Different variations of a passive safety containment for a BWR with active and passive safety systems

Nuclear Engineering and Design 235 (2005) 2125–2139

Different variations of a passive safety containment for a BWR with active and passive safety systems Takashi Sato ∗ , Hirohide Oikawa, Makoto Akinaga, Tsunekazu Murakami Toshiba Corporation, IEC, Gen-SS, 8 Shinsugita-cho, Isogo-ku, Yokohama, Japan Received 9 August 2004; received in revised form 20 January 2005; accepted 1 March 2005

Abstract The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR+ (ASBWR+ ) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper. © 2005 Elsevier B.V. All rights reserved.

1. Introduction Recently, passive-safety reactors are prevailing in the licensing review by the U.S. NRC (General Electric Company, 2002; Cummins and Schulz, 2004; Brettschuh and Hudson, 2004; Snell et al., 2004). Among them, there is the ESBWR as a passive safety BWR. It is based on the technology developed for the ∗ Corresponding author. Tel.: +81 45 770 2066; fax: 81 45 770 2179. E-mail address: [email protected] (T. Sato).

0029-5493/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2005.03.008

SBWR (Rao et al., 1991). Toshiba also participated in the international joint study for the SBWR development. For example, we conducted GIRAFFE tests for the passive containment cooling system (PCCS) development (Nagasaka et al., 1991). The basic design of the gravity-driven cooling system (GDCS) and the PCCS is almost the same between the SBWR and the ESBWR. The containment configuration is also very similar between them. However, the plant output is quite different. The SBWR is only about 600 MWe, while the ESBWR is 1380 MWe. The larger plant output is, the bigger impact on containment design. We thought

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of the use of a kind of improved Mark III type containment to reduce the containment pressure at severe accidents. The Mark III containment is the containment for BWR/6 (General Electric Company, 1980). It has the biggest free volume of the pressure suppression type containments for BWRs. This can provide the potential to reduce the pressure due to hydrogen compression at severe accidents. The Mark III containment also has an advantage of double barrier function against fission product (FP) release to some extent. Therefore, the Mark III containment is very useful for passive safety BWRs. In order to apply the Mark III type containment to the SBWR, however, we have to modify it and make a brand-new containment. We call it Mark X containment tentatively in the paper. The Mark X containment has a very large free volume. This is effective to reduce the peak pressure at accidents. However, if we do not inert the containment, there will be potential of hydrogen detonation. In order to remove the potential risk, we decided to inert the Mark X containment with nitrogen. The Mark X containment has double containment function, to some extent, against FP release not only for particulates but also for noble gases and organic iodine. This safety function is very rare among the passivesafety reactors under the licensing review of the U.S. NRC. Most of them have only a single containment barrier. Only the SWR 1000 has double containment barriers (Brettschuh and Hudson, 2004). The advanced safer BWR (ASBWR+ ) has walkaway safety owing to the GDCS and the PCCS. The terminology of walk-away safety here means safety that requires neither operator actions for initiation nor mechanical or human intervention for continuation. It has been used for inherent safe reactors to show their ability of grace period (Waltar et al., 1985; Sefidvash, 1996). The ASBWR+ is not an inherent safe reactor. It is rather an intended safe reactor (Sato et al., 1995). However, an intended safe reactor also can have walk-away safety. If a severe accident occurs, in the ASBWR+ , the GDCS floods the lower dry well (DW) to cool the core debris spread on the core catcher. The squib valves on the lower DW flooding lines are opened by passive means when signaled by thermocouples in the lower DW. After the debris cooling is established, the containment is cooled by the PCCS. The PCCS is a fully passive safety system and no valve operation is needed to initiate its safety function of containment cooling.

The water source of the PCCS is available for 3 days. After depletion of the PCCS water source, the Mark X containment can be cooled by natural circulation of outside air. Owing to this function, the ASBWR+ has the capability of permanent walk-away safety. The latest concept of active safety BWRs, such as ABWR2010+ has passive safety systems in addition to active safety systems (Sato et al., 2004). The Mark X containment is also useful for these kinds of half-passive safety BWRs. They can also establish permanent walk-away safety using their passive safety systems and the Mark X containment. The combination of active safety systems and the passive safety function of the Mark X containment can provide in-depth hybrid safety.

2. Design objectives of the ASBWR+ The design objectives of the advanced safer BWR are as follows: (1) to provide a containment that can accommodate severe accidents within the design pressure; (2) to provide a containment that has a double containment function against fission products release to a certain extent without relying on active safety systems; (3) to provide a containment that has no potential risk of hydrogen detonation without relying on uncertain measures, such as igniters; (4) to provide a passive safety BWR that has permanent walk-away safety; (5) to provide economical competitiveness to any other passive-safety reactors. The design pressure of a containment vessel (CV) has been decided based on the design basis accident (DBA), namely, loss of coolant accident (LOCA). However, when we assume the DBA LOCA, we also assume the integrity of the core owing to the single failure proof emergency core cooling system (ECCS). Originally, if the integrity of the core is assured, there is no acute need for the containment. It is only when the core is damaged that a containment is truly required. Our safety discipline has taught that a nuclear power plant is to be designed based on the defense in-depth principle (U.S. AEC, 1973). However, in reality, there are no specific safety criteria that explicitly require the defense in-depth principle. Our safety criteria rather allow us to design a containment not to obey the defense in-depth principle. The design pressure of the containment can be decided without assuming a large

