Analysis of the Fukushima Daiichi Nuclear Accident by Severe Accident Analysis Code SAMPSON

Transcription

1 Analysis of the Fukushima Daiichi Nuclear Accident by Severe Accident Analysis Code SAMPSON M. Naitoh, H. Suzuki, H. Mizouchi, K. Hirakawa, A. Takahashi, M. Pellegrini, and H. Okada The Institute of Applied Energy, , Shinbashi SY bldg. 8F, Nishi-Shinbashi, Minato-ku, Tokyo, 15-3, JAPAN, ABSTRACT The severe accident analysis code, SAMPSON, has been applied for analysis of accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Units 1-3. SAMPSON was designed as a large-scale simulation system of inter-connected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to severe accident conditions. The discussions here are focused on the in-vessel phenomena. Since there were some events and phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP, modellings of systems specific to the Fukushima Daiichi NPP were implemented for realistic and detailed analyses: (1) damage model of in-core monitor housings, (2) deterioration model of gaskets of safety relief valves, and (3) model for partial load operation of the reactor core isolation cooling system (RCIC). The models (1) and (2) allowed direct release of steam from the reactor pressure vessel (RPV) to the drywell. The model (3) allowed extraction of steam-water two-phase flow and decrease of water injection flow rate by the RCIC pump. The results with the modified SAMPSON showed that the weight ratios of melts to the total core materials were 67% for Units 1, 38% for Unit 2, and 4% for Unit-3, and that amounts of hydrogen which were generated by metal-water reaction in the RPVs were 572 kg for Unit-1, 93 kg for Unit-2, and 88 kg for Unit-3. 1 INTRODUCTION The Great East Japan Earthquake of magnitude 9. occurred at 2:46 PM on March 11, 211, and was subsequently followed by the huge tsunami. The Fukushima Daiichi Nuclear Power Plant (NPP) Units suffered serious damages from the tsunami, involving core meltdowns in some units, and they released large amounts of fission products to the environment. The station blackout (SBO) occurred after emergency power supply equipment facilities were submerged by the sea water. Passive systems, such as the reactor core isolation cooling system (RCIC) and the high pressure coolant injection system (HPCI) which are driven by steam turbines, and the isolation condensers (ICs), were the only devices for decay heat removal after the reactor scram. Since there was some time delay of alternative water injection into the cores after termination of the passive systems, the core meltdowns occurred, resulting in bottom failure of the reactor pressure vessels (RPVs). It has been determined that molten debris fell into the cavities of the primary containment vessels (PCVs) through the RPV bottoms in Units 1, 2 and 3. Moreover, hydrogen explosions occurred at the operating floors of the reactor buildings (RBs) of Units 1, 3, and 4, resulting in releases of large amounts of fission products to the environment. There was no hydrogen explosion at the RB of Unit 2, because the blowout panel on the RB wall was broken due to the blast from the explosion of Unit-3 and the atmosphere in the operating floor was released to the air. Since Unit 4 was undergoing regular maintenance, it was not operating on March 11 and no core damage occurred to it. The cause of its RB damage was hydrogen inflow from Unit 3 through shared piping of the stand-by gas treatment system. TEPCO, the operator of Fukushima Daiichi NPP Units 1 to 4, has no choice but to carry out their decommissioning. Confirmation of the locations, amounts and characteristics of the debris are required in order to proceed with the decommissioning smoothly. However, this 1/11 pages

