The primary circuit untightening subsystem is designed to ensure reliable water flow from atmospheric ECCS tanks and natural circulation of coolant along the circuit ‘reactor-emergency pool’. This system operates during LOCAs when the primary circuit pressure decreases to 0.6 MPa. The system consists of two independent trains with redundancy of 2 x 100%. Each train comprises two pipelines and valves connecting the primary hot and cold legs to the emergency pool located in the lower part of the containment. This provides long term residual heat removal after the ECCS accumulator tanks and atmospheric tanks are emptied. The system train includes untightening valve units, pipeline and valves for connecting the refueling pool with the emergency pool, repair valves, and a screen filter.

The SCOR pressurizer is integrated into the upper part of the riser, just below the steam generator. The pressurizer region is designed as an annular shape in the form of an inverted U, see Figure XIX-1. The coolant flows through the central part of the pressurizer. The bottom portion of the inverted U contains the opening to allow water insurge and outsurge to/from the pressurizer region

The electric heaters are located in a small volume tank outside the reactor vessel and act as a steam source. The cold water supply is tapped off just downstream the pumps and the two-phase mixture is reinjected at the top of the pressurizer.

Owing to the low pressure and low temperature operating point leading to smaller variations of water density versus temperature, the volume of the pressurizer is smaller than those used in plants with a classical, separate, pressurizer vessel. The SCOR pressurizer has a total volume of ~21 m3. This volume is large enough to manage a blackout without steam release through the safety valve of the pressurizer.

The reactor is operated at high pressure (7.2 MPa) which needs to be depressurized to a value closer to the containment pressure in order to gain a driving head between the GDCS tanks and the RPV. The depressurization of the RPV is achieved through the ADS during blowdown period. Once the RPV is depressurized, the GDCS injects water into the RPV without relying on any active systems by using the gravitational head between the GDCS tanks and the RPV. The GDCS functions to remove the core decay heat to avoid the core uncovering and melting. Although the PCCS is actuated during the GDCS injection period, the PCCS is not functional since there are no driving forces between the DW and WW. In other words, the GDCS water cools down the RPV and the containment pressure keeps reducing during this period.

The CAREM nuclear power plant design is based on a light water integrated reactor. The whole high — energy primary system, core, steam generators, primary coolant and steam dome, is contained inside a single pressure vessel (Fig. XIV-1).

For low power modules (below 150 MW(e)), the flow rate in the reactor primary systems is achieved by natural circulation. Figure XIV-1 shows a diagram of the natural circulation of the coolant in the primary system. Water enters the core from the lower plenum. After it’s heated the coolant exits the core and flows up through the riser to the upper dome. In the upper part, water leaves the riser through lateral windows to the external region. Then it flows down through modular steam generators, decreasing its enthalpy. Finally, the coolant exits the steam generators and flows down through the down-comer to the lower plenum, closing the circuit. The driving forces obtained by the differences in the density along the circuit are balanced by the friction and form losses, producing the adequate flow rate in the core in order to have the sufficient thermal margin to critical phenomena. Reactor coolant natural circulation is produced by the location of the steam generators above the core.

II — 1. Introduction

The APWR+ is a four loop type PWR with 1750 MW(e) output, which is being developed as the successor of APWR and conventional PWRs, aiming at more enhancements in economy, safety, reliability, reduction of the operators’ workload, and harmony with the environment. The development is being carried out by Japanese PWR utilities (Kansai Electric Power, Hokkaido Electric Power, Shikoku Electric Power, Kyushu Electric Power, and The Japan Atomic Power Co.) and Mitsubishi Heavy Industries.

APWR+ employs the following concepts for its safety system including passive features:

a) Passive safety equipment and system

• Core cooling using steam generator and natural circulation of the primary system during

accident

• ‘Advanced Accumulators’ and ‘Advanced Boric Acid Injection Tank’.

b) Four train configuration

c) Confinement of energy release to the containment

• The core and the loop piping are submerged in case of a LOCA

d) Enhancement of the design diversity

• Emergency power supply by Diesel and gas turbine

• Heat removal paths from the containment

Figure VI-1 shows the features of APWR+ safety related systems.

The main goal of this advanced BWR is to replace the active safety systems used in current designs with passive safety systems enabling:

• Reliable control of the various design basis accidents;

• Low probability of beyond-design-basis accidents (core damage frequency);

• Limitation of the consequences of a core melt accident to the plant itself;

• High plant availability;

• Economic competitiveness.

Various features have been changed compared to existing BWR designs, including:

• Larger water inventory in the reactor pressure vessel (RPV) above the core permits passive core cooling;

• Larger water storage capacities inside and outside the reactor containment providing long grace periods and avoiding the need for prompt operator intervention, especially during and after accidents;

• For transients as well as for accident control, emergency condensers and containment cooling condensers to passively remove decay heat from the core and containment, respectively;

• Activation of key safety functions such as reactor scram, containment isolation and automatic depressurization is backed up by passive systems (passive pressure pulse transmitters);

• Passive cooling of the reactor pressure vessel exterior in the event of a core melt accident ensures in-vessel melt retention;

• Despite the introduction of passive safety systems for accident control the operating experience gained from current BWR plants constitutes the basis for the new concept;

• Simplification of reactor auxiliary systems and systems used for normal power operation.