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amount of hydrogen generation based on our safety criteria (U.S. NRC, 2003). However, we learned that from probabilistic safety assessment (PSA), the DBA LOCA is not a dominant or representative accident scenario for containment design, because the core is not damaged. Nevertheless, the DBA LOCA still decides the design pressure of a containment vessel. On the contrary, if a dominant sequence occurs, the core must be damaged and the containment is truly needed. Recently, all the advanced light water reactor (ALWR) requirements require us to assume 75–100% metal–water reaction (EDF et al., 1995; Taylor, 1990; EPRI, 1997). However, they still allow that peak pressure of the containment at severe accidents can be higher than the design pressure, because they still regard severe accidents as design extension condition (DEC) or beyond DBA. If the core is damaged, there must be a large amount of hydrogen. However, the containment design pressure is still decided without considering a large amount of hydrogen. This is a compromise that has existed since the containment was introduced to artificially minimize the exclusion distance or the radius of exclusion area (Okrent, 1981). We thought that the design pressure of the Mark X containment must be decided considering a large amount of hydrogen generation, because it is the true design basis of the containment, if we indeed respect the defense in-depth principle. Passive-safety reactors do not have any active safety systems that can treat fission products released from the containment. Passive-safety reactors usually do not have the secondary containment, either. Their primary containments stand alone in the atmosphere without being covered by the secondary containment. Therefore, FP can be released to the environment directly much easier than the current active-safety reactors, such as the ABWR, where the primary containment is completely covered with the reactor building (R/B). Therefore, the containment for passive-safety reactors should have enhanced capability to hold up FP in order to compensate for the lack of the secondary containment and the active FP treatment system. In order to enhance the containment capability to hold up FP passively, we thought that the Mark X containment should have double containment function to a certain extent even for noble gases and organic iodine. If a large amount of hydrogen generation is considered as one of the design basis of the containment, we need to increase the free volume of the containment

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to reduce the containment peak pressure like a large dry containment. If the free volume of the containment is increased so much we encounter a new problem, namely, hydrogen detonation. If the containment is not inerted and the concentration of oxygen is not controlled, there is some risk of hydrogen detonation. One of the protection systems for hydrogen detonation is an igniter. However, igniters are very uncertain devices to avoid hydrogen detonation. If ignition timing is wrong, they can rather cause hydrogen detonation. We thought that the Mark X containment should be inerted even if it has a large free volume in order to eliminate the risk of hydrogen detonation completely. Passive-safety reactors have walk-away safety during the grace period. During the grace period, the passive safety systems keep the plant in safe condition. After the grace period, however, they need to use active systems for accident management, such as replenishing the cooling water. We thought that the ASBWR+ should have permanent walk-away safety to reduce the necessity of accident management after the grace period. All the things stated above are safety enhancement. Normally, safety enhancement causes cost increase. It loses cost competitiveness of the ASBWR+ . Therefore, we thought the ASBWR+ should rather increase cost competitiveness even with the safety enhancement.

3. Reactor system and safety systems 3.1. Reactor system of the ASBWR+ In order to benchmark, we adopted the reactor system of the ESBWR for the advanced safer BWR at first. We assumed exactly the same core, reactor pressure vessel (RPV), and control rod drive (CRD) as the ESBWR. Namely, the plant output is 1380 MWe. The RPV height is 27.7 m. It is also a natural circulation reactor. Table 1 shows the comparison of design parameters of the reactor system among ABWR, ESBWR, and ASBWR+ . The plant output is very large as a passive-safety reactor. It is more than twice of the SBWR. The SBWR was only about 600 MWe. It caused a very severe condition for the containment design, because much more hydrogen is generated at severe accidents. The ESBWR adopted almost the same containment configuration that has larger

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Table 1 Comparison of reactor system parameters Parameters

ABWR

ESBWR

ASBWR+

Thermal power (MWt) Electric power (MWe) Vessel height (m) Vessel diameter (m) Number of fuel bundles Active fuel height (m) Power density (kw/l) Number of CRD

3926 1350 21.1 7.1 872 3.7 51 205

4000 1380 27.7 7.1 1020 3.0 54 121

4000 1380 27.7 7.1 1020 3.0 54 121

free volume (General Electric Company, 2002). The ASBWR+ adopted a completely different containment configuration to moderate the large plant output. The ESBWR and the ASBWR+ use a brand-new control rod (CR) that has wider blades. With this new CR we can reduce the number of CRD. 3.2. Safety systems of the ASBWR+ Safety systems of the ASBWR+ are based on the SBWR (Rao et al., 1991). Configuration of the safety systems is similar to those of the SBWR. Capacity of the safety systems, however, is the same as the ESBWR, because the plant output is 1380 MWe. Fig. 1 shows the safety systems of the ASBWR+ . The ASBWR+ has the isolation condenser (IC), the passive containment cooling system and the gravity-driven cooling system. The PCCS can cool the primary containment vessel (PCV) directly. The PCCS condensate is returned to the GDCS pool that is installed in the dry well. Therefore, the GDCS can continue to inject water into the reactor pressure vessel (RPV) up to depletion of the PCCS pool. The combination of the PCCS and the GDCS provides a recycling core cooling capability in the ASBWR+ . The GDCS pool is open to the DW atmosphere. Therefore, the backpressure of the DW is useful to inject cooling water to the RPV. On the contrary, the GDCS pool is installed in the isolated volume connected to the wet well (WW) air space in the case of the ESBWR. The PCCS condensate is returned to not to the GDCS pool but to the drain tank installed in the DW (General Electric Company, 2002). The drain tank water is then injected into the RPV by gravity. Therefore, the GDCS is once-through injection system and need not work as recycling core cooling system in the case of the ESBWR. After the once-through injec-