2 will take long time, and require remote inspections due to the very high dose rates. Thus, a national project sponsored by the Japanese government has been started in order to grasp the current status inside the reactors and containment vessels through accident progression analyses. In the project, two codes, MAAP and SAMPSON, are being applied for accident progression analyses. This paper describes analysis of the accident progression in the Fukushima Daiichi NPP Units1-3 with the SAMPSON code, especially the discussion is focused on the in-vessel phenomena. 2 SAMPSON CODE AND NEW MODELLING 2.1 Outline of SAMPSON SAMPSON was designed as a large-scale simulation system of inter-connected hierarchical modules covering a wide spectrum of scenarios ranging from normal operation to severe accident conditions [1]. Figure 1 shows the running sequence of SAMPSON modules. The analysis control module (ACM) does not include physical models, but calls and terminates analysis modules as appropriate with respect to time in the event and physical location. The analysis modules for steam explosion (VESUVIUS), hydrogen explosion (DDOC), and 3-D hydrogen distribution in containment (HYNA) are for off-line use, receiving boundary and initial conditions from SAMPSON results. In Vessel Normal Operation Heat up Core Melt Relocation RPV Failure Analysis Control Module (ACM) Ex Vessel Melt Ejection MCCI Pressurization CV Failure Reactor Thermal Hydraulics (THA) Core Heat Up (FRHA) Molten Core Relocation (MCRA) [Steam Explosion (VESUVIUS)] [Hydrogen Explosion (DDOC)] Debris Cooling & Spreading (DCA, DSA) FP Release from Fuel (FPRA) Debris Concrete Interaction (DCRA) FP Behavior in RCS (FPTA) FP Behavior in Containment (FPTA) Containment Thermal Hydraulics (CVPA), [3 D Hydrogen Distribution in Containment (HYNA)] In Vessel Analysis Module Ex Vessel Analysis Module Module for off line use Figure 1 Running sequence of SAMPSON modules The major features of SAMPSON are as follows. (1) Minimum use of empirical correlations to eliminate tuning parameters as much as possible. (2) Maximum use of mechanistic models and theoretical-base equations. (3) Composition of 11 analysis modules and 3 other modules (DDOC, VESUVIUS, and HYNA) for off-line use. (4) Validation by a wide range of analyses for separate effect tests and integral tests, mainly through participation in OECD/NEA projects [2, 3]. Distinctive models incorporated in SAMPSON and their validations are described in references [2-8]. Typical mechanistic models are included in the molten core relocation analysis (MCRA) module as follows. Multi-component model 2/11 pages

3 - 9 liquid components (particles are treated as liquid components): Water; liquid fuel; liquid steel; liquid Zircaloy; liquid control material; fuel particles; steel particles; Zircaloy particles; and control material particles. - 6 gas components: Steam; oxygen; nitrogen; hydrogen; argon; and helium. Since argon was sometimes used as a thermal insulator in tests and helium was sometimes used in flow behavior tests as a substitute for hydrogen, these gases have been applied for validation test analyses. - 4 solid components: Rods; steel; Zircaloy; and crust. Mass conservation equations for 9 liquid components and 6 gas components. Energy conservation equations for 9 liquid components and for a mixture of 6 gas components. Momentum conservation equations for 2 groups of liquid components (e.g. water and others) and for a mixture of 6 gas components. Interaction between three phases described by an interfacial area considering phase change: - For example, melting and freezing of all solids. The MCRA module with mechanistic models enables independent calculations of the intact state, continuous molten phase, particle phase, and continuous crust phase as core behavior. The major models incorporated in MCRA and their validations are described in reference [5]. The Phebus-FPT1 test analysis results are shown in Figure 2 as an example of SAMPSON validation [3]. SAMPSON could well reproduce behaviors of molten core relocation and fission product release (iodine in the example of Fig. 2(b)) from fuel. Elevation (m) Analysis Test Melt down Release rate Mass of fuel (kg/m) Time ( 1 3 s) (a) Relocation of fuel after finishing the test (b) Iodine release from fuel Figure 2 OECD/NEA ISP-46: Phebus-FPT1 test analysis results Test Analysis Applications of SAMPSON to PWR and BWR severe accident analyses are described in references [9, 1]. 2.2 Modelling of systems specific to Fukushima Daiichi NPP In order to analyze the accident progression realistically and in detail, modellings are newly required, since there were some events and phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP. The typical models to be added to the original SAMPSON are the following. (1) A damage model for the source range monitors (SRMs) and the intermediate range monitors (IRMs), which are types of in-core monitors; their damage resulted in direct release of fluid from the RPV to the primary containment vessel (PCV). (2) A model for leakage from gaskets of SRV piping; the deterioration of the gaskets resulted in direct release of steam from the RPV to the PCV. 3/11 pages

