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The development of passive design features for the Korean Next Generation Reactor
Nuclear Engineering and Design 201 (2000) 259 – 271 www.elsevier.com/locate/nucengdes
The development of passive design features for the Korean Next Generation Reactor Sung Jae Cho *, Byong Sup Kim, Myung Gi Kang, Han Gon Kim Center for Ad6anced Reactors De6elopment, Korea Electric Power Research Institute, 103 -16 Munji-Dong, Yusong-Gu, Taejon 305 -380, South Korea Accepted 3 April 2000
Abstract Four passive design features, such as a fluidic device, a passive secondary condensing system (PSCS), a passive cavity flooding system, and a passive hydrogen ignitor, are under development as part of the Korean Next Generation Reactor (KNGR), an advanced PWR, to increase its safety. The fluidic device, which is located at the discharge of the safety injection tank, is a system to inject the borated water into the reactor coolant system in a passively regulating way to elongate the allowable start-up time of emergency diesel generators and to enhance performance against the loss of coolant accidents. The PSCS, which supplements the auxiliary feedwater system, secures the heat removal through steam generators in the case of the loss of feedwater event. The PSCS takes inlet flow from the steam line and returns condensate into the feedwater line after condensation through condenser tubes. Fusible plugs have been adopted for passive cavity flooding. If the ambient temperature is high enough, the plugs between the in-containment refueling water storage tank (IRWST) and the reactor cavity melt so that the IRWST water starts to flow into the reactor cavity by gravity. Passive hydrogen ignitors (i.e. catalytic ignitors) have been adopted in addition to the active hydrogen ignitors to maintain containment hydrogen concentrations below a detectable limit of 10 volume percent. Applicability of these passive features on KNGR has been studied following the three-step approach of (1) preliminary analyses using computer codes and small-scale experimental facilities, (2) detailed analyses through large-scale tests, code running, and uncertainty validation, and (3) quality assurance by design verification and analyses on interfaces with other systems. © 2000 Elsevier Science S.A. All rights reserved.
1. Introduction
Extended and Updated Selected Paper from Post SMiRT 14th Seminar on Passive Safety Features in Nuclear Installations, Pisa, Italy, August 25–27, 1997 * Corresponding author. Tel.: + 82-42-865-5701; fax: +8242-865-5704. E-mail address: [email protected] (S.J. Cho).
Since the 1980s, many nuclear power plant vendors have developed advanced light water reactors (ALWR), which are safer and more economic than current nuclear power plants. The typical examples of ALWR are System 80+ by ABB-CE, ABWR by General Electric, and EPR by NPI, etc. In Korea, the Korean Next Genera-
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tion Reactor (KNGR) program, continued since 1992 to prepare for the domestic nuclear power demand expected early in the 21st century, probably after 2005, is on a line extended from the present Korea Standard Nuclear Power Plant (KSNPP) in an overall view. Nonetheless, KNGR aims to have many advanced and different characteristics from its predecessor, which are summarized in Table 1 (Choi et al., 1994a; Cho and Jerng, 1995; Choi and Kim, 1995). As shown in the table, KNGR has many advanced design goals and features. Some important design goals for KNGR are as follows. The design lifetime of KNGR should be 60 years. Core damage frequency (CDF) and containment failure frequency (CFF) of KNGR should be less than 10 − 5 per RY and 10 − 6 per RY, respectively. The thermal margin of KNGR should be greater than 10%. The target of thermal margin is 15%. The economic goal of KNGR is to secure around 20% cost advantage over competing energy sources; for example, coal-fired power generation. Major advanced design features of KNGR are as follows. Double cylindrical containment. Four-train emergency core cooling system that is injected to the reactor vessel directly. Fluidic device in safety injection tank in order to regulate flow rate passively. In-containment refueling water storage tank (IRWST). Main control room designed under the human factors engineering principles and digital I&C technique. The electric power of KNGR is 1350 MW, which is about 30% higher than KSNPP. The design simplification in the emergency core cooling system and auxiliary feedwater system would be an example of the design improvements. KNGR aims at both enhanced safety and economic competence (Choi et al., 1994a). The philosophy of enhancing safety for an advanced reactor is based upon three levels of safety. The first level comes from designing a
system with a high degree of reliability. The sound design with sufficient safety margins and strict quality assurance program in engineering and design will ensure this first level of safety. Despite the assurance offered by careful system design and engineering, it is necessary to anticipate that some incidents and malfunctions will occur during the service life of the system. The second level of safety is to provide means that will forestall or cope with such events. Protection systems are designed to prevent, arrest, or accommodate safely a range of conceivable abnormal situations. The final level of safety is based upon the view that it is prudent to go beyond the levels of safety already described by designing engineered safety features to prevent and mitigate a set of designbasis and severe accidents even though they are considered highly unlikely. The implementation of safety enhancement could be achieved by employing passive safety features in each level of safety described. Passive features eliminate the need for operator involvement at abnormal plant situations including accident conditions by solely relying on natural forces like gravity and natural convection. For example, the core cooling capability in emergency situations could be highly upgraded by automatic delivery of water to cool the core through gravity flow introduction from a tank located above the core, and natural circulation of coolant induced by the temperature difference through heat exchangers that are connected to the steam generator. Passive safety features sometimes reduce plant complexity, but in most cases accompany economic penalty. The dependence of safety enhancement solely on passive features most likely asks for a big economic sacrifice. One must realize that the design objectives include better economics as well as enhanced safety in developing advanced concepts, which means that safety and economics are inter-related rather than independent. Hence, passive features should be adopted taking into account the state of technology and the economics of improvements. As long as the expected safety is within the acceptable level, further enhancement of safety should be based upon cost–benefit studies.