Fig. 1. Safety systems of the ASBWR+ .

tion, the GDCS free volume can work like the WW air space and mitigate containment pressure. In the case of ASBWR+ , containment peak pressure can be limited very low owing to its unique containment design. Therefore, we need not connect the GDCS volume to the WW air space in the case of the ASBWR+ . 4. Containment of the ASBWR+ 4.1. Main features Fig. 2 shows the containment of the ASBWR+ . We call it Mark X containment tentatively in the paper. The Mark X containment is a double containment. It has the primary containment vessel and the secondary containment vessel (SCV). The PCV consists of two compartments, the dry well and the wet well. The DW installs the reactor pressure vessel and the WW installs the suppression pool (SP). This PCV configuration is almost the same as the ESBWR containment. However, the design pressure of the PCV is much lower. This is because the WW air space is connected to the

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Fig. 2. Containment of the ASBWR+ (Mark X containment).

SCV through the isolation and connection switching system (ICSS). The ICSS can be rupture disks, vacuum breakers, or automatic valves. The ICSS is closed during normal operation and opened automatically, if the pressure difference through it reaches to the preset pressure difference. For transients and small LOCA, significant pressure increase in the WW air space does not occur. The ICSS does not open for these events. FP release is limited in the PCV and contamination of the SCV does not happen at these minor events. On the contrary, if a large LOCA or a severe accident occurs, pressure increase in the WW air space activates

the ICSS automatically. Non-condensable gases, such as nitrogen and hydrogen compressed in the WW air space can be released into the SCV. Therefore, the pressure increase of the PCV is limited very low. The SCV has about 50,000 m3 free volume. With this volume, you can accommodate a large amount of hydrogen generated at severe accidents. The PCV has about 15 vent pipes in order to moderate the overshoot pressure at a DBA LOCA. The SCV is made of steel in order to cool it by natural circulation of outside air. There is a dome over the operating floor. It is called as operating dome (OD). The operating dome is a key element of

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Table 2 Comparison of containments Characteristics

ABWR

ESBWR

ASBWR+

Containment type PCV design pressure (kPa) (psig) SCV design pressure (kPa) (psig) Free volume (m3 ) PCV outer diameter (m) PCV inner height (m) Building height (m) LOCA pressure (kPa) (psig) SA pressure (kPa) (psig)

RCCV 310 (45) – 13310 33.0 29.5 65.0 267 (39) 690 (100)

RCCV 310 (45) – 10600 36.5 34.6 69.4 207 (30) 517 (75)

Mark X 207 (30) 207 (30) 80000 36.5 34.6 69.4 69 (10) 103 (15)

the Mark X containment and provides marvelous merits that will be explained later in the paper. Both the PCV and the SCV are inerted with nitrogen in order to eliminate the risk of hydrogen detonation completely. Table 2 shows the comparison of containments among ABWR, ESBWR and ASBWR+ . The design differential pressure to activate the ICSS will be decided in actual design phase. For pressure transient analyses, we assumed 98 kPa tentatively. 4.2. Safety performance of the Mark X containment The SBWR had excellent safety performance originally owing to the passive safety systems. However, the plant output was only about 600 MWe. This capacity was not enough to accomplish a scale merit for plant capital cost. It also had a large reactor building surrounding the primary containment vessel. The PCV was not self-standing and connected to the R/B. This also caused cost impact. The ESBWR modified these points of the SBWR. The plant output is increased to 1380 MWe to pursue a scale merit. The large surrounding R/B is removed and upper portion of the PCV is self-standing. Based on the advanced light water reactor requirements, however, you have to consider a large amount of hydrogen generation at severe accidents. Safety performance of the PCCS for hydrogen rich containment atmosphere has been well studied (Yokobori et al., 1995; Auban et al., 2003). The large amount of hydrogen is released to the dry well from the RPV through the depressurization valves (DPV) or the vessel melt-through opening. Firstly, stratification in the DW might occur. The hydrogen tends to occupy the upper portion of the DW, because it is lighter than