4 (3) A partial load operation model for the RCIC; the deterioration of RCIC turbine performance resulted in the change of extracted and exhaust fluid mass flow rate and in the decrease of water injection flow rate into the core. (4) A partial condensation model for the suppression pool. (5) A model for leakage from the top flange of the PCV; the leakage resulted in release of fluid from the PCV to the reactor building. The above items (1), (2), and (3) directly relate to the in-vessel phenomena, and the others to the ex-vessel phenomena. Here, the discussion is focused on the in-vessel phenomena Damage of SRM/IRM housing There are several kinds of in-core monitors such as SRMs, IRMs, traversing in-core probes (TIPs), and local power range monitors (LPRMs). These housings penetrate the RPV bottom. The pressure boundary of the SRMs and IRMs is inside the core, since their bottom ends are open to the PCV drywell, while the pressure boundaries of the others are outside the RPV, as shown in Figure 3. Under the normal operating condition, however the SRM/IRM housings always have the pressure difference of about 7 MPa between outside and inside of the housing, they cannot be broken. Under the severe accident conditions, they might buckle when their temperatures rise up even before their melting point (about 1,7 K). If the buckling would actually occur, it results in direct release of steam into the drywell. Figure 3 Possible leak location of RPV fluid into the PCV The buckling condition is given by simple Von Mises relation as follows [11]..27, (1) where P cr : critical pressure to buckling E: modulus of elasticity t: thickness r: internal radius 4/11 pages

5 Since the modulus of elasticity of steel (E) is a 12 function of temperature [12], the critical buckling pressure (P cr ) was expressed by a 1 function of temperature as shown in Figure 4. 8 The Figure shows that the buckling occurs when the housing temperature reaches 1,3 6 K, which is lower than the melting point of the housing (about 1,7 K), even under the RPV 4 rated pressure of about 7 MPa. 2 1,3 K It was now supposed that a crack was 7 MPa generated at the location of the buckling 5 1, 1,5 2, occurrence. The unknown is the crack size Temperature (K) when the buckling occurs. Here, the Figure 4 Temperature dependency of Pcr following conditions were assumed. Width of the crack:.5 mm. Length of the crack: circumferentially 6. When the SRM/IRM housings reached the melting temperature, the leak flow area from the melt was supposed to be equal to the cross sectional area of the housing tubes. And a mass flow rate from the leak location was calculated based on the choked flow condition Deterioration of SRV gaskets The other possibility of leakage was thought to be from the gaskets at the flanges of SRV piping. As shown in Figure 3, the SRV line was connected to the main steam line and the other end was opened in the suppression pool. The SRV itself was installed in the drywell. The design maximum temperature of the SRV gaskets was 723 K. Since high temperature steam would flow into the SRV piping during the SRV opening under the severe accident conditions, the sealing function of the gaskets should be deteriorated when the temperature exceeded 723 K, resulting in steam release from the RPV to the PCV drywell. The unknown is the leak area of the deteriorated gasket. Here, the following conditions were assumed. Width of the leak portion of the gasket:.2 mm. Length of the leak portion of the gasket: circumferentially 6. The above SRM/IRM crack size and the leak area of the SRV gaskets were just the assumptions without any evidence. Therefore, a parametric survey by changing the crack size and the leak area are required Partial load operation of RCIC For Unit-2,the RCIC was the only decay heat removal system available during the period after the reactor scram and until the alternative water injection by the fire pump. The RCIC had a function to inject fresh water into the core, and it was originally designed to automatically start on receiving a signal of low water level in the core region and to automatically stop on receiving a signal of high water level (so called L-8 level). The RCIC was initially activated by manual operation just after the reactor scram at Unit-2, but it automatically stopped with the L-8 signal. Then this on-off operation was repeated until the SBO. The Unit-2 DC power supply system was also submerged by the sea water and its function was lost at the time of SBO occurrence. Since the RCIC had been working at the time of SBO occurrence, the RCIC valve might have been stuck open as it was at the total loss of power, and it continued to work for about 66 hours (until 9:, March 14) even after the total loss of power, without receiving the L-8 signal to stop it. When the RCIC continued to work without L-8 signal, the water level in the RPV must continue to rise to the steam extraction line. Thus, steam to drive the RCIC turbine must include some water. Since the deterioration of the RCIC performance under such Pcr (MPa) 5/11 pages

