Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. 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.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

Safety evaluation of the inherent and passive safety features of the smart design

SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integral PWR, which is under development at KAERI for seawater desalination and electricity generation. The conceptual design of the SMART desalination plant produces 40,000 m³/day of potable water and generates about 90 MW of electricity, which are assessed as sufficient for a population of about 100,000. The SMART enhances safety by adopting the inherent safety design features such as the elimination of large break loss of coolant accidents, substantially large negative moderator temperature coefficients, etc. In addition, the safety goals of the SMART are achieved through the adoption of passive engineered safety systems such as an emergency core cooling system, passive residual heat removal system, safeguard vessel, and reactor and containment overpressure protection systems. This paper describes the design concept of the major safety systems of the SMART and presents the results of the safety analyses using a MARS/SMR code for the major limiting accidents including transient behaviors due to desalination system disturbances. The analysis results employing conservative initial/boundary conditions and assumptions show that the safety systems of the SMART conceptual design adequately remove the core decay heat and mitigate the consequences of the limiting accidents, and thus secure the plant to a safe condition.     

Basic concept of a near future BWR

This paper shows a basic concept of a near future boiling water reactor (BWR) aiming at evolutional safety and cost savings with minimum change from the current advanced BWR (ABWR). The plant output is uprated to 1500 MWe from 1356 MWe. This power uprate can bring about potential of 11% cost saving per MWe base. Safety improvement as a next generation large reactor is also achieved.

The advanced reinforced concrete containment vessel (ARCCV) is used for the containment vessel to improve safety for severe accidents. The peak pressure of the containment at severe accidents can be kept close to the design pressure. The advanced passive containment cooling system (APCS) is also provided and can accomplish no primary containment vessel (PCV) venting.

The advanced emergency core cooling system (AECCS) consists of four divisions in the front line. The advanced passive cooling system (APCS) is also provided. The combination of the four divisional emergency core cooling system (ECCS) and the passive safety system improves the plant performance in probabilistic safety assessment (PSA).

This plant concept is designed based on the heritage of the current ABWR. No more major research and development (R&D) are necessary. Therefore, construction and operation is possible in the early 2010s.     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

IAEA activities on passive safety systems and overview of international development

Although a major focus of nuclear power reactor development efforts in industrialised countries is on large evolutionary units and design modifications that take maximum advantage of successful proven features and components, consideration is also given to utilisation of passive safety systems and inherent safety features. The various advanced reactor designs: evolutionary, large water-cooled reactor designs; evolutionary, medium size water-cooled reactor designs; concepts requiring substantial development; gas-cooled reactor concepts; and liquid metal-cooled fast reactors, incorporate a wide variety of passive safety features for initiation of safety systems, for residual heat removal and for containment heat removal. Organizations have established testing programmes to confirm their performance. A number of IAEA activities have lead to the conclusion that the use of passive safety features can be a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions. However, care should be taken to evaluate possible new failure mechanisms, and both passive and active systems should be assessed from the standpoint of reliability and economics. Key technical issues include: the quantification of reliability over a wide range of conditions: economics, speed of action, plant ageing, demonstration of technical feasibility, in-service testing, ease of maintenance and minimization of personnel radiation exposure. Many member states conduct substantial work on the design, modelling, development and reliability assessments of passive safety systems. Continued information exchange can benefit the involved member states, and the IAEA is providing a forum for review of programmes, project directions, and the results achieved.     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Issues affecting advanced passive light-water reactor safety analysis

Next generation commercial reactor designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved via safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet, the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs will necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing U.S. advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes may require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

On the failure modes of alternative containment designs following postulated core meltdown

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR representative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best estimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. These include the temperature of the non-condensibles escaping the ice condenser into the upper compartment, the performance of the pressure suppression system, the distribution of non-condensibles between compartments, and the degree and rate of combustion of hydrogen generated from metal-water reactions. For the PWR ice condenser case, results indicate that the containment would be breached by (i) steam overpressurization during the blowdown period (time less than 20 sec) if the ice condenser fails to perform its function, (ii) by overpressurization and thermal stress during the core melt period if 25% or more of the core zirconium reacts with water followed by hydrogen burning and, and (iii) by the overpressurization due to non-condensibles before containment floor penetration is completed. For the BWR Mark III case, similar conclusions can be drawn for the loss of vapor suppression, and for the hydrogen burning if the extent of zirconium-water reaction is more than 35% of the core inventory. If the hydrogen burning fails to materialize, the containment can retain its integrity until containment meltthrough provided the melting is confined to the reactor pedestal area. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded.     

