injection system consisted of DPV and GDCS can achieve high reliability of the water coverage of the reactor core following an accident.

The containment wall with ship hull structure is filled with cooling water that is boiled off to the atmosphere to cool the PCV passively during an accident. This containment wall cooling system is also used for the drywell cooling during the normal operation and therefore the drywell arrangement is simplified without drywell cooling components used in the current BWR containment.

When cooling water in the PCCS pool above PCV is exhausted, external pool or seawater is supplied by gravitation so that the highly reliable and long term PCV cooling is achieved.

The double cylindrical raised suppression pool with the ship hull structure is installed around the cylindrical drywell and above the core elevation. This makes the structure stronger and simpler, and the suppression pool water can be easily used for GDCS and drywell lower part flooding. LSBWR safety system concept including PCCS is shown in Figure VII-4.

The performance of the safety system has been analyzed for a feedwater line break accident. The analysis has been performed using TRAC code incorporated with the heat transfer models for the natural convection cooling and the steam condensation cooling with a noncondensable, which have been used to estimate the heat transfer coefficients in the containment space and the containment wall coolant channels. The analysis results for the containment pressure and the heat removal rate by the passive containment cooling system are shown in Figure VII-5 and VII-6. After taking its peak value during the blowdown phase, the containment pressure keeps decreasing while the GDCS coolant flow is sufficient to suppress the steam production in the reactor core. The containment pressure begins to increase around 3 hours since the GDCS flow decreases and the steam is produced by the decay heat. The pressure increase is, however, suppressed by the containment wall cooling and is maintained well below the design pressure for 24 hours. The heat removal rate of the containment wall cooling becomes almost comparable with the decay heat after 12 hours. The condensate produced by the containment wall cooling flows from the drywell to the RPV through the GDCS injection line and the reactor core is kept covered.

VI — 4. Conclusions

The LSBWR design is still at the conceptual design stage and licensing reviews have not yet been started. Recently, Compact Containment BWR development is going on based on various experience obtained at LSBWR development.

FIG. VII-6. Long term containment pressure transient.

Mitsubishi, Japan

XII — 1. Introduction

The Integrated Modular Water Reactor (IMR) is one of the integrated primary system reactors (IPSRs) with the reference output of 1,000 MWt (350 MW(e)). The design targets of IMR are to achieve the electricity generating cost comparable to that of a large-scale nuclear reactor and higher-level safety by removing the sources that cause fuel failures by design. To achieve these targets, IMR employs integrated design with in-vessel CRDM (control rod drive mechanisms), hybrid heat transport system (HHTS) which employs two-phase natural circulation for the primary heat transportation, and stand­alone direct heat removal system (SDHS) for accident heat removal from the primary system.

IMR started its conceptual design study in 1999 at Mitsubishi Heavy Industries (MHI) reflecting changes of business environment such as less growth of economy and electricity demand, and deregulation of electricity market in Japan. An industry-university group led by MHI (Kyoto University, Central Research Institute of Electric Power Industries (CRIEPI), the Japan Atomic Power Company (JAPC), and MHI) has been developing related key technologies funded by Japan Ministry of Economy, Trade and Industry from 2001 to 2004. In this project, the concepts and the feasibility of HHTS and SDHS have been built and tested through three series of experiments. They are (1) air — water scale test to confirm void distribution and void behavior in the reactor, (2) high temperature natural circulation test to study two-phase natural circulation in the reactor with the actual temperature, pressure, and axial dimension of IMR, (3) SDHS test to study passive heat transport with the actual temperature, pressure, and axial dimension of SDHS. These test facilities were built and set at MHI Takasago R&D centre. In-vessel CRDM technology is based on marine reactor (MRX) development by Japan Atomic Energy Agency (JAEA) and MHI.

Here, the concepts and the feasibility test results of HHTS and SDHS are summarized.