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Table 1 Summary of top-tier requirements for KNGR Item (a) General Reactor type Design life Design Philosophy (b) Safety Safety goal
Thermal margin Operator action time Station blackout coping time LOCA protection (c) Design bases Design criteria Source term Seismic analysis
Requirement
Comparison with current plants
Evolutionary PWR with 4000 MWH 60 years Simplification, design margin, human factor engineering, proven technology
– 40 years –
Core damage frequency, 10−5 per RY Containment failure frequency, 10−6 per RY Radioactive release limit targets: 1 rem/24 h at site boundary; Cs-137 100 TBq at site boundary; probability exceeding above limits shall be less than 10−6 per RY 10–15% \30 min
CDF10−4 per RY CFF10−5 per RY
Coping time, 8 h minimum; installation of AAC
4h
No fuel damage to 15 cm diameter break
–
Licensing design basis (LDB), safety margin basis (SMB) Use the physically based source term Eliminate OBE seismic analysis
LDB only TID-14844 OBE analysis
(d) Se6ere accident mitigation system Combustible gas Hydrogen concentration lower than 10% for 100% clad oxidation control Direct containment Automatically depressurize the core to prevent high pressure core heating control melt; provide the reactor cavity and cooling means Containment Maintain ASME code service level or equivalent criteria for 24 h performance criteria Emergency planning Provide technical basis for reduction of emergency planning boundary (e) System features NSSS
ECCS RHRS Containment I&C and control room design Spent fuel storage (f) Performance Availability goal Load follow capability
Hot leg temperature lower than current design; pressurizer large enough to eliminate PORVs; SG: In-690 and 10% plugging margin for tubes, minimum 30 min dry-out time; depressurization system: safety grade system for depressurization and feed-bleed operation Direct vessel injection; RWST, inside containment; four trains without interconnection Higher ultimate rupture strength than RCS operating pressure; mid-loop operation improvement Cylindrical double concrete Use digital I&C technology and HFE in top-down scheme
8% 10 min
No provision No provision No provision
EP is required
S/G: In-600 and 2–80% plugging
Cold leg injection; RWST outside; two trains – – Analogue technology
10 years+1 core; 20 years using advanced technology
10 years+1 core
Availability, 90%; unplanned trip, 0.8 per RY Daily load follow with frequency control
80–87% in Korea –
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Table 1 (Continued) Item
Requirement
Comparison with current plants
Load rejection capability Refueling cycle Low-level radwaste Occupational radiation exposure
100% load rejection without reactor trip
–
1824 months Solid, 50 m3 per year; liquid, 0.05 Ci per year; gas: 200 Ci per year Accumulated exposure, 100 man-rem per year; individual exposure, 2 rem per year for 5 year average
12–18 months – Accumulated exposure, 200 man-rem per year
(g) Economic Life time cost goal 10-Year cost goal
20% advantage over coal 10% advantage over coal
– –
(11) Site en6elope parameters Seismic design 0.3g Others Seawater temperature, 35.5°C for safety systems, 32.1°C for non-safety systems; ambient air temperature, 33.3°C maximum and −10.6°C minimum Reliability assurance Develop the program to assure the consistency of risk-significant program structures, systems, and components with design bases Configuration Develop computerized database to manage design process management systematically program
In this paper, we will introduce the passive safety features that are incorporated in KNGR at present and the new development of passive safety features for the future application. Through the main body of the paper there is (1) a summary of the Phase I study to select passive features for KNGR, (2) passive safety features for application to KNGR, and (3) future passive design features for the application will be described focusing on the functional requirements, systematic layout, major analyses, and detailed design procedures
2. Passive design features for KNGR As shown in Table 3, the design stages of KNGR are divided into three stages, Phase I, Phase II, and Phase III. In the Phase I development stage, we finished the conceptual design of KNGR. At that time, we set up the basic design requirements of KNGR as shown in Table 1. During the Phase II stage, we developed basic design and detailed design requirements of KNGR. In Phase III, we will proceed with the detailed design and will obtain the design certification to construct KNGR.
0.2 g –
No provision No provision
During the early stage of Phase I development, we considered many passive design features for KNGR such as the following (Choi et al., 1994b). Passive secondary condensing system (PSCS) (Abdollahian et al., 1994). Fuel cells as alternate on-site power supply. Passive external containment cooling system (PCCS) (Kim et al., 1998). Catalytic hydrogen ignitor. Passive containment spray system. Containment overpressure protection system. Passive cooling for specific components. Passive spent fuel pool cooling system. Passive quenching system. Permanent supplemental interfaces. Passive safety injection system. Passive primary residual heat removal system. Passive cavity flooding system (Jae et al., 1993). Emergency boration system. Passive pressurizer spray control system. Ceramic core catcher. Fluidic device in safety injection tank to regulate flow rate (Sugizaki et al., 1992). Flooding exterior of reactor vessel. The above mentioned features have been studied to be adopted in the KNGR design on the
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basis of the screening criteria shown in Table 2. At first, each item went through the technical review in aspects of simplification, integrated plant impact, licensing impact and/or the need of testing. As a result of this technical review, ten items were selected. We then reviewed the safety and economics of each item concerning safety benefits, the overall core melt and/or radiation release frequency not being increased by the passive safety feature. In relation to economics, the passive safety features should not cause a significant increase in overnight plant cost and in overall construction periods. Finally, we performed the detailed cost – benefit analysis to select the final candidate. As a result of the Phase I study, four passive features were retained for application to KNGR Table 2 Screening criteria for passive features Category
Screening criteria
Cost
No significant increase in overnight plant cost No increase in overall construction schedule Essentially no reduction in plant output No fundamental changes to existing licensing principles or criteria No concepts previously rejected by licensing authorities Development/testing required to fit in overall schedule No significant development cost No fundamental R&D required (i.e. proof of principle) No construction of new test facilities required No fundamental methods development No increase in safety risk (core melt or radiation release) No unquantifiable impact on plant design, construction, operation, and maintenance No negative impact on the fundamental reference design configuration No significant decrease in ‘…abilities’a
Licensing impact
Development and testing
Safety benefits Integrated plant impact
Operating and maintenance
No perceived increase in complexity No significant increase in O&M costs a
ity.
Constructability, operability, reliability and maintainabil-
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design to enhance the safety margin (Choi et al., 1994b). Passive secondary condensing system (PSCS). Fluidic device in safety injection tank to regulate flow rate. Passive cavity flooding system. Catalytic hydrogen ignitor. Another passive feature (i.e. PCCS) is under application to KNGR design. Its adoption to KNGR has not yet been determined. Applicability of these passive features on KNGR has been assessed following a three-step approach. 1. Preliminary analyses using computer codes and small-scale experimental facilities. 2. Detailed analyses through large-scale tests, code calculation and uncertainty validation. 3. Quality assurance by design verification and analyses on interface with other systems. The major design activities through the design stages and the major performance analyses for design implementation are shown in Tables 3 and 4, respectively. Details for the passive features and their current design status are presented in the next section.
3. Current design status of the passive design features for KNGR
3.1. Fluidic de6ice The fluidic device is located at the discharge of the safety injection tank (SIT) and controls the injection flow of borated water into the reactor coolant system (RCS) in a passively regulating way, as shown in Fig. 1. It consists of a stand pipe, a control port, and the main body. The borated water in the SIT can be injected into the reactor vessel upper plenum directly following a two-step process. First, the main stream flows through the stand pipe. Once it gets a switch point, the flow through the stand pipe is terminated, and then the injection flow continues through the control port only. The major function of the control port is to generate a vortex flow in the body to make flow resistance, such that the injection flow rate can be passively regulated by
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Table 3 Major design activities through the design stages Passive features
Design stage
Remarks
Phase I (1992–1994)
Phase II (1995–1999)
Phase III (1999–2001)
Passive secondary condensing system
Conceptual design
Detailed design, design certification
Fluidic device in SIT
Conceptual design, preliminary analysis Conceptual design
Preliminary analysis, performance analysis, small-scale test, large-scale test Performance analysis, performance test Performance analysis
Passive cavity flooding system (fusible plug) Catalytic hydrogen ignitor Passive containment condensing system
Conceptual design –
Detailed design, design certification Material selection, detailed design, design certification Performance analysis design implementation, design certification Performance analysis and Performance analysis, test implementation planning
the vortex flow intensity. The system will be capable of reducing the discharge flow to 10% of the maximum flow rate. This system is expected to help extend the allowable start-up time of emergency diesel generators and enhance loss of coolant accident (LOCA) cooling performances (Choi et al., 1994a). After large break loss of coolant accident (LBLOCA), the safety injection system (SIS) provides a big flow rate to the reactor to reflood the core in a short period to meet 10CFR50.46 peak cladding temperature acceptance criteria. SIS in KNGR consists of four high-pressure injection pumps (HPSIPs) and four safety injection tanks. Low-pressure injection pumps, which have much larger capacity compared with HPSIPs, are removed by adopting the direct vessel injection method. At the early stage of LBLOCA, i.e. the blowdown phase, a large flow rate from SIT is required to fill the reactor core lower plenum in order to meet the acceptance criteria of 10CFR50.46. After the downcomer is filled with the injected water, most water injected from the SIT is spilled through the broken pipe into the bottom of the containment system. By adopting the fluidic device in SIT, the injection mode from the SIT can be divided into two steps, i.e. a large amount before the downcomer is reflooded and a relatively small amount after that time.