6 two-phase flow condition was unknown, it was assumed by giving the deterioration parameter,, as follows..3, (2) where : ratio of the RCIC performance to the rated condition It was supposed that the RCIC had worked under the rated condition until the SBO occurrence; =1.. 3 ANALYSIS CONDITIONS The core was divided into 8 rings and 13 axial nodes with r-z 2-D coordinates; 4 rings for fuel region and 4 rings for control rod/bypass region, as shown in Figure 5. The exact dimensions of Units 1-3 geometries [12] were reflected in the input data. The decay heat was provided by TEPCO [13]. The boundary conditions such as working period of the ICs, the RCIC, the HPCI, and the SRVs were decided based on open information which TEPCO released [13]. The melting of the constituent materials were assessed by their own melting points in the analysis. Since the eutectic reaction was also considered, some materials were considered to melt at lower temperature than the melting point of each material. 1,5 K: Eutectic temperature of B 4 C+Steel. 2,473 K: Eutectic temperature of UO 2 +Zr. Figure 5 Node division The fuel cladding burst was evaluated by stress. The break of the RPV bottom was evaluated by either its melting point or creep rupture criterion by the Larson-Miller parameter. 4 ACCIDENT ANALYSIS OF UNITS 1-3 Ring 4 (24*) Ring 6 (28*) Ring 2 (19*) Ring 1 (76*) Control Rod/Bypass Region Ring 7 (116*) Ring 3 (96*) Ring 5 (112*) Fuel Region Ring 8 (26*) 13 Nodes Axially *: Number of fuel bundles or control rod blades included in each ring The accident progressions were analyzed by the modified SAMPSON code, SAMPSON-B1.3, which included above mentioned modellings. The results described in the following sections were focused on the in-vessel phenomena. Therefore, for the calculations of hydrogen mass, only metal-water reactions in the RPVs were considered. To evaluate total mass of hydrogen, molten core-concrete reaction in the cavity of the drywell shall be considered in future. 4.1 Unit-1 Figure 6 shows the calculated RPV pressure transient with the measured data. Before the SBO occurrence, the RPV pressure was recorded on a chart, and after the SBO, the measurements were manually and intermittently made by reactor personnel using devices powered by portable DC batteries. The ICs, which were the only devices for decay heat removal, had worked until the SBO. After the SBO, the SRVs had repeated opening and closing to release excess steam generated by decay heat. The RPV depressurization must occur sometime between the 2 measurements, 7 MPa at 2:7 on March 11 and.9 MPa at 2:45 on March 12. The analysis result was summarized as follows. 6/11 pages

7 RPV pressure (MPa) 6th European Review meeting on Severe Accident Research (ERMSAR-213) RPV pressure (MPa) : IC working period (A) and (B) (A) (A) (A) SBO occurrence Time after scram (min) Close-up March 11 16: 18: 2: 22: Leakage from SRV gasket at18:52 Leakage from SRM at19:1 Initiation of fuel cladding burst at 21:9 24: RPV bottom damage at 21:16 The dose rate in the RB increased to 288 msv/h at 21: Time after scram (h) Figure 6 Pressure transient at Unit-1 6 March 12 2: Fresh water injection at 5:46 on 3/12 : SAMPSON : Measurement The calculation showed that the steam leakage from the SRV gaskets started at 18:52. There were 4 SRMs and 8 IRMs dispersively installed in Unit 1. The first SRM damage occurred at 19:1. At this timing, depressurization was very slow because of balance of the leakage flow rate through the SRV gasket and the SRM, and the steam generation rate by decay heat. Then a depressurization became faster with increase of a number of damaged SRM/IRM. The RPV pressure and the drywell pressure became almost equivalent at about 7.5 hours after scram (22:2 on March 11). During this depressurization process, fuel cladding burst, fuel melt due to generation of eutectic compounds, and finally creep rupture at the RPV bottom sequentially occurred. The calculated pressure transient could generally reproduce the measured data. Major events calculated by SAMPSON-B1.3 are tabulated in Table 1. The calculated time of initiation of fuel cladding burst was 21:9 on March 11 (6.37 hours after the scram). The fission products were released into the drywell by direct steam leakage through the SRM/IRM and the SRV gaskets. This caused dose rate increase in the RM at 21:51. The core melt initiation was 6.38 hours after the scram, which was due to Table 1 Major events at Unit-1 Event Time after scram Collapsed water level to TAF Leakage from SRV gasket Collapsed water level to BAF Leakage from SRM housing Fuel cladding burst Initiation of fuel melt due to generation of eutectic compound Failure of RPV bottom 2h 46m 4h 5m 4h 18m 4h 23m 6h 22m 6h 23m 6h 29m generation of eutectic compound. And finally the creep rupture at the RPV bottom occurred at 6.48 hours after the scram (21:16 on March 11), resulting in falling of debris into the cavity of the drywell. The alternative water injection to the reactor core by the fire pump started at 5:46 on March 12, which was after the RPV bottom break. 7/11 pages