nitrogen and steam. This can happen and remain, if the pressure of the DW does not increase. However, actually the pressure of the DW must increase due to heat-up or steam generation. The PCCS works using the pressure difference between the DW and the SP as a driving force. The intake of the PCCS tube is located at the upper portion of the DW. Therefore, the hydrogen stratified in the upper portion of the DW is pushed into the PCCS tube and vented to the suppression pool through the PCCS vent lines dedicated for non-condensable gas venting. The steam in the DW is also pushed into the PCCS tube and condensed therein. The condensed water is returned to the RPV via the GDCS pool. In the long-term cooling of the PCV, not LOCA vent pipes but only the PCCS vent lines are used as venting passes to the SP. This is because the submergence of the PCCS vent line is kept shallower than that of the LOCA vent line (Yokobori et al., 1995). Therefore, the steam in the DW is cooled by the PCCS and does not cause pressure increase so much. Unfortunately, however, the hydrogen is vented to the SP and compressed in the wet well air space. The hydrogen compressed in the WW air space causes a very high PCV pressure. The ESBWR adopted an efficient safety measure of containment overpressure protection system (COPS) to avoid the high peak pressure at severe accidents (General Electric Company, 2002). The content of the COPS is containment venting into the equipment room under the containment. If the pressure in the WW air space increases, the burst diaphragm on the WW wall opens and the hydrogen is released to the equipment room. The additional volume available for the COPS is the equipment room, where shutdown cooling and cleanup heat exchangers are installed. The free volume of the equipment room is about 15,000 m3

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and provides a pressure mitigation function at severe accidents without venting the containment atmosphere to the environment. The Mark X containment also has the safety measure of COPS. It has a large free volume of the secondary containment. Therefore, we can use the steel secondary containment vessel (SSCV) as the additional volume available for the COPS. The SSCV has more than 50,000 m3 free volume, including the operating dome. It is quite enough as the additional volume to mitigate the PCV pressure. If the pressure in the WW air space increases at severe accidents, the isolation and connection switching system on the WW wall, which are rupture disks, open and the hydrogen is released into the SSCV. Then, the ICSS on the operating dome, that is a vacuum breaker, also opens and the hydrogen in the SSCV flows into the operating dome. Hence, the Mark X containment can use all the free volume in the SSCV including the operating dome. The amount of hydrogen generated by 100% metal–water reaction of active fuel cladding is about 19,500 m3 at atmospheric pressure. The amount of nitrogen in the DW is about 6500 m3 . The total of 26,000 m3 of non-condensable gases is released into the 50,000 m3 free volume of the SSCV. If the passive containment cooling system is available, the containment pressure is determined mainly by the compression of the non-condensable gases. The estimated SSCV peak pressure is only about 98 kPa (1 kg/cm2 g) with the PCCS and only about 196 kPa (2 kg/cm2 g) without the PCCS. The SSCV is inerted with nitrogen, including the operating dome. Therefore, there is no risk of hydrogen detonation in the Mark X containment completely. The equipment room is not used as the COPS volume. If necessary, of course, it can be also used as an additional available volume for the COPS. The operating dome is very effective to minimize the time needed to inert and deinert the operating floor. The balance of the SSCV volume is inerted permanently and basically need not be deinerted even during refueling shutdown period. When a severe accident occurs, the Mark X containment is cooled by the PCCS at first. There is no pressure increase due to steam as long as the PCCS works. However, if hydrogen generation initiates, mixture of steam and hydrogen is supplied into the PCCS tubes. PCCS cooling performance is somewhat degraded by the existence of hydrogen. During this time phase the PCCS is mainly used for venting of the hydrogen from the DW

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to the SP through the PCCS vent pipes. Therefore, PCV pressure begins to increase due to compression of the hydrogen in the WW air space. In the case of conventional containment design, the WW air space volume is limited and the vented hydrogen must cause a big pressure increase of the PCV. After the ceasing of hydrogen generation from the molten core, the PCCS cooling performance is recovered and PCV pressure increase also ceases. After that time phase, PCV pressure stay almost constant (Yokobori et al., 1995; Auban et al., 2003). In the case of the Mark X containment, however, the WW air space can be connected to the large volume of the SSCV via the rapture disks and PCV pressure can stay much lower value. Almost all of the hydrogen is vented to the SP through the PCCS vent pipes owing to its buoyancy and steam driving force. LOCA vent pipes are hardly used for venting of the hydrogen and the steam because of their deeper submergence. Therefore, all the steam generated by the core debris cooling is cooled by the PCCS. However, the cooling water in the PCCS pool is not indefinite. After depletion of the PCCS pool, the steam is vented to the suppression pool through the PCCS vent pipes and the SP becomes saturated. After the saturation, a lot of steam flows into the WW air space and then into the SSCV volume through the opened ICSS. In this stage, pressure and temperature in the SSCV also increase. If the temperature of the SSCV wall increases enough, cooling by natural circulation of outside air initiates automatically. Normally, depletion of the PCCS pool occurs after 3 days of the initiation of a severe accident. Therefore, the decay heat becomes smaller. In this stage, cooling by natural circulation of outside air is enough for the containment vessel cooling. In the Mark X containment, some of the non-condensable gases are purged into the operating dome and trapped there. This can maximize the partial pressure of steam in the SSCV annulus and enhance efficient steam cooling. Moreover, the ICSS is installed very close to the lower wall of the SSCV. This can assure the steam getting close to the wall of the SSCV, where cool outside air is available. Furthermore, it is noticed that stratification will establish even in the SSCV annulus with more or less pure nitrogen below the ICSS, mainly steam above the ICSS and mainly hydrogen in the upper dome. Actually, the stratified hydrogen in the upper dome is more easily purged into the operating dome. Therefore, the SSCV wall cools the released steam very efficiently. Fig. 3 shows the pres-