Beyond design basis events
Design basis events Severe accident
Severe accident Future application
The important factors to design the fluidic device are the change rate of the flow amount and the time to change the flow rate. To decide these factors, sensitivity analyses have been performed for KNGR using CEFLASH-4A and COMPERC-II codes. These are design codes for large break LOCA developed by ABB Combustion Engineering Corporation. CEFLASH-4A code simulates blowdown phenomena of the early stage of LBLOCA, and COMPERC-II code simulates reflooding phenomena of the late stage of the accident. The major design parameters of KNGR are presented in Table 5. As a result of preliminary analyses, the time to change the flow rate is set to 36 s after the initiation of LBLOCA, and the minimum required flow rate by SIT after that time is 12% of maximum flow. Fig. 2 shows the SIT injection flow rate with respect to two change rates, i.e. 20 and 30%. In cases of 20 and 30% flow rates, the injected water is maintained until 198 and 140 s after the initiation of accident, respectively. Without a fluidic device, the SIT would be empty 72 s after the initiation of accident. This means that we can use SIT water for 126 s more than typical SIT in the case of 20% flow rate. Therefore, safety injection pumps are not required until 198 s after the initiation of the accident when the fluidic device, which has a 20% flow rate changing capability,
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and the start of the emergency diesel generator could then be prolonged until that time. The emergency diesel generator starting time, however, should be determined carefully because it is used as the power supply under many other accident conditions. During Phase II, we will complete the experiments to determine the geometric specification and the performance of the fluidic device in order to install it in safety injection tanks. The scales of the
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experiment are set to 1:30 and 1:1 with respect to the flow rate and operating pressure, respectively. The major parameters for the experiments are presented in Table 6, and the schematic diagram of the experimental facility is shown in Fig. 3. As shown in the figure, a compressor will pressurize SIT before the experiment. The fluidic device is located outside SIT so that the height of standpipe can be changed.
Table 4 Major performance analyses for design implementation Passive features
Computer code analysis
Experiment
Passive condensing secondary system
Steady-state natural circulation Flow stability Noncondensable gas effect PSCS valves opening/closing set point System pressure loss Analysis of over-cooling of primary side Determination of condenser capacity Alarm setpoint Water tank capacity evaluation I&C installation location Transient analysis Malfunctioning effect
Pressure loss through the condenser Non-condensable gas effect Transient analysis Heat transfer coefficient CHF initiation verification Thermal stratification in the water tank
Fluidic device in SIT
CFD analysis in fluidic device Safety injection capability Analysis on nitrogen gas entrainment Malfunctioning effect
Effect of the stand pipe elevation Stand pipe installation angle and location Flow resistance coefficient K Probability of nitrogen gas entrainment Flow rate change during service period
Passive cavity flooding system (fusible plug)
Analysis on flooding capability Flow area due to plug melting Malfunctioning effect
Fusible metal melting temperature Metal properties Plug geometry effect
Catalytic hydrogen ignitor
Ignition capability under accident condition Optimization in ignitor number and location Malfunctioning effect
Thermal aging resistance Test under accident conditions Functional test Ignition capability Vibration resistance Impurities effect
Passive containment condensing system
Steady state natural circulation Heat removal capability from containment System pressure loss Determination of condenser capacity Water tank Capacity evaluation I&C installation location Transient analysis
Pressure loss through the system Flow stability Transient analysis Heat transfer coefficient Thermal stratification in the water tank
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Fig. 1. Schematic diagram of the fluidic device in the SIT.
3.2. Passi6e secondary condensing system One of the most effective ways to improve current pressurized water reactor safety is to adopt a passive decay heat removal system for the purpose of long-term core cooling without any external power supply and no operator action. The PSCS, which supplements the auxiliary feedwater system (AFWS), is to secure the heat removal through the steam generators (S/Gs) when AFWS has failed, especially the total loss of feedwater events. The PSCS is installed at the outside of the containment as shown in Fig. 4 (Choi et al., 1994a). The current version of the design requirement for the PSCS requires 8 h of mission time (i.e. it can remove decay heat through PSCS condenser tubes to the secondary water storage tanks without any additional water supply for 8 h). The normal operating condition for the inside of the condenser tubes is saturated
steam for the full operating pressure and temperature of the secondary side The decay heat from the reactor core is transferred to the S/Gs, which transfer heat to the naturally circulating water to make it boil. The Table 5 The major design parameters of KNGR Design parameters
Initial value
RCS Pressure (psia) Thermal power (102% of nominal power) (MW) Peak linear heat generation rate (kw ft−1) Hot leg temperature (F) Cold leg temperature (F) RCS flow rate (lb h−1) Number of safety injection pumps Number of SITs Containment net free volume (ft3) SIT gas/water volume (ft3)
2250 4063 13.94 615 555 166.6×106 4 4 3.3×106 800/1600
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Fig. 2. Required/supplied flow rate during reflooding after LBLOCA. Table 6 The major parameters for the fluidic device experiment Parameters
Maximum
Minimum
Radius of stand pipe (mm) Length of stand pipe (m) Gas pressure in SIT (bar g) Length of SIT (m) Range of flow rate (m3 h−1)
100 8 50 14 20–100
25 4 40 14 10–50
steam discharged from the S/Gs goes into the PSCS condensers, and is condensed on the inside of the tubes due to heat transfer to the water in the secondary water storage tank by the nucleate pool boiling. The condensate is then returned to the S/G via a feedwater line. When the PSCS flow is driven by natural circulation only, its decay heat removal capability is determined mainly by the natural circulation flow rate of the PSCS. The natural circulation flow rate in the PSCS, on the other hand, depends primarily on the buoyancy effects induced by the temperature differences between the cold and the hot regions, and the vertical elevations of the PSCS. Since the vertical elevation is limited by the cost and constructability, the major remaining
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and effective way to increase the heat removal capability of PSCS depends on the heat removal capacity of the condensers. In KNGR, PSCS has two condensers per steam generator. In the Phase I study, the capacity of PSCS was determined to 111 MW per steam generator. That capacity is equivalent to decay heat 5 min after the reactor trip. In Phase II, we modified the capacity of PSCS to 80 MW. This capacity is equivalent to mitigate the most limiting accident, i.e. total loss of feedwater flow accident. We assumed that PSCS is not used for pipe break accidents such as steam line break or feedwater line break. The major performance analyses for design implementation of PSCS are pressure loss, heat transfer coefficient, flow stability, and transient response, as summarized in Table 4. Throughout the Phase II design process, the (1) preliminary system analyses by numerical code, (2) small-scale test (1/480 volume scale, full pressure), (3) detailed system analysis and (4) full-scale, full-pressure test will be executed to ensure its operability and obtain proper design data for KNGR. Currently, RELAP5/Mod3.2 code is used for system performance analysis. The heat transfer coefficient between condenser inside and outside, which is used for code analysis, has been obtained by the results of a single tube test. According to the preliminary probabilistic safety analysis (PSA) on PSCS, it can reduce CDF by about 50%.