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.     

Dynamic effects on structures and equipment due to Safety Relief Valve discharge loads

This paper presents analytical models, procedure, and results of a sensitivity study performed to investigate how the Safety Relief Valves (SRV) response of the steel containment of a Boiling Water Reactor (BWR) Mark III plant is influenced by the high frequency energy or noise associated with the idealized bubble forcing function. Various possible structural modifications to those plants already designed to have a steel containment are also presented and discussed with regard to minimizing the dynamic effects of SRV discharge loads.Special attention is given to the concept of filling concrete in the annulus between the steel containment and the shield building. The dynamic analyses of the containment structures with and without concrete in the annulus were performed and the results compared. Throughout the paper, several aspects of the SRV dynamic problem are emphasized and the relevant areas are identified for further investigations and studies.     

Simulation of hydrogen deflagration and detonation in a BWR reactor building

A systematic study was carried out to investigate the hydrogen behaviour in a BWR reactor building during a severe accident. BWR core contains a large amount of Zircaloy and the containment is relatively small. Because containment leakage cannot be totally excluded, hydrogen can build up in the reactor building, where the atmosphere is normal air. The objective of the work was to investigate, whether hydrogen can form flammable and detonable mixtures in the reactor building, evaluate the possibility of onset of detonation and assess the pressure loads under detonation conditions. The safety concern is, whether the hydrogen in the reactor building can detonate and whether the external detonation can jeopardize the containment integrity. The analysis indicated that the possibility of flame acceleration and deflagration-to-detonation transition (DDT) in the reactor building could not be ruled out in case of a 20 mm² leakage from the containment. The detonation analyses indicated that maximum pressure spike of about 7 MPa was observed in the reactor building room selected for the analysis.     

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC.

The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensable gas partial pressure in the suppression chamber, and low suppression pool temperature.

Calculations of other pipe breaks were also implemented for the reference plant adopting I/C as PCCS. The results show the effectiveness of the I/C cooling over a wide range of break spectra.     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Protection of the containment of the twin BWR KRB II plant against overpressure failure during severe accidents

The 2 × 1310 MWel plant KRB II as the newest design BWR plant in Germany (with prestressed concrete containment) had to go, with respect to the other BWRs, slightly different ways in solving the problem of containment overpressure protection during severe accidents. The basic concept of the plant and the special boundary conditions of the twin unit concept required modified solutions for the realized containment filtered venting system and the measures against energy release due to HZ reactions.The paper reports on the objectives, design, arrangement and installation of the equipment provided hereto. Special emphasis is layed upon the operation experience with the equipment installed. Furthermore, an outlook is given on possible complementary installations such as HZ-igniters and -recombiners and containment atmosphere monitoring and sampling.     

Analysis of SBLOCA on CPR1000 with a passive system

Since the Fukushima accident in 2011,more and more attention has been paid to nuclear reactor safety.A number of evolutionary passive systems have been developed to enhance the inherent safety of reactors.This paper presents a passive safety system applied on CPR1000,which is a traditional generation Ⅱ+ reactor.The passive components selected are as follows:(1) the reactor makeup tanks (RMTs);(2) the advanced accumulators (A-ACCs);(3) the passive emergency feedwater system (PEFS);(4)the passive depressurization system (PDS);(5) the incontainment refueling water storage tank (IRWST).The model of the coolant system and the passive systems was established by utilizing a system code (RELAP5/MOD3.3).The SBLOCA (small-break loss of coolant) was analyzed to test the passive safety systems.When the SBLOCA occurred,the RMTs were initiated.The water in the RMTs was then injected into the pressure vessel.The RMTs' low water level triggered the PDS,which depressurized the coolant system drastically.As the pressure of the coolant system decreased,the A-ACCs and the IRWST were put to work to prevent the uncovering of the core.The results show that,after the small-break loss-of-coolant accident,the passive systems can prevent uncovering of the core and guarantee the safety of the plant.     

Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste.The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology.The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s.International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.     

Reduced scale PWR passive safety system designing by genetic algorithms

This paper presents the concept of “Design by Genetic Algorithms (DbyGA)”, applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement.