The ADS is a part of the ECCS. The ADS depressurizes the RPV in the event of LOCA to allow the GDCS water injection into the vessel, preventing the core uncovery. Once the ADS actuates, it continuously operates to keep the reactor depressurized for GDCS injection after an accident initiation. The ADS in the ESBWR is composed of ten safety relief valves (SRVs) and eight depressurization valves (DPVs). The SRVs are mounted on the top of the main steam lines in the DW and discharge the steam through lines routed to quenchers in the SP. Four DPVs are horizontally mounted on horizontal stub tubes connected to the RPV at about the elevation of the main steam lines. The other four DPVs are horizontally mounted on horizontal lines branching from each main steam lines. Main function of the DPVs is to discharge the steam directly from the RPV to the DW in order to depressurize the RPV during the initial phase of LOCA.

The ADS automatically actuates on a reactor low level (Level 1) signal that persists for at least 10 seconds. A two-out-of-four Level 1 logic is used to activate the SRVs and DPVs. The 10-second persistence requirement for the Level 1 signal ensures that momentary system perturbations do not actuate the ADS when it is not required. The two-out-of-four logic ensures that a single failure does not cause spurious system actuation.

The quick boron supply system (QBSS), being developed as an additional reactor trip system, comprises a system of 4 special loops bypassing the main coolant pumps. Each loop consists of a hydroaccumulator containing concentrated boron acid solution and pipelines with quick-acting valves that do not require electric power for their opening. These valves are opened when the reactor fails to scram. When this occurs concentrated boron solution is pressed out of the hydro-accumulators into the primary loops and further into the reactor. Tn case of a station blackout the boron solution delivery occurs in the period of reactor coolant pump (RCP) coast-down.

The RCPs have a large flywheel inertia which provides the possibility of ejecting all boron concentrate from the QBSS hydro-accumulators. The amount and concentration of the boron solution are chosen to provide a certain equivalency from the viewpoint of reactivity inserted by this system and by the solid absorber scram. Tn fact, this system, being a part of the primary coolant circulation system increases the ‘inherent’ safety of the reactor design.

FIG. XIII-1. Containment and the systems of WWER-1000/392 that used passive principles. (I — reactor; 2 — steam generator; 3 — steam path; 4 — condensate path; 5 — inlet circular header; 6 — PHRS heat exchanger; 7 — PHRS slide valve; 8 — PHRS draught tube; 9 — PHRS exit header ; 10 — deflector; 11 — quick boron supply system; 12 — HA-2; 13 — HA-1; 14 — filtering unit; 15 — tube of passive filtering system; 16 — steam header; 17 — valve).

In hot conditions, the residual heat is removed through the steam generator. The steam is discharged to the atmosphere, and the Steam Generator is fed by the start-up shutdown system (SSS). The system is not safety grade. Then, at low temperature, the RRP with the air-cooling tower (RRPa) removes the decay heat.

When the vessel is opened, especially during the refuelling operating, decay heat is removed by the twelve RRPa cooled by chilled water in order to obtain a very low primary temperature compatible with the maintenance action conditions. The primary circuit operates in natural convection and the RRPa loops operate in active condition.

Chilled water is only used during the refuelling operation. In case of chilled water circuit or RRPa pump failure, the heat sink is backed up by the air-cooling tower of the RRPa. The RRP system replaces the normal residual heat removal system with an external loop like in standard PWRs.

XIX-4.3. Safety injection

Since large LOCAs are eliminated by design and since the primary system thermal inertia is larger than that of loop-type PWRs, the safety injection system requires devices with a small flow rate. Given the intrinsic low pressure option for the reactor, there is only one type of safety injection with a pressure of about 20 bars. The needed pump power for the safety injection is very small, about 35 kWe.

II — l. Introduction

The Indian nuclear power program consists of three stages. The first stage envisaged setting up of pressurized heavy water reactors (PHWRs) and the necessary fuel cycle facilities. Comprehensive capability in the design, construction, and operation of PHWR has been achieved. The second stage envisages development of fast breeder reactors using Plutonium and Depleted Uranium obtained from the first stage. Fast breeder test reactor (FBTR) has been operated successfully for 21 years and the construction of 500 Mw(e) prototype fast breeder reactor (PFBR) has started. The third stage aims at the development of reactors based on Uranium-233 obtained from irradiated thorium. The Kamini reactor uses Uranium-233 as a fuel and has operated since 1995. To transition to thorium based systems, an advanced heavy water reactor (AHWR) is being developed at Bhabha Atomic Research Centre.

The Bhabha Atomic Research Centre (BARC), located in Mumbai, is the premier multidisciplinary nuclear research centre of India having excellent infrastructure for advanced research and development. It has expertise covering the entire spectrum of nuclear science, engineering, and related areas. Detailed design, research, and development work on AHWR is being conducted in BARC.

The AHWR is a 300 MW(e) boiling light water cooled, heavy water moderated, vertical pressure tube type reactor designed to produce most of its power from thorium with an associated 500 m3/day capacity desalination plant. The core consists of (Th-U233) O2 and (Th-Pu)O2 fuel. A simplified sketch of the reactor is depicted in Fig. III-1. Some important features of the reactor are given below.

• Thorium based fuel with a negative void coefficient of reactivity,

• Advanced coolant channel design with easily replaceable pressure tubes,

• Passive systems for core heat removal (under both normal operating and shutdown condition), containment cooling and containment isolation,

• Direct injection of ECCS water into fuel bundle,

• Accumulator for high pressure ECC,

• Gravity driven water pool (GDWP) at high elevation,

• No emergency planning in public domain,

• Design life of 100 years,

• Associated desalination plant.

A number of passive systems that utilize natural circulation have been incorporated in AHWR. Some of them are briefly described below.

The isolation condenser will be utilized for the 330 MW(e) RMWR, which is basically the same as the conventional one and is, therefore, not described in here. The passive containment cooling system (PCCS) with horizontal heat exchangers will be utilized for the large-size RMWR, details of which are described in the section for the ABWR-II in this report.

FIG. VIII-4. Schematic of axial core configuration for the
330 MW(e) RMWR.

FIG. VIII-5. Plant system concept for the 330 MW(e) RMWR.

Research and development (R&D) activities for the RMWR have been conducted by the JAERI in collaboration with the Japanese industries, which have clarified the favourable characteristics of the reactor including high conversion ratio of more than one, high burn-up, long operation cycle, and multiple recycling of plutonium. So far, the R&D has been conducted under several domestic frameworks, including a) the research corporation program between the Japan Atomic Power Company (JAPC) and JAERI where many RMWR systems were studied, b) the innovative and viable nuclear energy technology (IVNET) development project where the 330 MW(e) RMWR was developed with JAPC, Hitachi Ltd. and Tokyo Institute of Technology (IWAMURA, T., 2002), c) the program sponsored by the ministry of education, culture, sports, science and technology (MEXT) on the innovative nuclear reactor technologies where the bundle core heat transfer tests are conducted together with broad ranges of R&D for fuel and neutronics (OHNUKI, A., 2004, KURETA, M., 2004, YOSHIDA, H., 2004), and d) the feasibility study on the commercialized fast reactor cycle systems conducted by the Japan nuclear cycle development institute. In addition to the above domestic frameworks, the JAERI entered into the agreement with the USDOE on the tight lattice core design in December, 2004.

REFERENCES TO ANNEX VIII

[1] IWAMURA, T., et al., Core and System Design of Reduced-Moderation Water Reactor with Passive Safety Features, Proc. of ICAPP ’02-220 Int. Cong. On Advan. Nucl. Pow. Plants, Florida, USA (2002) (CD-ROM) 8page.

[2] KURETA, M. et al., Development of Predictable Technology for Thermal/Hydraulic

Performance of Reduced-Moderation Water Reactors (2) — Large — scale Thermal/Hydraulic Test

and Model Experiments, ICAPP’04, 4056 Pittsburgh, USA (2004).

[3] OHNUKI, A. et al., Development of Predictable Technology for Thermal/Hydraulic

Performance of Reduced-Moderation Water Reactors (1) — Master Plan, ICAPP’04, 4055 Pittsburgh, USA (2004).

[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs 2004, IAEA-TECDOC-1391 (2004), pp.418-435.

[5] YOSHIDA, H. et al., Development of Predictable Technology for Thermal/Hydraulic

Performance of Reduced-Moderation Water Reactors (3) — Current Status of Development of Three-Dimensional Two-Phase Flow Simulation Method, ICAPP’04, 4057 Pittsburgh, USA (2004).