3.3. Passi6e ca6ity flooding system KNGR takes the post-flooding process to mitigate the impact of a severe accident on the system and to maintain containment integrity. This flooding process consists of two methods. One is active flooding by motor-operated valves, and the other is passive flooding by fusible metal plugs located inside the pipe between the IRWST and the reactor cavity. During the passive cavity process (i.e. in conditions of loss of active systems), once molten corium falls down through the broken crevice of the reactor vessel to the bottom of the reactor cavity in case of severe accident, the ambient temperature of the cavity increases until the melting point of the metal plugs is reached. If end plug melts, IRWST will supply cooling water to
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the reactor cavity through the pathway by gravity. As shown in Fig. 5, the reactor cavity and IRWST are directly connected to each other. The
end plug sides are normally closed by the metal plugs. The cavity and IRWST can be isolated by valves to maintain the end plugs.
Fig. 3. Schematic diagram of the fluidic device experiment.
Fig. 4. Schematic diagram of the passive secondary condensing system.
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Fig. 5. Schematic diagram of the IRWST and cavity flooding system.
3.4. Passi6e hydrogen ignitor Hydrogen ignitors have been adopted in addition to active hydrogen recombiners to maintain hydrogen concentrations below a detectable limit of 10 volume percent. Hydrogen ignitors consist of active and passive systems. For the active systems, glow plug ignitors that are manually controlled in the main control room will be adopted. For the passive system, catalytic ignitors will be adopted to mitigate severe accident impact on the primary side components. Due to the proven design of the catalytic ignitors, the hydrogen ignition system is expected to minimize the threat of containment failure caused by large hydrogen deflagrations or hydrogen detonations. The catalytic hydrogen ignitor burns hydrogen passively using the noble metals under special conditions. It consists of a catalyst, ignition wires, and a tightly closed metal container as shown in Fig. 6. Once a severe accident occurs, temperature in the containment increases, and then the cover of the hydrogen ignitor opens automatically. After this opening, gases containing hydrogen enter into the metal container. In the container, a catalyst reacts with hydrogen and generates high temperature to activate hydrogen ignition element. Then, hydrogen in the containment is burned by the ignition element.
To adopt this on KNGR, some tests and analyses must be executed through the whole design phases. The major points are functional tests under severe accident conditions and ignition capability. Besides the catalytic hydrogen ignitor, a passive auto-catalytic recombiner (PAR) is con-
Fig. 6. Schematic diagram of the catalytic ignitor.
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Fig. 7. Schematic diagram of the PCCS.
sidered in KNGR because a PAR does not require any operator action in order to actuate it, and the ignitor can impact upon the integrity of containment or equipment near the ignitor.
4. Passive design feature for future application The passive external containment cooling system is the preferred means of decay heat removal following a LOCA and is under development for the future application. The PCCS (shown in Fig. 7), similar to the passive containment cooling system in the simplified boiling water reactor, provides a way of external containment cooling through a natural circulation circuit. It consists of two heat exchangers, relevant lines, and water tanks. Heat transferred from the containment atmosphere to the coolant through the primary heat exchanger tube, which is in the containment, is removed by the condenser tubes to the water tank, which is outside the containment. The condensate then returns to the containment. The PCCS condensers are installed in the external water storage tank at atmospheric pressure. They
cool the steam–air mixture from the containment atmosphere by cold water in the PCCS tank via heat transfer through condensing tubes, and return the condensate into the containment. The major points of tests and analyses will be focused on the heat transfer coefficient of the heat exchangers and flow stability through the natural circulation circuit. The key physical phenomena of the PCCS are as follows (Kim et al., 1998). Condensation in the presence of noncondensable gases. Natural (buoyancy induced) convection and mixing. Stratification in liquid or gas volume. In order to estimate the performance of PCCS, the computer codes should have the capability of simulating the presented phenomena. Therefore, we reviewed some available computer codes, i.e. GOTHIC version 4.1c, GASFLOW version 1.0, etc. Based on the review results, GOTHIC code has been selected to analyze the performance of PCCS in KNGR. We are now performing sensitivity analyses using GOTHIC code for mass and energy release data for KNGR. According to the preliminary analysis results, PCCS has a theoretical possibility of meeting the design criteria. However, to design PCCS as a back-up system for the containment spray system, it is estimated that more than 14 units are required to meet DBA design criteria for the pressure at 24 h after large break LOCA. This would require more than 7000 m2 of evaporator heat exchanger surface in containment, which would introduce serious and perhaps insurmountable equipment arrangement problems. Therefore, more study is required to optimize the size and capacity of the PCCS. Although the study of PCCS design is focused on the case of the design basis accident at present, we will extend it to the case of severe accidents if the results of the present work are positive.
5. Conclusion The approach of developing advanced nuclear systems in Korea is with the continuing goals of enhanced safety and better economics. The safety
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enhancement of the nuclear systems can be achieved by employing passive safety features because of their very high reliability. However, they should be discretely adopted to avoid a big economic sacrifice that might impact on the better economics. Through the inter-related safety and economic study for many passive design features during the Phase I works, four major passive features (i.e. fluidic device, PSCS, passive cavity flooding system, and passive hydrogen ignitor) have been selected for application to KNGR design. Based upon Phase I study results, the following conclusions can be obtained for the current and future design applications. 1. Fluidic device located at the discharge of the SIT injects the borated water into RCS in a passively regulating ways. According to the analysis results, fluidic device will lengthen the water injection by more than 100 s. After demonstration experiments, it will be installed in SIT for KNGR. 2. PSCS using condensers submerged in the PSCS tank cools S/Gs by natural circulation. Until now, we have determined overall configuration and capacity. According to PSA results, the core damage frequency is reduced by about 50% through PSCS. 3. Fusible plugs between reactor cavity and IRWST supply cooling water from IRWST to the reactor cavity by melting and passive hydrogen ignitors using catalytic ignitors. 4. For future safety enhancement, there is one major R&D area of adopting passive safety feature: the PCCS.
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5. R&D for adopting passive safety features is a continuing process for KNGR. It consists of conceptual design study, actual design work, and experimental back-up. References Abdollahian, D., Johnson, C.B., Yedidia, J., 1994. Performance of the SBWR isolation condenser for overpressure protection. Transactions on American Nuclear Society (Trans. ANS), p, 701. Cho, S.J., Jerng, D.W., 1995. Research activities and design requirement for the next generation reactor in Korea. The ZAEA International Workshop on Future LWR, Tokyo, Japan, July 1995. Choi, Y.S., Kim, B.S., 1995. Progress in design, research & development and testing of safety systems for the Korean next generation reactor. IAEA Technical Committee Meeting on the Progress of Water Cooled Reactor Design, Piacenza, Italy, May 1995. Choi, Y.S. et al., 1994a. Research & development on next generation reactor (phase 1) — design concepts of next generation reactor, KRC-92N-J11. Korea Electric Power Corporation, Seoul, Korea, December. Choi, Y.S. et al., 1994b. Research & development on next generation reactor (phase 1) — a feasibility study for incorporating passive safety features into large evolutionary LWR, KRC-92N-J11. Korea Electric Power Corporation, Seoul, Korea, December. Jae, M., Milici, A., Kastenberg, W.E., Apostolakis, G., 1993. Evaluation of accident management strategies for PWRs: cavity flooding. PSA International Topical Meeting, Florida, USA, 26 – 29 January 1993. Kim, Y.H., Todreas, N.E., Driscoll, M.J., 1998. Distributed parameter modeling of the KNGR containment using GOTHIC, MIT-ANP-TR-059. MIT (internal report), February 1998. Sugizaki, T., Matsuoka, T., Shiraishi, T., Nakahara, Y., Kawai, K., Makahara, Y., Okabe, K., 1992. Design study for a passive safeguards system. 5th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-5).