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Fig. 3. Pressure transient at SA (TQUX sequence).

sure transient of the Mark X containment at a severe accident. TQUX sequence is analyzed as an example. It is a severe accident scenario, where a loss of feed water transient is followed by all loss of high-pressure ECCS and a failure of vessel depressurization. The amount of metal–water reaction resulted in 46% of the active fuel cladding. MAAP code is modified to address the ICSS and the operating dome. This is still a preliminary analysis, because the operating dome effect is not completely modeled, yet. The effect of stratification in the SSCV annulus is conservatively not considered, either. However, the result shows that the peak pressure is kept very low even after the PCCS pool depletion owing to the natural circulation cooling of outside air. There is a very important function of the operating dome for the CV cooling. After the saturation of the SP, steam flows into the SCV annulus continuously, and the pressure of the annulus increases. This pressure increase drives the steam into the operating dome. This steam flow carries the non-condensable gases, namely, hydrogen and nitrogen, also into the operating dome. As mentioned above, before steam generation complete stratification might be established with more hydrogen in the upper dome area, where the vacuum breaker on the top of the operating dome is closely located. The stratified hydrogen around the vacuum breaker tends to be easily and selectively purged into the operating dome. Therefore, the non-condensable gases are purged and trapped in the operating dome to some extent and steam mainly occupies the rest of the SSCV volume. This can provide very efficient continuous steam condensation on the SSCV wall. If there is much amount of non-condensable gases in the

rest of the SSCV, after the initial condensation of the steam, the vicinity of the wall is mainly occupied by non-condensable gases. After this situation is established, the non-condensable gases insist on being there and steam cannot get close to the wall. Therefore, the non-condensable gases work as an insulator. This noncondensable gas insulation effect deteriorates CV cooling very much. The operating dome can hold some of the non-condensable gases and can support efficient CV cooling by natural circulation of outside air. Without this operating dome effect, you cannot cool a CV by outside air efficiently. You always suffer from the non-condensable gas insulation effect and you have to assure mixing of the CV atmosphere sufficiently. Even if the mixing is enough, you can cool only the mixture of steam and non-condensable gases. Therefore, non-condensable gases always work for deteriorating cooling efficiency. It is uncertain that a large reactor can cool its containment vessel by natural circulation of outside air under the existence of a large amount of non-condensable gases. Small-scale experiments are not enough to verify the coolability if mixing of the containment atmosphere is a critical factor for the coolability, and there is no strong driving force for mixing of the atmosphere like a pressure difference between containment compartments. Normally, these experiments simulate plant condition only by keeping the ratio between the plant thermal output and effective surface area available for CV cooling. However, when you verify the sufficient mixing in the CV, you rather have to keep the ratio between the plant thermal output and the free volume of the CV. You cannot do the both things at the same time, because surface area and free volume stand on quite different scaling laws. You also have to consider the disturbing effects of obstacles existing in the CV, such as steam generators (SGs) and the crane to the mixing of steam and non-condensable gases. You have to combine independent experimental data and develop a computer code to analyze the mixing of gases and cooling of the CV. All these things cause uncertainty. The operating dome of the Mark X containment is effective to reduce the uncertainty. Moreover, the ICSS is installed very close to the lower wall of the SSCV. This can assure the steam getting close to the wall of the SSCV, where cool outside air is available. Therefore, the SSCV wall cools the released steam very efficiently. We need not depend on uncertain mixing phenomena in the containment.

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4.3. Double containment function against FP release The Mark X containment has double containment function against fission products release to some extent. Particulates and non-organic iodine can be scrubbed in the suppression pool and only very limited amount is released to the secondary containment vessel. The SCV further provides a double containment barrier for them. The leak rate of the SCV is very small as stated above. Therefore, the Mark X containment has an excellent safety performance to the soluble FP. The amount of soluble FP that can be released to the environment is very limited. In order to establish this safety level, the Mark X containment does not need any other safety features, such as a surrounding reactor building, the standby gas treatment system (SGTS), and emergency power supply. Noble gases and organic iodine are not soluble. Nearly, 100% of the non-soluble gases go into the SCV through the opened isolation and connection switching system. Namely, the primary containment vessel does not constitute a barrier for the non-soluble gases. Normally, release of the non-soluble gases from the core occurs within about 2 h of the initiation of a severe accident. Most of the release is finished at least within 1 day. Within 1 day, the containment is cooled by the passive containment cooling system. No steam is released to the SCV. Mainly hydrogen and nitrogen are released to the SCV. There is no significant pressure increase. Therefore, leak rate of the SCV is limited very low. However, after depletion of the PCCS pool, steam is released to the SCV and pressure increase close to the design pressure occurs. In this stage, leak rate up to the design leak rate need to be considered. However, the non-soluble gases existing in the annulus of the SCV is purged into the operating dome by the driving force of the generated steam. After being purged into the operating dome, the vacuum breaker on the wall of the operating dome prohibits the non-soluble gases from returning into the annulus of the SCV, thereby trapped and held up in the operating dome the nonsoluble gases cannot be released to the environment. With this mechanism, the Mark X containment has double containment function even for the non-soluble gases to some extent. More than factor of 2 can be expected for the retention function. If the discharge line of the ICSS on the operating dome is submerged into an