IMR realizes ‘no cause of fuel failure’ concept by reactor design and the core is always submerged in primary coolant without safety injection. Therefore, only a heat removal function is required as the safety system. Figure XV-2 shows the configuration of the stand-alone direct heat removal system (SDHS). SDHS is a closed natural circulation system that removes residual heat directly from inside the RV to the atmosphere via SGs and passive SG coolers (PSGCs). The reactor is cooled down and depressurized without opening the primary system pressure boundary. Residual heat in the early stage of an accident is removed by latent heat of cooling water by submerging PSGCs. After the cooling water was dried out, residual heat will be removed by air-cooling, since the dried water pool and ducts form a wind tunnel. Therefore, the heat transfer mode in PSGCs automatically changes from water­cooling to air-cooling following the pool water evaporation and SDHS works a long time without any operator action and external support.

The reactor design and SDHS concept greatly simplify the safety system of IMR compared with conventional PWRs. There is no safety injection system and containment spray system, and SDHS makes support systems to change into non-safety systems, which are component cooling water system, essential service water system, and emergency AC power system. To show the feasibility of SDHS, the safety system of IMR, natural circulation and heat removal capability of the system including effect of non-condensable gas has been experimentally examined. SDHS is designed to accumulate non-condensable gas into the lower header of PSGC and the feed water tank of the SGs.

XIV — 4. Conclusions

The Integrated Modular Water Reactor (IMR) employs two natural circulation systems, the hybrid heat transport system (HHTS) for the primary system and the stand-alone direct heat removal system (SDHS) for the safety heat removal. The design concepts of IMR and these systems have been built and the test results showed the feasibility of the system. In the next design phase of IMR, 3D effects of two-phase natural circulation flow behaviour will be tested as a part of basic design study.

REFERENCES TO ANNEX XV

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs, p715-732, IAEA, IAEA-TECDOC-1391, Vienna (2004).

[2] TAKANO, K., et al., Integrated Modular Water Reactor (IMR), Development for Practical Application in the Near Future, Proc. of 14th Pacific Basin Nuclear Conference (PBNC 14th), Honolulu, USA (2004).

[3] HIBI, K., et al., Integrated Modular Water Reactor (IMR) Design, Nuclear Engineering and Design, 230, 253-266 (2004).

[4] KANAGAWA, T., et al., The Design Features of Integrated Modular Water Reactor (IMR), Proc. of ICONE-12, #49528, Arlington, USA (2004).

[5] SUZUTA, T., et al., Development of Integrated Modular Water Reactor — Natural Circulation Tests in Reactor Vessel, Proc. of ICAPP04, #4086, Pittsburgh, USA (2004).

[6] SERIZAWA, et al., Two-phase Flow in Natural Circulation System of the Integrated Modular Water Reactor (IMR), Proc. of the 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6), #N6P132, Nara, Japan (2004).

[7] TAKANO, K., et al., Assessment of Bubble Behavior in Two-phase Flow Natural Circulation Utilized for Primary System of the Integrated Modular Water Reactor (IMR), Proc. of Fourth Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS4), #062, Sapporo, Japan (2004).

[8] SUZUTA, T., et al., Steam-Water Natural Circulation Tests for Integrated Modular Water Reactor (IMR), Proc. of Japan-US Seminar on Two-Phase Flow Dynamics, Nagahama, Japan (2004).

[9] HIBI, K., et al., Improvement of Reactor Design on Integrated Modular Water Reactor (IMR) Development, Proc. of ICAPP05, #5215, Seoul, Korea (2005).

[10] TANI, et al., Plant dynamics and Controllability of IMR, Proc. of ICAPP05, #5181, Seoul, Korea (2005).

[11] INOUE, K., et al., Safety system design and Stand-alone Direct Heat Removal System (SDHS) for Integrated Modular Water Reactor (IMR), Proc. of ICAPP05, #5180, Seoul, Korea (2005).

[12] SUZUTA, T., et al., Steam-Water Natural Circulation Tests Simulating the Integrated Modular Water Reactor (IMR), Proc. of ICONE-13, #50641, Beijing, China (2005).

[13] SUBKI, M. H., et al., Steam-water Test Simulation by RELAP5/MOD3.2 for Two-phase Flow Natural Circulation System on the Integrated Modular Water Reactor (IMR), Proc. of ICONE — 13, #50591, Beijing, China (2005).

[14] TAKANO, K., et al., Bubble Behavior in Two-phase Flow Natural Circulation Employed in the Primary System on the Integrated Modular Water Reactor (IMR), Analyzed by a-flow Code, Proc. of ICONE-13, #50590, Beijing, China (2005).