8 RPV pressure (MPa) th European Review meeting on Severe Accident Research (ERMSAR-213) Table 2 shows calculated debris composition just before the start of alternative water injection. Almost 67% of the core materials had melted down. It was considered that the rest remained in the RPV. 22% of the total melt materials remained in the liquid phase and 78% became particulates. The calculated mass of generated hydrogen due to metal-water reaction in the RPV was 572 kg. Other calculations with a parameter of the leak area from the SRM/IRM and from the SRV gaskets were performed: 5% smaller leak areas than the ones discussed in the sections and (Case A) and 5% larger leak areas (Case B). In the case A calculation, the RPV pressure was kept high even at 2:45 on March 12, at which the measured pressure was.9 MPa. The case B result showed rapid pressure decrease earlier than 2:7, at which the measured pressure was 7 MPa. Thus, it was concluded that the other calculations were not reproducing the accident behavior of Unit Unit-2 Figure 7 shows the calculated RPV pressure transient with the measured data. The calculated pressure transient showed good agreement with the measured data by considering the partial load operation of the RCIC. The temporal pressure increase at about 13.5 hours after scram (4:2 on March 12) was due to change of water source from the condensate storage tank (CST) to the suppression pool in which the water temperature was much higher than the one in the CST. The pressure increase at 66 hours after scram (9: on March 14) was due to RCIC trip. And after that, the SRVs had worked for a while. At 18:2 on March 14, the SRV was opened manually, resulting in rapid depressurization. The water source was changed from the condensate storage tank to the suppression pool. SAMPSON Measured RCIC operation under two-phase flow condition (Partial load operation) RCIC operation under rated condition Table 2 Debris composition at Unit-1 Component Weight (kg) Liquid Fuel Zr Control rod Steel Particle Fuel Zr Control rod Steel Total 18,858 3,83 4,72 1,956 66,983 32,832 19, ,24 85,841 SRV manual open Time after scram (h) Figure 7 Pressure transient at Unit-2 Major events calculated by SAMPSON-B1.3 are tabulated in Table 3. The calculated time of initiation of fuel cladding burst was 19:23 on March 14 (76.6 hours after the scram), which was soon after the RPV depressurization (18:2 on March 14). And the release of the fission products started due to the cladding burst. Even after the alternative water injection by the fire pump at 77 hours after Table 3 Major events at Unit-2 Event Time after scram Collapsed water level to TAF Collapsed water level to BAF Fuel cladding burst Initiation of fuel melt due to generation of eutectic compound Failure of core support plate Failure of RPV bottom 74h 48m 75h 3m 76h 36m 77h m 78h 3m 78h 42m 8/11 pages

9 8 6th European Review meeting on Severe Accident Research (ERMSAR-213) scram (19:54 on March 14), the RPV bottom broke due to the creep rupture at 78.7 hours after scram (2:29 on March 14). Table 4 shows calculated debris composition. Almost 4% of the core materials had melted down and fell down into the cavity of the drywell. 2% of the total melt materials remained in the liquid phase and 8% became particulates. The calculated mass of generated hydrogen due to metal-water reaction in the RPV was 93 kg. Table 4 Debris composition at Unit-2 Component Weight (kg) Liquid Fuel Zr Control rod Steel Particle Fuel Zr Control rod Steel Total 15,276 5,65 6, ,372 65,389 32,639 25,94 1,222 5,588 77, Unit-3 Figure 8 shows the calculated RPV pressure transient with the measured data. The calculated pressure transient showed good agreement with the measured data. Since DC batteries were available at Unit-3, the RCIC and HPCI were manually operated by controlling their valves. The pressure decreased to about 1 MPa after the HPCI operation because it had enough capacity to remove decay heat. The pressure of 1 MPa was almost the lower limit at which the HPCI could work. At 9:8 on March 13, the SRV was opened manually, resulting in rapid depressurization. 7 6 RCIC Operation HPCI Operation SRV manual open RPV preeure (MPa) SAMPSON Measured Time after scram (h) Figure 8 Pressure transient at Unit-3 Major events calculated by SAMPSON-B1.3 are tabulated in Table 5. The calculated time of initiation of fuel cladding burst was 1:4 on March 13 (43.28 hours after the scram), which was about 1 hour after the RPV depressurization by manual open of the SRV (9:8 on March 13). And the release of the fission products started due to the cladding burst. At 11:37 on March 13 Table 5 Major events at Unit-3 Event Time after scram Collapsed water level to TAF Collapsed water level to BAF Fuel cladding burst Failure of RPV bottom 35h 7m 35h 29m 43h 17m 44h 5m (44.83 hours after scram), the RPV bottom broke due to melt. The alternative water injection by the fire pump started at 9:25 on March 13 (42.63 hours after scram), but it was not effective to remove heat from debris in the lower plenum since the injection flow rate into the core was very small. 9/11 pages