The development of passive design features for the Korean Next Generation Reactor Sung Jae Cho *, Byong Sup Kim, Myung Gi Kang, Han Gon Kim Center for Ad6anced Reactors De6elopment, Korea Electric Power Research Institute, 103 -16 Munji-Dong, Yusong-Gu, Taejon 305 -380, South Korea Accepted 3 April 2000
Abstract Four passive design features, such as a fluidic device, a passive secondary condensing system (PSCS), a passive cavity flooding system, and a passive hydrogen ignitor, are under development as part of the Korean Next Generation Reactor (KNGR), an advanced PWR, to increase its safety. The fluidic device, which is located at the discharge of the safety injection tank, is a system to inject the borated water into the reactor coolant system in a passively regulating way to elongate the allowable start-up time of emergency diesel generators and to enhance performance against the loss of coolant accidents. The PSCS, which supplements the auxiliary feedwater system, secures the heat removal through steam generators in the case of the loss of feedwater event. The PSCS takes inlet flow from the steam line and returns condensate into the feedwater line after condensation through condenser tubes. Fusible plugs have been adopted for passive cavity flooding. If the ambient temperature is high enough, the plugs between the in-containment refueling water storage tank (IRWST) and the reactor cavity melt so that the IRWST water starts to flow into the reactor cavity by gravity. Passive hydrogen ignitors (i.e. catalytic ignitors) have been adopted in addition to the active hydrogen ignitors to maintain containment hydrogen concentrations below a detectable limit of 10 volume percent. Applicability of these passive features on KNGR has been studied following the three-step approach of (1) preliminary analyses using computer codes and small-scale experimental facilities, (2) detailed analyses through large-scale tests, code running, and uncertainty validation, and (3) quality assurance by design verification and analyses on interfaces with other systems. © 2000 Elsevier Science S.A. All rights reserved.
1. Introduction
Extended and Updated Selected Paper from Post SMiRT 14th Seminar on Passive Safety Features in Nuclear Installations, Pisa, Italy, August 25–27, 1997 * Corresponding author. Tel.: + 82-42-865-5701; fax: +8242-865-5704. E-mail address: [email protected] (S.J. Cho).
Since the 1980s, many nuclear power plant vendors have developed advanced light water reactors (ALWR), which are safer and more economic than current nuclear power plants. The typical examples of ALWR are System 80+ by ABB-CE, ABWR by General Electric, and EPR by NPI, etc. In Korea, the Korean Next Genera-
0029-5493/00/$ - see front matter © 2000 Elsevier Science S.A. All rights reserved. PII: S 0 0 2 9 - 5 4 9 3 ( 0 0 ) 0 0 2 5 7 - 0
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tion Reactor (KNGR) program, continued since 1992 to prepare for the domestic nuclear power demand expected early in the 21st century, probably after 2005, is on a line extended from the present Korea Standard Nuclear Power Plant (KSNPP) in an overall view. Nonetheless, KNGR aims to have many advanced and different characteristics from its predecessor, which are summarized in Table 1 (Choi et al., 1994a; Cho and Jerng, 1995; Choi and Kim, 1995). As shown in the table, KNGR has many advanced design goals and features. Some important design goals for KNGR are as follows. The design lifetime of KNGR should be 60 years. Core damage frequency (CDF) and containment failure frequency (CFF) of KNGR should be less than 10 − 5 per RY and 10 − 6 per RY, respectively. The thermal margin of KNGR should be greater than 10%. The target of thermal margin is 15%. The economic goal of KNGR is to secure around 20% cost advantage over competing energy sources; for example, coal-fired power generation. Major advanced design features of KNGR are as follows. Double cylindrical containment. Four-train emergency core cooling system that is injected to the reactor vessel directly. Fluidic device in safety injection tank in order to regulate flow rate passively. In-containment refueling water storage tank (IRWST). Main control room designed under the human factors engineering principles and digital I&C technique. The electric power of KNGR is 1350 MW, which is about 30% higher than KSNPP. The design simplification in the emergency core cooling system and auxiliary feedwater system would be an example of the design improvements. KNGR aims at both enhanced safety and economic competence (Choi et al., 1994a). The philosophy of enhancing safety for an advanced reactor is based upon three levels of safety. The first level comes from designing a
system with a high degree of reliability. The sound design with sufficient safety margins and strict quality assurance program in engineering and design will ensure this first level of safety. Despite the assurance offered by careful system design and engineering, it is necessary to anticipate that some incidents and malfunctions will occur during the service life of the system. The second level of safety is to provide means that will forestall or cope with such events. Protection systems are designed to prevent, arrest, or accommodate safely a range of conceivable abnormal situations. The final level of safety is based upon the view that it is prudent to go beyond the levels of safety already described by designing engineered safety features to prevent and mitigate a set of designbasis and severe accidents even though they are considered highly unlikely. The implementation of safety enhancement could be achieved by employing passive safety features in each level of safety described. Passive features eliminate the need for operator involvement at abnormal plant situations including accident conditions by solely relying on natural forces like gravity and natural convection. For example, the core cooling capability in emergency situations could be highly upgraded by automatic delivery of water to cool the core through gravity flow introduction from a tank located above the core, and natural circulation of coolant induced by the temperature difference through heat exchangers that are connected to the steam generator. Passive safety features sometimes reduce plant complexity, but in most cases accompany economic penalty. The dependence of safety enhancement solely on passive features most likely asks for a big economic sacrifice. One must realize that the design objectives include better economics as well as enhanced safety in developing advanced concepts, which means that safety and economics are inter-related rather than independent. Hence, passive features should be adopted taking into account the state of technology and the economics of improvements. As long as the expected safety is within the acceptable level, further enhancement of safety should be based upon cost–benefit studies.