Fig. 4. Radiological dose evaluation of ASBWR+ .

available suppression pool, such as the buffer pool of the ESBWR, you can increase the driving force of the steam into the operating dome, thereby increasing the retention factor. If the containment vessel cooling by natural circulation of outside air is weaker, more steam exists in the annulus of the SCV. If more steam exists in the SCV, the pressure of the SCV increases and the leak rate also increases. However, if more steam exists in the SCV, more non-soluble gases are purged into the operating dome. Therefore, there is a negative feedback between the leak rate and the amount of non-soluble gases available for leakage. The higher the leak rate is, the less the amount of non-soluble gases. We can also use the alternate source term for the ASBWR+ (U.S. NRC, 1995, 2000). Based on the alternate source term, only 50% noble gases, 0.075% organic iodine and 2.5% non-organic iodine are in the SSCV and available for the release to the environment at most. Fig. 4 shows the result of preliminary dose evaluation for the ASBWR+ using the alternate source term and different operating dome transfer ratios (ODTR). Dose level of the ASBWR+ is the same order of magnitude as the advanced boiling water reactor (ABWR) and two orders magnitude smaller than the dose limits.

5. Permanent walk-away safety of the ASBWR+ As stated above, the ASBWR+ can cool the containment vessel permanently by natural circulation of outside air even after depletion of the passive containment cooling system pool. As for the core cooling, the suppression pool keeps the core covered using the equalizing line. The gravity-driven cooling system

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cannot cool the core after the GDCS pool depletion. After the SP is saturated steam is released to the secondary containment vessel and the SP water level also decreases. If the SP is depleted, it also loses the core cooling capability. However, the Mark X containment can condense the steam by natural circulation of outside air. The ASBWR+ has the condensate return system (CRS) in the annulus of the SCV. The CRS can keep the water level of the SP. Therefore, the combination of the air cooled Mark X containment, the CRS, and the SP equalizing line injection provides permanent walkaway safety for the ASBWR+ at severe accidents. The ASBWR+ also has the isolation condenser condensate return system (IC CRS). Normally, steam from the IC pool is vented to the atmosphere. You can switch it into the annulus of the SCV. Then, the steam is cooled and condensed by natural circulation of outside air. The condensate is returned to the IC pool by the IC CRS. You can use this operation mode for a prolonged station blackout (SBO). Therefore, the ASBWR+ also has permanent walk-away safety for SBO. The ASBWR+ has the PCCS and the GDCS. Before depletion of the PCCS pool the GDCS keeps the core covered and the PCCS cool the primary containment vessel directly. The steam generated in the PCCS pool is vented to the atmosphere. No SCV heat-up occurs at this stage. After depletion of the PCCS pool, steam is released to the SCV from the SP. The steam is condensed and returned to the SP. The SP water level is kept by the recycling cooling mode. This recycling cooling mode enables the ASBWR+ to have permanent walk-away safety. If the depressurization valves (DPV) fail to open and the IC valves also fail to open, the GDCS cannot inject water into the RPV and the IC cannot work, either. In this situation, even the ASBWR+ can suffer from severe core damage. In such a case, a large amount of hydrogen might be generated from the molten core and it can degrade the PCCS cooling performance for a certain time period. Even in such a severe accident scenario, the GDCS can flood the dry well about 1 m above the SP normal water level without any human intervention. For this purpose, we can use passive means to activate the corresponding squib valves when signaled by thermocouples installed in the DW. The core is uncovered in the RPV but the outside of the RPV is flooded completely up to about 1 m above the SP normal water level. Therefore, even in this time pe-

riod, the local DW temperature never reaches very high temperature that can threat the PCV integrity. After a vessel melt-through occurs, a large amount of steam generates due to debris cooling. The steam purges the hydrogen to the SP through the PCCS vent pipes. The PCCS cooling performance recovers. The PCCS initiates its safety function without any valve openings and continues its safety function by using only natural forces, namely, differential pressure and gravity. It is a complete passive safety system by itself. After the depletion of the PCCS pool, the ASBWR+ can cool the Mark X containment by natural circulation of outside air. This natural circulation cooling also uses only natural forces, gravity and buoyancy, and can continue forever. Therefore, the ASBWR+ can have permanent walk-away safety even for a severe accident scenario, where a large amount of hydrogen is generated.