Idaho National Laboratory, Oregon State University, Nexant, USA

XVII — 1. Introduction

The multi-application small light water reactor (MASLWR) is a 150 MW(t) modular nuclear reactor that uses natural circulation for primary loop cooling. The MASLWR system was designed by the Nuclear Energy Research Initiative (NERI), a program of the United States’ Department of Energy (DOE). Other collaborators included Idaho National Engineering and Environmental Laboratory (INEEL), Oregon State University (OSU), and NEXANT-Bechtel, all in the United States of America. The design philosophy was to use existing pressurized water reactor (PWR) knowledge to develop a safe and economical small power source using only natural circulation in the primary system. Through economic considerations, the project moved towards design of factory-fabricated modules, designed to operate in a power farm of 30 units producing 1050 MW(e), although a fraction of that, down to a single module, could also be used in a given location. A single MASLWR module produces 35 MW(e), using 8% enriched fuel and can last for 5 full-power years without replacement.

Figure XVII-1 shows a single power generation module for the MASLWR design. MASLWR implements an integrated reactor vessel with an internal helical coil steam generator. The reactor vessel is enclosed in a high pressure steel containment vessel that is partially filled with water to serve as a suppression pool. The containment vessel in turn resides in a large exterior cooling pool that acts as the ultimate heat sink.

FIG. XVII-1. Schematic of the MASLWR exterior cooling pool and turbine-generator set.

VI — 1. Introduction

General Electric (GE) has developed a new passive safe boiling water reactor called the economic simplified boiling water reactor (ESBWR), which is based on the previous simplified boiling water reactor (SBWR) design with some modifications of safety systems and the containment size relative to the reactor power [1, 2]. Major differences between the current boiling water reactors (BWR) and the ESBWR are in the simplification of the coolant circulation system and the implementation of a passive emergency cooling system. The ESBWR reactor core has a rated thermal output of 4500 MW*th.

The ESBWR relies on natural circulation and proven passive systems to improve safety, economics, and performance. In ESBWR concepts, the safety is accomplished by eliminating the recirculation pump, thus relying on natural circulation cooling. The coolant is circulated by natural circulation as a result of the density difference between the high void, two-phase fluid in the chimney and the exterior single-phase liquid in the downcomer. The tall chimney not only enhances the natural circulation flow, but also ensures the ample time for core uncovery before the emergency core cooling system (ECCS) comes in play. The emergency core cooling and containment cooling systems do not have an active pump injecting flows and the cooling flows are driven by pressure differences. Large volumes of suppression pool (SP) functions not only as a primary heat sink during the initial blow down, but also as coolant inventory to prevent the core uncovery through the gravity equalization lines.

By relying on natural circulation at operating pressures (7.2 MPa) and increased chimney height, the ESBWR has enhanced natural circulation flow inside the vessel. The schematic of natural circulation inside the reactor pressure vessel (RPV) is shown in Figure VI-1. The driving head of core flow is proportional to the core and chimney height and void fraction inside the downcomer shroud. In ESBWR, the differential water level is increased by approximately 8.2 m compared to the conventional BWRs. The greatly increased driving head enhances the natural circulation flow in the ESBWR compared to the conventional BWRs. Aforementioned ESBWR design features results in an average core flow per bundle over three times greater than that of a conventional BWR under natural circulation at similar bundle power. The use of natural circulation eliminates pumps, motors, controls, piping and many other components that could possibly fail.

The steam generator passive heat removal system (SG-PHRS) for the V-407 reactor does not require a power supply and is designed to remove the decay heat in case of a non-LOCA. It also supports the emergency core cooling function in case of a LOCA. The reactor coolant system and passive heat removal equipment arrangement provides heat removal from the core following reactor shutdown via steam generator heat transfer to the tanks of chemically demineralized water (CWT) outside the containment and further to the atmosphere by natural circulation as shown in Figure XII-4. Reactor power that can be removed from the core by coolant natural circulation is about 10% of the nominal value, which ensures a reliable residual heat removal. In case of a non-LOCA the decay heat is removed by coolant natural circulation to steam generator boiler water. The steam generated comes into the passive heat removal system where steam is condensed on the internal surface of the tubes that are cooled on the outside surface by the water stored in the demineralized water tank. These tanks are stored outside the containment. The water inventory in this tank is sufficient for the long term heat removal (at least 24 hours) and can be replenished if necessary from an external source.

FIG. XII-4. Passive heat removal for non-LOCAs.

There is only one U tube boiler steam generator. Like propulsion reactors, the SG is placed above the core. In contrast to standard SGs, the present generator has an axial symmetry. The hot leg is located in the centre and the cold leg is located in periphery.

XIX-3.6. Operating point

According to the results from a low pressure PWR studies, SCOR operates at a pressure of 88 bars. The secondary pressure at the turbine inlet is 30 bars, which is lower than that of standard PWRs. The net thermodynamic efficiency is 31,5%, slightly smaller than that of standard PWRs.

XIX-4. Description of the safety systems

XIX-4.1. Decay heat removal systems

Since the reactor has only one steam generator, the decay heat removal systems are diversified on both the primary and secondary circuit.