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Table 1 Summary of top-tier requirements for KNGR Item (a) General Reactor type Design life Design Philosophy (b) Safety Safety goal
Thermal margin Operator action time Station blackout coping time LOCA protection (c) Design bases Design criteria Source term Seismic analysis
Requirement
Comparison with current plants
Evolutionary PWR with 4000 MWH 60 years Simplification, design margin, human factor engineering, proven technology
– 40 years –
Core damage frequency, 10−5 per RY Containment failure frequency, 10−6 per RY Radioactive release limit targets: 1 rem/24 h at site boundary; Cs-137 100 TBq at site boundary; probability exceeding above limits shall be less than 10−6 per RY 10–15% \30 min
CDF10−4 per RY CFF10−5 per RY
Coping time, 8 h minimum; installation of AAC
4h
No fuel damage to 15 cm diameter break
–
Licensing design basis (LDB), safety margin basis (SMB) Use the physically based source term Eliminate OBE seismic analysis
LDB only TID-14844 OBE analysis
(d) Se6ere accident mitigation system Combustible gas Hydrogen concentration lower than 10% for 100% clad oxidation control Direct containment Automatically depressurize the core to prevent high pressure core heating control melt; provide the reactor cavity and cooling means Containment Maintain ASME code service level or equivalent criteria for 24 h performance criteria Emergency planning Provide technical basis for reduction of emergency planning boundary (e) System features NSSS
ECCS RHRS Containment I&C and control room design Spent fuel storage (f) Performance Availability goal Load follow capability
Hot leg temperature lower than current design; pressurizer large enough to eliminate PORVs; SG: In-690 and 10% plugging margin for tubes, minimum 30 min dry-out time; depressurization system: safety grade system for depressurization and feed-bleed operation Direct vessel injection; RWST, inside containment; four trains without interconnection Higher ultimate rupture strength than RCS operating pressure; mid-loop operation improvement Cylindrical double concrete Use digital I&C technology and HFE in top-down scheme
8% 10 min
No provision No provision No provision
EP is required
S/G: In-600 and 2–80% plugging
Cold leg injection; RWST outside; two trains – – Analogue technology
10 years+1 core; 20 years using advanced technology
10 years+1 core
Availability, 90%; unplanned trip, 0.8 per RY Daily load follow with frequency control
80–87% in Korea –
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Table 1 (Continued) Item
Requirement
Comparison with current plants
Load rejection capability Refueling cycle Low-level radwaste Occupational radiation exposure
100% load rejection without reactor trip
–
1824 months Solid, 50 m3 per year; liquid, 0.05 Ci per year; gas: 200 Ci per year Accumulated exposure, 100 man-rem per year; individual exposure, 2 rem per year for 5 year average
12–18 months – Accumulated exposure, 200 man-rem per year
(g) Economic Life time cost goal 10-Year cost goal
20% advantage over coal 10% advantage over coal
– –
(11) Site en6elope parameters Seismic design 0.3g Others Seawater temperature, 35.5°C for safety systems, 32.1°C for non-safety systems; ambient air temperature, 33.3°C maximum and −10.6°C minimum Reliability assurance Develop the program to assure the consistency of risk-significant program structures, systems, and components with design bases Configuration Develop computerized database to manage design process management systematically program
In this paper, we will introduce the passive safety features that are incorporated in KNGR at present and the new development of passive safety features for the future application. Through the main body of the paper there is (1) a summary of the Phase I study to select passive features for KNGR, (2) passive safety features for application to KNGR, and (3) future passive design features for the application will be described focusing on the functional requirements, systematic layout, major analyses, and detailed design procedures
2. Passive design features for KNGR As shown in Table 3, the design stages of KNGR are divided into three stages, Phase I, Phase II, and Phase III. In the Phase I development stage, we finished the conceptual design of KNGR. At that time, we set up the basic design requirements of KNGR as shown in Table 1. During the Phase II stage, we developed basic design and detailed design requirements of KNGR. In Phase III, we will proceed with the detailed design and will obtain the design certification to construct KNGR.
0.2 g –
No provision No provision
During the early stage of Phase I development, we considered many passive design features for KNGR such as the following (Choi et al., 1994b). Passive secondary condensing system (PSCS) (Abdollahian et al., 1994). Fuel cells as alternate on-site power supply. Passive external containment cooling system (PCCS) (Kim et al., 1998). Catalytic hydrogen ignitor. Passive containment spray system. Containment overpressure protection system. Passive cooling for specific components. Passive spent fuel pool cooling system. Passive quenching system. Permanent supplemental interfaces. Passive safety injection system. Passive primary residual heat removal system. Passive cavity flooding system (Jae et al., 1993). Emergency boration system. Passive pressurizer spray control system. Ceramic core catcher. Fluidic device in safety injection tank to regulate flow rate (Sugizaki et al., 1992). Flooding exterior of reactor vessel. The above mentioned features have been studied to be adopted in the KNGR design on the
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basis of the screening criteria shown in Table 2. At first, each item went through the technical review in aspects of simplification, integrated plant impact, licensing impact and/or the need of testing. As a result of this technical review, ten items were selected. We then reviewed the safety and economics of each item concerning safety benefits, the overall core melt and/or radiation release frequency not being increased by the passive safety feature. In relation to economics, the passive safety features should not cause a significant increase in overnight plant cost and in overall construction periods. Finally, we performed the detailed cost – benefit analysis to select the final candidate. As a result of the Phase I study, four passive features were retained for application to KNGR Table 2 Screening criteria for passive features Category
Screening criteria
Cost
No significant increase in overnight plant cost No increase in overall construction schedule Essentially no reduction in plant output No fundamental changes to existing licensing principles or criteria No concepts previously rejected by licensing authorities Development/testing required to fit in overall schedule No significant development cost No fundamental R&D required (i.e. proof of principle) No construction of new test facilities required No fundamental methods development No increase in safety risk (core melt or radiation release) No unquantifiable impact on plant design, construction, operation, and maintenance No negative impact on the fundamental reference design configuration No significant decrease in ‘…abilities’a
Licensing impact
Development and testing
Safety benefits Integrated plant impact
Operating and maintenance
No perceived increase in complexity No significant increase in O&M costs a
ity.
Constructability, operability, reliability and maintainabil-
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design to enhance the safety margin (Choi et al., 1994b). Passive secondary condensing system (PSCS). Fluidic device in safety injection tank to regulate flow rate. Passive cavity flooding system. Catalytic hydrogen ignitor. Another passive feature (i.e. PCCS) is under application to KNGR design. Its adoption to KNGR has not yet been determined. Applicability of these passive features on KNGR has been assessed following a three-step approach. 1. Preliminary analyses using computer codes and small-scale experimental facilities. 2. Detailed analyses through large-scale tests, code calculation and uncertainty validation. 3. Quality assurance by design verification and analyses on interface with other systems. The major design activities through the design stages and the major performance analyses for design implementation are shown in Tables 3 and 4, respectively. Details for the passive features and their current design status are presented in the next section.
3. Current design status of the passive design features for KNGR
3.1. Fluidic de6ice The fluidic device is located at the discharge of the safety injection tank (SIT) and controls the injection flow of borated water into the reactor coolant system (RCS) in a passively regulating way, as shown in Fig. 1. It consists of a stand pipe, a control port, and the main body. The borated water in the SIT can be injected into the reactor vessel upper plenum directly following a two-step process. First, the main stream flows through the stand pipe. Once it gets a switch point, the flow through the stand pipe is terminated, and then the injection flow continues through the control port only. The major function of the control port is to generate a vortex flow in the body to make flow resistance, such that the injection flow rate can be passively regulated by
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Table 3 Major design activities through the design stages Passive features
Design stage
Remarks
Phase I (1992–1994)
Phase II (1995–1999)
Phase III (1999–2001)
Passive secondary condensing system
Conceptual design
Detailed design, design certification
Fluidic device in SIT
Conceptual design, preliminary analysis Conceptual design
Preliminary analysis, performance analysis, small-scale test, large-scale test Performance analysis, performance test Performance analysis
Passive cavity flooding system (fusible plug) Catalytic hydrogen ignitor Passive containment condensing system
Conceptual design –
Detailed design, design certification Material selection, detailed design, design certification Performance analysis design implementation, design certification Performance analysis and Performance analysis, test implementation planning
the vortex flow intensity. The system will be capable of reducing the discharge flow to 10% of the maximum flow rate. This system is expected to help extend the allowable start-up time of emergency diesel generators and enhance loss of coolant accident (LOCA) cooling performances (Choi et al., 1994a). After large break loss of coolant accident (LBLOCA), the safety injection system (SIS) provides a big flow rate to the reactor to reflood the core in a short period to meet 10CFR50.46 peak cladding temperature acceptance criteria. SIS in KNGR consists of four high-pressure injection pumps (HPSIPs) and four safety injection tanks. Low-pressure injection pumps, which have much larger capacity compared with HPSIPs, are removed by adopting the direct vessel injection method. At the early stage of LBLOCA, i.e. the blowdown phase, a large flow rate from SIT is required to fill the reactor core lower plenum in order to meet the acceptance criteria of 10CFR50.46. After the downcomer is filled with the injected water, most water injected from the SIT is spilled through the broken pipe into the bottom of the containment system. By adopting the fluidic device in SIT, the injection mode from the SIT can be divided into two steps, i.e. a large amount before the downcomer is reflooded and a relatively small amount after that time.