6. Other probable variations 6.1. Forced recirculation reactor in the Mark X containment Instead of a natural circulation reactor, we can also use a forced recirculation reactor. We call this concept ASBWR II in the paper tentatively. The ASBWR II has internal recirculation pumps (RIP) and advanced locking piston CRD (ALPCRD). The plant concept is shown in Fig. 5. It can control the power by changing the core flow. On the contrary, the ASBWR+ is a natural circulation reactor and cannot control the power by changing the core flow. It always uses the control rods to control the power. The power is very sensitive to CR movement and fine motion of CR is necessary in the ASBWR+ . It uses the fine motion control rod drive (FMCRD). However, in the ASBWR II, we can control the power much more easily and frequently with the RIP. We do not need fine motion of the CR to control the thermal power. Therefore, we can use the advanced locking piston control rod drive (ALPCRD) instead of the FMCRD in the ASBWR II (Sato et al., 2004). With the ALPCRD we use the conventional CR, thereby we can avoid T&D for a brand-new CR. We need not rely on the natural circulation in the reactor pressure vessel any more, and we can reduce the RPV height. In the case of a natural circulation reactor, in order to keep enough core flow, we had to

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output also can be increased depending on coolability of the containment by natural circulation of outside air. We can also optimize the plant output of the ASBWR II in the future. The ASBWR II has a potential for plant output more than 1500 MWe. 6.2. ABWR in the Mark X containment

Fig. 5. Containment of the ASBWR II (forced recirculation reactor in the Mark X containment).

have much water head, and thus, a long RPV. Now we can reduce the RPV height. However, the RPV height is not changed right now, because the gravity-driven cooling system requires much coolant inventory in the RPV. The RPV height is 27.7 m. We can optimize the design and reduce the vessel height in the future. The ASBWR II also uses the Mark X containment as the containment. It has a lot of safety margin for the amount of hydrogen generation at severe accidents. Therefore, the amount of hydrogen is not a limiting factor of the plant output. Using the RIP, we can increase the plant output easily. We can also increase the fuel length to the conventional length of 3.7 m in the future. The plant

The advanced boiling water reactor is the latest nuclear power plant that is constructed and operated successfully in Japan. There are two operating units of the ABWR in Kashiwazaki–Kariwa site (Kobayashi et al., 2003). Both have 1356 MWe plant output. Several units are also under construction or expected to be constructed in the near future both in Japan and Taiwan (Yamazaki and Takahashi, 2003). The U.S. ABWR has acquired standard design certification from the U.S. NRC in 1997 (U.S. NRC, 1997). Therefore, it is the most reliable and probable option, when a nuclear renaissance emerges in the U.S. in the near future. It is a forced recirculation plant with internal recirculation pumps and also can use the Mark X containment as the containment. We call this concept ASBWR III in the paper tentatively. The plant concept is shown in Fig. 6. The ABWR originally has three-division active emergency core cooling system based on level 1 probabilistic safety assessment insights (Sato, 1992). The ASBWR III, however, also has the passive containment cooling system and the isolation condenser based on the safety philosophy of in-depth hybrid safety (Sato et al., 1999, 2004). Therefore, it is a half-passive-safety reactor. Using the Mark X containment the ASBWR III also can reduce the peak pressure of the containment at severe accidents below the design pressure. It also can have permanent walk-away safety for the containment vessel cooling. It also can eliminate the surrounding reactor building, because the Mark X containment is a double containment by itself. It is protected against external events including an airplane crash. Therefore, the Mark X containment has a potential to dramatically enhance safety and cost competitiveness of the current ABWR. Moreover, even for core cooling, we can provide walk-away safety for the ASBWR III. Using a slightly longer RPV, we can set the top of active fuel (TAF) level under the diaphragm floor. Therefore, if we can make up the lower dry well up to the diaphragm floor, we can also keep the TAF covered. Depressurization

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DBA LOCA. The ASBWR III can further enhance the excellent performance of the current ABWR passively. 6.3. Raised SP in the Mark X containment The raised suppression pool is a primary containment vessel concept that has the SP at the highest portion of the PCV. This PCV also can be installed in the Mark X containment. We call this plant concept as ASBWR IV in the paper tentatively. The plant concept is shown in Fig. 7. The dimensions shown in the figure are tentative and conservative. They can be smaller in final design stage. The raised SP was originally proposed for the future advanced boiling water reactor. However, the water management in the PCV is very difficult, because

Fig. 6. Containment of the ASBWR III (ABWR in the Mark X containment).

valves and equalizing line injection are also installed like other passive safety BWRs. In order to make up the lower DW, the active ECCS is used in the short-term. However, in the long-term, we use the gravity-driven core flooding system (GDCF) installed in the annulus of the secondary containment vessel. In the Mark X containment, the pressure difference between the dry well and the SCV is very small. It is less than 49 kPa (0.5 kg/cm2 ) in the long-term. Therefore, we can easily inject water into the DW from the SCV by gravity. Using the GDCF, the ASBWR III has permanent walkaway safety even for core cooling. The current ABWR has the safety performance of no core uncovery for the

Fig. 7. Containment of the ASBWR IV (raised SP in the Mark X containment).