The passive residual heat removal (PRHR) consists of a C-Tube type heat exchanger that resides in the water-filled In-containment refueling water storage tank (IRWST) as shown in the schematic given in Figure V-4. The PRHR provides primary coolant heat removal via a natural circulation loop. Hot water rises through the PRHR inlet line attached to one of the hot legs. The hot water enters the tubesheet in the top header of the PRHR heat exchanger at full system pressure and temperature. The IRWST is filled with cold borated water and is open to the containment Heat removal from the PRHR heat exchanger occurs by boiling on the outside surface of the tubes. The cold primary coolant returns to the primary loop via the PRHR outline line that is connected to the steam generator lower head.

The flooding pools are equipped with active coolers. If the active cooling fails, the 3600 m3 pool will not start to evaporate for twelve hours. During a failure the pressure in the containment is increased and the heat is transferred by four containment cooling condensers into the upper shielding pool (see Fig. XI-1, XI-2 and XI-4). Each condenser is able to transfer up to 5.5 MW at a pressure of 0.25 MPa. Each condenser consists of 124 finned tubes arranged in two layers (design state 1998). The tubes have a length of 4 m, an inner diameter of 32 mm and a tube wall thickness of 3 mm. The tubes are arranged with a twelve degree incline. In addition, a second version of the cooling condenser was designed to remove non condensable gases.

FIG. XI-4. Working principle of the containment cooling condenser.

The core is cooled mostly by the single-phase free convection due to natural circulation. The natural circulation loop consists of the core, the chimney, the downcomer and the lower plenum. In order to establish natural circulation at different RPV liquid levels during normal operations and accidents, the core barrel has large communication-holes above the SG top elevation and small holes at several elevations below that. The bypass flow through the small holes is approximately 5% of the core flow during the full-power operational condition. Since the RPV coolant is self-pressurized, that is, the pressure is determined by the temperature at the liquid level in the RPV, the core outlet is saturated. The fuel assembly is based on the 17×17 fuel assembly design with Zircaloy-4 cladding UO2 pellets that is used for the current PWRs except for the fuel pin pitch. The pitch of the PSRD is 13.9mm, which is 1.3mm wider than that of the current PWRs to enhance burn-up by increasing the moderation effect. This geometry decreases the flow resistance along the core.

TABLE XVin-1. MAJOR CHARACTERISTICS OF THE PSRD-100

FIG. XVIII-1. Concept of PSRD

So far, the PSRD natural circulation core cooling has been evaluated for steady and transient states by using the thermal-hydraulic analysis codes. During such analyses, the load following capability was analyzed using the RETRAN-02 code by decreasing the feedwater flow rate from 100% to 50% in 200 sec, keeping it at 50% for 200 sec, and then increasing from 50% to 100% in 200 sec to represent a typical day-load change as shown in Fig. XVIII-2(a) (ISHIDA, T., 2003-2). The automatic reactor control system was not used to clarify the inherent self-controllability of the system. The calculation results showed that the reactor power responded well with the delay time of 50 to 100 sec and the overshoot of up to 10% as shown in Fig. XVIII-2 (b). The response of the natural circulation flow rate was also stable without showing high-frequency oscillations as shown in Fig. XVIII-2 (c). The results indicated the inherently stable nature of the PSRD natural circulation cooling system. The results also indicated that the responses will be more stable for the actual load follow condition with the operation of the reactor automatic control system without excessively depending on the use of the control rod operation.

During a LOCA, the reactor shuts down and the RPV is isolated by closing the main steam line isolation valves. The ICS removes decay heat after any reactor isolation. In other words, the ICS passively removes sensible and core decay heat from the reactor when the normal heat removal system is unavailable. Decay heat removal limits further increases in steam pressure and keeps the RPV pressure below the safety set point. The arrangement of the IC heat exchanger is shown in Fig. VI-4. The ICS consists of four independent loops, each containing two heat exchanger modules that condense steam inside the tube and transfers heat by heating/evaporating water in the IC pool, which is vented to the atmosphere. This transferring mechanism from IC tubes to the surrounding IC pool water is accomplished by natural convection, and no forced circulation equipment is required.

The ICS is initiated automatically by any of the following signals: high reactor pressure, main steam line isolation valve (MSIV) closure, or an RPV water Level 2 signal. To operate the ICS, the IC condensate return valve is opened whereupon the standing condensate drains into the reactor and the steam water interface in the IC tube bundle moves downward below the lower headers.