Beyond design basis events
Design basis events Severe accident
Severe accident Future application
The important factors to design the fluidic device are the change rate of the flow amount and the time to change the flow rate. To decide these factors, sensitivity analyses have been performed for KNGR using CEFLASH-4A and COMPERC-II codes. These are design codes for large break LOCA developed by ABB Combustion Engineering Corporation. CEFLASH-4A code simulates blowdown phenomena of the early stage of LBLOCA, and COMPERC-II code simulates reflooding phenomena of the late stage of the accident. The major design parameters of KNGR are presented in Table 5. As a result of preliminary analyses, the time to change the flow rate is set to 36 s after the initiation of LBLOCA, and the minimum required flow rate by SIT after that time is 12% of maximum flow. Fig. 2 shows the SIT injection flow rate with respect to two change rates, i.e. 20 and 30%. In cases of 20 and 30% flow rates, the injected water is maintained until 198 and 140 s after the initiation of accident, respectively. Without a fluidic device, the SIT would be empty 72 s after the initiation of accident. This means that we can use SIT water for 126 s more than typical SIT in the case of 20% flow rate. Therefore, safety injection pumps are not required until 198 s after the initiation of the accident when the fluidic device, which has a 20% flow rate changing capability,
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and the start of the emergency diesel generator could then be prolonged until that time. The emergency diesel generator starting time, however, should be determined carefully because it is used as the power supply under many other accident conditions. During Phase II, we will complete the experiments to determine the geometric specification and the performance of the fluidic device in order to install it in safety injection tanks. The scales of the
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experiment are set to 1:30 and 1:1 with respect to the flow rate and operating pressure, respectively. The major parameters for the experiments are presented in Table 6, and the schematic diagram of the experimental facility is shown in Fig. 3. As shown in the figure, a compressor will pressurize SIT before the experiment. The fluidic device is located outside SIT so that the height of standpipe can be changed.
Table 4 Major performance analyses for design implementation Passive features
Computer code analysis
Experiment
Passive condensing secondary system
Steady-state natural circulation Flow stability Noncondensable gas effect PSCS valves opening/closing set point System pressure loss Analysis of over-cooling of primary side Determination of condenser capacity Alarm setpoint Water tank capacity evaluation I&C installation location Transient analysis Malfunctioning effect
Pressure loss through the condenser Non-condensable gas effect Transient analysis Heat transfer coefficient CHF initiation verification Thermal stratification in the water tank
Fluidic device in SIT
CFD analysis in fluidic device Safety injection capability Analysis on nitrogen gas entrainment Malfunctioning effect
Effect of the stand pipe elevation Stand pipe installation angle and location Flow resistance coefficient K Probability of nitrogen gas entrainment Flow rate change during service period
Passive cavity flooding system (fusible plug)
Analysis on flooding capability Flow area due to plug melting Malfunctioning effect
Fusible metal melting temperature Metal properties Plug geometry effect
Catalytic hydrogen ignitor
Ignition capability under accident condition Optimization in ignitor number and location Malfunctioning effect
Thermal aging resistance Test under accident conditions Functional test Ignition capability Vibration resistance Impurities effect
Passive containment condensing system
Steady state natural circulation Heat removal capability from containment System pressure loss Determination of condenser capacity Water tank Capacity evaluation I&C installation location Transient analysis
Pressure loss through the system Flow stability Transient analysis Heat transfer coefficient Thermal stratification in the water tank
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Fig. 1. Schematic diagram of the fluidic device in the SIT.
3.2. Passi6e secondary condensing system One of the most effective ways to improve current pressurized water reactor safety is to adopt a passive decay heat removal system for the purpose of long-term core cooling without any external power supply and no operator action. The PSCS, which supplements the auxiliary feedwater system (AFWS), is to secure the heat removal through the steam generators (S/Gs) when AFWS has failed, especially the total loss of feedwater events. The PSCS is installed at the outside of the containment as shown in Fig. 4 (Choi et al., 1994a). The current version of the design requirement for the PSCS requires 8 h of mission time (i.e. it can remove decay heat through PSCS condenser tubes to the secondary water storage tanks without any additional water supply for 8 h). The normal operating condition for the inside of the condenser tubes is saturated
steam for the full operating pressure and temperature of the secondary side The decay heat from the reactor core is transferred to the S/Gs, which transfer heat to the naturally circulating water to make it boil. The Table 5 The major design parameters of KNGR Design parameters
Initial value
RCS Pressure (psia) Thermal power (102% of nominal power) (MW) Peak linear heat generation rate (kw ft−1) Hot leg temperature (F) Cold leg temperature (F) RCS flow rate (lb h−1) Number of safety injection pumps Number of SITs Containment net free volume (ft3) SIT gas/water volume (ft3)
2250 4063 13.94 615 555 166.6×106 4 4 3.3×106 800/1600
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Fig. 2. Required/supplied flow rate during reflooding after LBLOCA. Table 6 The major parameters for the fluidic device experiment Parameters
Maximum
Minimum
Radius of stand pipe (mm) Length of stand pipe (m) Gas pressure in SIT (bar g) Length of SIT (m) Range of flow rate (m3 h−1)
100 8 50 14 20–100
25 4 40 14 10–50
steam discharged from the S/Gs goes into the PSCS condensers, and is condensed on the inside of the tubes due to heat transfer to the water in the secondary water storage tank by the nucleate pool boiling. The condensate is then returned to the S/G via a feedwater line. When the PSCS flow is driven by natural circulation only, its decay heat removal capability is determined mainly by the natural circulation flow rate of the PSCS. The natural circulation flow rate in the PSCS, on the other hand, depends primarily on the buoyancy effects induced by the temperature differences between the cold and the hot regions, and the vertical elevations of the PSCS. Since the vertical elevation is limited by the cost and constructability, the major remaining
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and effective way to increase the heat removal capability of PSCS depends on the heat removal capacity of the condensers. In KNGR, PSCS has two condensers per steam generator. In the Phase I study, the capacity of PSCS was determined to 111 MW per steam generator. That capacity is equivalent to decay heat 5 min after the reactor trip. In Phase II, we modified the capacity of PSCS to 80 MW. This capacity is equivalent to mitigate the most limiting accident, i.e. total loss of feedwater flow accident. We assumed that PSCS is not used for pipe break accidents such as steam line break or feedwater line break. The major performance analyses for design implementation of PSCS are pressure loss, heat transfer coefficient, flow stability, and transient response, as summarized in Table 4. Throughout the Phase II design process, the (1) preliminary system analyses by numerical code, (2) small-scale test (1/480 volume scale, full pressure), (3) detailed system analysis and (4) full-scale, full-pressure test will be executed to ensure its operability and obtain proper design data for KNGR. Currently, RELAP5/Mod3.2 code is used for system performance analysis. The heat transfer coefficient between condenser inside and outside, which is used for code analysis, has been obtained by the results of a single tube test. According to the preliminary probabilistic safety analysis (PSA) on PSCS, it can reduce CDF by about 50%.