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it has a very large volume of the lower dry well. In order to make up the lower DW it needs a large amount of water inside the PCV. In the conventional design of PCV, the SP is installed at the lowest position. This enabled that emergency core cooling system injection water could return to the SP by gravity and keep the SP water level. Owing to this passive coolant returning function, ECCS suction switching from the SP to the lower DW can be avoided in the conventional boiling water reactors (BWR). In the case of the raised SP, however, in order to establish this passive coolant returning function, the ECCS injection water must make up the DW up to the SP water level. However, the SP is not enough to provide that amount of water source. We cannot increase the SP water level, because the wet well air space also requires a large volume for hydrogen generation at severe accidents. If you try to inject water from outside of the PCV you need additional active systems, because the PCV pressure is very high. This causes more cost and lower reliability. Therefore, the raised SP is not adopted for the future ABWR yet. However, if it is installed in the Mark X containment, the water management problem can be solved very easily. In the Mark X containment, hydrogen can be released into the secondary containment vessel. Therefore, you need not a large volume of the WW air space. You can increase the water height of the SP and keep enough water. If you still need additional water, you can inject it from inside of the SCV by gravity. There is no pressure difference between the SCV and the SP in the Mark X containment. You can easily make up the SP. Therefore, if raised SP is installed in the Mark X containment it becomes quite feasible. There is another very important advantage of the raised SP. The water level of the SP can be kept higher than the TAF, because the SP is installed at the highest portion of the PCV. Therefore, you can make up the core using the SP very easily. The Mark X containment condenses the steam released from the SP and returns it to the SP. Using this mechanism, you can establish walk-away safety more firmly in ASBWR IV. It has also both active safety systems and passive safety systems based on the safety philosophy of IDHS (Sato et al., 1999, 2004). It has active high-pressure emergency core cooling system, namely, two high-pressure core flooders (HPCF) and two low-pressure flooders (LPFL). Therefore, it need not have a very long RPV. The active ECCS is much powerful to inject water into

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Fig. 8. Basic concept of in-depth hybrid safety in the ASBWR IV.

the RPV than the gravity-driven cooling system. With the HPCF available, you do not have to depressurize the vessel quickly. It also has reactor internal pumps (RIP). You can control thermal power by changing core flow with the RIP. You do not need fine motion of the control rods like a natural circulation reactor. Therefore, you can use the advanced locking piston control rod drive for cost reduction. The ECCS injects water in the short-term and the SP keeps the core covered in the long-term. This combination might rather provide more realistic and feasible plant concept than fully passive-safety reactors. The ASBWR IV has two-division configuration of active ECCS. The active ECCS copes with only the design basis accident. If all of the active safety systems fail, the passive safety systems back up. Fig. 8 shows the basic concept of in-depth hybrid safety in the ASBWR IV. The passive safety systems work independently of the active safety systems. Thus, the IDHS can provide in-depth safety. Fig. 9 shows the three-division configuration of the hybrid safety systems of the ASBWR IV. There is no reactor core isolation cooling system (RCIC), because

Fig. 9. Hybrid three-division configuration of the ASBWR IV safety system.

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the Isolation condenser is more reliable and inexpensive. The DBA LOCA is a pipe break accident of the HPCF injection line. Assuming a single failure of another HPCF, there is no high-pressure injection system. However, a slightly longer vessel of the ASBWR IV has enough coolant inventory, and can establish no core uncovery at the DBA LOCA. Elimination of the RCIC can bring about a certain cost saving. Two-division reactor coolant water system (RCW) and reactor sea water system (RSW) also contribute much cost saving. The very simple two-division active ECCS is suitable for being installed under the raised SP.

7. Conclusions Plant concept of the ASBWR+ was presented. It uses the Mark X containment. The Mark X containment is a double containment that has many advantages. The peak pressure at severe accidents is very low and can be kept below the design pressure even considering large amount of hydrogen generation. It can be cooled by natural circulation of outside air. The condensate in the steel secondary containment vessel can be returned to the suppression pool and utilized to cool the core. Owing to this cooling mechanism, the ASBWR+ can have permanent walk-away safety. The Mark X containment has double containment function against fission product release to some extent, even for noble gases and organic iodine. Owing to this double barrier function of the Mark X containment, the ASBWR+ can eliminate the surrounding reactor building without increasing FP release to the environment. The Mark X containment is also well protected against an airplane crash owing to the external event shield and the operation dome. The Mark X containment is inerted with nitrogen. There is absolutely no risk of hydrogen detonation. Although it is inerted, owing to the operation dome, inerting and deinerting of the operation floor area is easy. The design pressure of the PCV is very low and there is no stainless liner in the PCV. The SCV of the Mark X containment is made of steel. Therefore, no liner is necessary for the SSCV, either. The ASBWR II is a plant concept that has the reactor internal pumps and the advanced locking piston control rod drive. With the RIP, it can control the core much easier. The ALPCRD uses the conventional con-

trol rods. With the ALPCRD we can avoid research and development (R&D) for a brand-new CR. We can also optimize the RPV height, fuel length, and the plant output in the future. The Mark X containment also can be used for some other boiling water reactors. For example, it can be used for the current advanced boiling water reactor and socalled raised suppression pool. In both cases, the plant can have the same advantage of in-depth hybrid safety. These IDHS BWRs have some active safety systems. However, it can provide more realistic and more convenient plants. In these concepts, you need not depressurize the vessel so quickly every time an accident occurs. You can have a shorter vessel. You need not install the gravity-driven cooling system pool in the dry well, and so on. Although you still need emergency power supply, you can use it for the heating, ventilation and air conditioning (HVAC) of the control room and the standby gas treatment system for the fuel pool building. The Mark X containment is applicable to any BWRs that have no external recirculation loops. It can offer evolutional safety including permanent walkaway safety, double containment function, no hydrogen detonation, and very low peak pressure at SA while increasing cost competitiveness. It is a general containment that comes after the Mark III containment for the next generation BWRs.

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