10 The calculated debris composition was similar to the one at Unit-2. Almost 33% of the core materials had melted down and fell down into the cavity of the drywell. 3% of the total melt materials remained in the liquid phase and 7% became particulates. The calculated mass of generated hydrogen due to metal-water reaction in the RPV was 88 kg, which was smaller than the one at Unit-2, because of shorter duration of high temperature of the fuel rods. 4.4 Ex-vessel analysis The modeling for ex-vessel analysis as described in the section 2.2 (4) and (5), is now in progress. The total accident progression analyses of Fukushima Daiichi NPP Units 1-3 including in- and ex-vessel phenomena, and also source term evaluations are the future issues. 5 CONCLUSIONS The accident progressions at Fukushima Daiichi NPP Units 1-3 were analyzed with SAMPSON-B1.3. Since there were some events and phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP, modellings of systems specific to the Fukushima Daiichi NPP were implemented for realistic and detailed analyses. The major models which were newly added to the original SAMPSON were as follows. (1) The buckling model of the SRMs and the IRMs. (2) The model for leakage from gaskets of SRV piping. (3) The partial load operation model for the RCIC. The results are summarized as follows. (1) The weight ratios of melts to the total core materials were 67% for Unit 1, 38% for Unit 2, and 4% for Unit-3. (2) The molten debris fell into the cavities of the PCVs through the RPV bottoms in Units 1, 2, and 3, because of delay of the alternative water injection by the fire pumps. (3) The calculated masses of generated hydrogen due to metal-water reaction in the RPVs were 572 kg for Unit 1, 93 kg for Unit 2, and 88 kg for Unit 3. The total accident progression analyses of Fukushima Daiichi NPP Units 1-3 including in- and ex-vessel phenomena, and also source term evaluations are the future issues. ACKNOWLEDGEMENT This work has been sponsored by the Japanese Ministry of Economy, Trade and Industry. REFERENCES [1] H. Ujita, et al., Development of severe accident analysis code SAMPSON in IMPACT project, J. Nucl. Sci. Technol., 36, 11, 1999, 176. [2] T. Ikeda, et al., Analysis of International Standard Problem No. 45, QUENCH6 Test at FZK by detailed severe accidents analysis code, IMPACT/ SAMPSON, J. Nucl. Sci. Technol., 4, 4, 23, 246. [3] T. Ikeda, et al., Analysis of core degradation and fission products release in Phebus FPT1 Test at IRSN by detailed severe accidents analysis code, IMPACT/SAMPSON, J. Nucl. Sci. Technol., 4, 8, 23, 591. [4] H. Ujita, et al., Development of debris coolability analysis module in severe accident 1/11 pages

11 analysis code SAMPSON for IMPACT project, J. Nucl. Sci. Technol., 36, 1, 1999, 94. [5] N. Satoh, et al., Development of molten core relocation analysis module MCRA in the severe accident analysis code SAMPSON, J. Nucl. Sci. Technol., 37, 3, 2, 225. [6] H. Ujita and M. Hidaka, Model verification of the debris coolability analysis module in the severe accident analysis code SAMPSON, J. Nucl. Sci. Technol., 38, 4, 21, 229. [7] M. Hidaka and H. Ujita, Verification for flow analysis capability in the model of three-dimensional natural convection with simultaneous spreading, melting and solidification for the debris coolability analysis module in the severe accident analysis code SAMPSON (I), J. Nucl. Sci. Technol., 38, 9, 21, 745. [8] M. Hidaka, N. Sato, and H. Ujita, Verification for flow analysis capability in the model of three-dimensional natural convection with simultaneous spreading, melting and solidification for the debris coolability analysis module in the severe accident analysis code SAMPSON (II), J. Nucl. Sci. Technol., 39, 5, 22, 52. [9] H. Ujita, et al., PWR and BWR plant analyses by severe accident analysis code SAMPSON for IMPACT project, Proc. GENES4/ANP23, Sep , 23, Kyoto, Japan, Paper 174. [1] M. Naitoh, S. Hosoda, and C. M. Allison, Assessment of water injection as severe accident management using SAMPSON code, Proc. ICONE-13, May 16-2, 25, Beijing, China, ICONE [11] Open book obtained from the following url, chapter 2, p [12] Open information obtained from the web site; [13] Open information obtained from the web site; 11/11 pages