3.3. Passi6e ca6ity flooding system KNGR takes the post-flooding process to mitigate the impact of a severe accident on the system and to maintain containment integrity. This flooding process consists of two methods. One is active flooding by motor-operated valves, and the other is passive flooding by fusible metal plugs located inside the pipe between the IRWST and the reactor cavity. During the passive cavity process (i.e. in conditions of loss of active systems), once molten corium falls down through the broken crevice of the reactor vessel to the bottom of the reactor cavity in case of severe accident, the ambient temperature of the cavity increases until the melting point of the metal plugs is reached. If end plug melts, IRWST will supply cooling water to
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the reactor cavity through the pathway by gravity. As shown in Fig. 5, the reactor cavity and IRWST are directly connected to each other. The
end plug sides are normally closed by the metal plugs. The cavity and IRWST can be isolated by valves to maintain the end plugs.
Fig. 3. Schematic diagram of the fluidic device experiment.
Fig. 4. Schematic diagram of the passive secondary condensing system.
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Fig. 5. Schematic diagram of the IRWST and cavity flooding system.
3.4. Passi6e hydrogen ignitor Hydrogen ignitors have been adopted in addition to active hydrogen recombiners to maintain hydrogen concentrations below a detectable limit of 10 volume percent. Hydrogen ignitors consist of active and passive systems. For the active systems, glow plug ignitors that are manually controlled in the main control room will be adopted. For the passive system, catalytic ignitors will be adopted to mitigate severe accident impact on the primary side components. Due to the proven design of the catalytic ignitors, the hydrogen ignition system is expected to minimize the threat of containment failure caused by large hydrogen deflagrations or hydrogen detonations. The catalytic hydrogen ignitor burns hydrogen passively using the noble metals under special conditions. It consists of a catalyst, ignition wires, and a tightly closed metal container as shown in Fig. 6. Once a severe accident occurs, temperature in the containment increases, and then the cover of the hydrogen ignitor opens automatically. After this opening, gases containing hydrogen enter into the metal container. In the container, a catalyst reacts with hydrogen and generates high temperature to activate hydrogen ignition element. Then, hydrogen in the containment is burned by the ignition element.
To adopt this on KNGR, some tests and analyses must be executed through the whole design phases. The major points are functional tests under severe accident conditions and ignition capability. Besides the catalytic hydrogen ignitor, a passive auto-catalytic recombiner (PAR) is con-
Fig. 6. Schematic diagram of the catalytic ignitor.
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Fig. 7. Schematic diagram of the PCCS.
sidered in KNGR because a PAR does not require any operator action in order to actuate it, and the ignitor can impact upon the integrity of containment or equipment near the ignitor.
4. Passive design feature for future application The passive external containment cooling system is the preferred means of decay heat removal following a LOCA and is under development for the future application. The PCCS (shown in Fig. 7), similar to the passive containment cooling system in the simplified boiling water reactor, provides a way of external containment cooling through a natural circulation circuit. It consists of two heat exchangers, relevant lines, and water tanks. Heat transferred from the containment atmosphere to the coolant through the primary heat exchanger tube, which is in the containment, is removed by the condenser tubes to the water tank, which is outside the containment. The condensate then returns to the containment. The PCCS condensers are installed in the external water storage tank at atmospheric pressure. They
cool the steam–air mixture from the containment atmosphere by cold water in the PCCS tank via heat transfer through condensing tubes, and return the condensate into the containment. The major points of tests and analyses will be focused on the heat transfer coefficient of the heat exchangers and flow stability through the natural circulation circuit. The key physical phenomena of the PCCS are as follows (Kim et al., 1998). Condensation in the presence of noncondensable gases. Natural (buoyancy induced) convection and mixing. Stratification in liquid or gas volume. In order to estimate the performance of PCCS, the computer codes should have the capability of simulating the presented phenomena. Therefore, we reviewed some available computer codes, i.e. GOTHIC version 4.1c, GASFLOW version 1.0, etc. Based on the review results, GOTHIC code has been selected to analyze the performance of PCCS in KNGR. We are now performing sensitivity analyses using GOTHIC code for mass and energy release data for KNGR. According to the preliminary analysis results, PCCS has a theoretical possibility of meeting the design criteria. However, to design PCCS as a back-up system for the containment spray system, it is estimated that more than 14 units are required to meet DBA design criteria for the pressure at 24 h after large break LOCA. This would require more than 7000 m2 of evaporator heat exchanger surface in containment, which would introduce serious and perhaps insurmountable equipment arrangement problems. Therefore, more study is required to optimize the size and capacity of the PCCS. Although the study of PCCS design is focused on the case of the design basis accident at present, we will extend it to the case of severe accidents if the results of the present work are positive.
5. Conclusion The approach of developing advanced nuclear systems in Korea is with the continuing goals of enhanced safety and better economics. The safety
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enhancement of the nuclear systems can be achieved by employing passive safety features because of their very high reliability. However, they should be discretely adopted to avoid a big economic sacrifice that might impact on the better economics. Through the inter-related safety and economic study for many passive design features during the Phase I works, four major passive features (i.e. fluidic device, PSCS, passive cavity flooding system, and passive hydrogen ignitor) have been selected for application to KNGR design. Based upon Phase I study results, the following conclusions can be obtained for the current and future design applications. 1. Fluidic device located at the discharge of the SIT injects the borated water into RCS in a passively regulating ways. According to the analysis results, fluidic device will lengthen the water injection by more than 100 s. After demonstration experiments, it will be installed in SIT for KNGR. 2. PSCS using condensers submerged in the PSCS tank cools S/Gs by natural circulation. Until now, we have determined overall configuration and capacity. According to PSA results, the core damage frequency is reduced by about 50% through PSCS. 3. Fusible plugs between reactor cavity and IRWST supply cooling water from IRWST to the reactor cavity by melting and passive hydrogen ignitors using catalytic ignitors. 4. For future safety enhancement, there is one major R&D area of adopting passive safety feature: the PCCS.
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5. R&D for adopting passive safety features is a continuing process for KNGR. It consists of conceptual design study, actual design work, and experimental back-up. References Abdollahian, D., Johnson, C.B., Yedidia, J., 1994. Performance of the SBWR isolation condenser for overpressure protection. Transactions on American Nuclear Society (Trans. ANS), p, 701. Cho, S.J., Jerng, D.W., 1995. Research activities and design requirement for the next generation reactor in Korea. The ZAEA International Workshop on Future LWR, Tokyo, Japan, July 1995. Choi, Y.S., Kim, B.S., 1995. Progress in design, research & development and testing of safety systems for the Korean next generation reactor. IAEA Technical Committee Meeting on the Progress of Water Cooled Reactor Design, Piacenza, Italy, May 1995. Choi, Y.S. et al., 1994a. Research & development on next generation reactor (phase 1) — design concepts of next generation reactor, KRC-92N-J11. Korea Electric Power Corporation, Seoul, Korea, December. Choi, Y.S. et al., 1994b. Research & development on next generation reactor (phase 1) — a feasibility study for incorporating passive safety features into large evolutionary LWR, KRC-92N-J11. Korea Electric Power Corporation, Seoul, Korea, December. Jae, M., Milici, A., Kastenberg, W.E., Apostolakis, G., 1993. Evaluation of accident management strategies for PWRs: cavity flooding. PSA International Topical Meeting, Florida, USA, 26 – 29 January 1993. Kim, Y.H., Todreas, N.E., Driscoll, M.J., 1998. Distributed parameter modeling of the KNGR containment using GOTHIC, MIT-ANP-TR-059. MIT (internal report), February 1998. Sugizaki, T., Matsuoka, T., Shiraishi, T., Nakahara, Y., Kawai, K., Makahara, Y., Okabe, K., 1992. Design study for a passive safeguards system. 5th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-5).