Simulations of Core Damage Progression for TMI-2 Severe Accident Using CINEMA Computer Code

. As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. Te CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fssion product analysis in the vessel and the containment, respectively. Te severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. Tis analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. Te CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. Te CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. Te CINEMA results and measured total hydrogen production are very similar, which were approximately 465kg and 460kg, respectively.


Introduction
As an integrated severe accident computer code development in Korea, CINEMA (Code for INtegrated severe accidEnt Management Analysis) has been developing for a severe accident sequence analysis from an initiation event to a containment failure [1].Te objective of this code development is to have a Korean severe accident sequence analysis code by using the existing domestic DBA (design basis analysis) code system for the severe accident sequence analysis.Figure 1 shows the code structure of the CINEMA.Te CINEMA computer code is composed of CSPACE [2], SACAP (Severe Accident Containment Analysis Package) [3], and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident) [4], which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fssion product analysis, respectively.
Te CSPACE is divided into the SPACE (Safety and Performance Analysis CodE for nuclear power plants) and the COMPASS (COre Meltdown Progression Accident Simulation Software) models, which is to calculate an overall RCS thermal hydraulic response during core melt progression in SPACE computer code [5][6][7] and an invessel melt progression to reactor vessel failure in COMPASS computer code [8,9].
Many computer codes have been developed for the safety analysis including the severe accident, such as MELCOR [10,11], MAAP5 [12] in USA, and ASTEC [13] in France for a severe accident sequence analysis.Many accident analysis have been performed to analyze accident progression and general reactor safety problems [14][15][16][17][18] and to solve the specifc thermal hydraulic issues on the safety analysis including the severe accident, such as natural circulation, external reactor vessel cooling, reactor vessel cooling, and so on [19][20][21][22].Tese results may be used to solve the nuclear license issues for a large power reactor and small reactors.
As a part of a validation of the CINEMA computer code, the TMI-2 (Tree Mile Island unit 2) core melt progression has been analyzed.At frst, the steady state simulation was performed to validate the CINEMA input model for TMI-2 by a comparison of the CINEMA results with a TMI-2 operating condition.After this, the transient of the TMI-2 severe accident has been analyzed to validate CSPACE modes of the CINEMA computer code on the core melt progression in-vessel, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core.Te CINEMA results on main parameters, such as a pressurizer pressure and an integrated hydrogen generation mass have been compared with the TMI-2 severe accident data.In addition, the mass fow rate through the pressurizer PORV (Pilot Operated Relief Valve), which is an RCS coolant loss location, collapsed water levels in the pressurizer and the core, and the fuel rod temperature and mass in the TMI-2 severe accident have been analyzed using the CINEMA computer code.

Detailed Description of TMI-2 Severe Accident
A severe accident including in-vessel melt progression with corium relocation to the lower plenum of the reactor vessel occurred in the TMI-2 PWR (pressurized water reactor) on March 28, 1979 [23,24].Many analyses on the TMI-2 severe accident were performed using many computer codes, such as MELCOR and SCDAP/RELAP5 [25,26].Te initiating event for TMI-2 severe accident was a TLOFW (total loss of feed water) with a SBLOCA (small break loss of coolant accident) that resulted in a signifcant cladding oxidation, a partial melting of the core material, a signifcant release of fssion products from the fuel, and melted core material relocation to the lower plenum.Te TMI-2 severe accident was terminated by a coolant injection into the in-vessel using the HPI (high pressure injection) water of the ECCS (emergency core cooling system).Te following four main phases occurred during the TMI-2 severe accident.
(i) Phase 1. Te initiating event of the total loss of feed water with the SBLOCA through the pressurizer PORV from 0 s to 4,440 s.Tis phase includes an operation and stop of the HPI, an operation of the makeup pump, and shutdown of the RCP.In this phase, boiling and two phase fow occurred in the core region.(ii) Phase 2. Heat up and melting of the fuel rod by core uncover 4,440 s to 10,440 s.All RCP shutdown and core uncover occurred, which resulted in the fuel cladding oxidation and fuel melting, and relocation of the melted core material to the lower part of the core.(iii) Phase 3. Core refood by an operation of the RCP and HPI from 10,440 s to 13,440 s.Tis phase includes the initial core quenching by a restart of the RCP from 10,440 s to 11,580 s.Te melted core material was quenched by the HPI water on the operation of feed and bleed for the RCS.In spite of this, the melted core material was relocated to the lower head in the reactor vessel.Science and Technology of Nuclear Installations (iv) Phase 4. Core relocation to the lower plenum of the reactor vessel and cooling in-vessel from 13,440 s to 18,000 s.Tis phase represents that the central region of the partially melted fuel was not coolable by HPI water because the corium pool with crust formation was generated in the middle of the core region.Between 13,440 s and 13,560 s, the crust in the corium pool was failed, which resulted in corium relocation to the lower plenum of the reactor vessel.However, the relocated molten material was quenched by coolant in the reactor vessel at this phase.
Table 1 shows detailed main events in the TMI-2 severe accident.Tis accident was initiated by the TLOFW which led to a pressure increase in the secondary and primary systems followed by turbine and main feedwater pump trip at 0 s (second).Pressure increase in the primary system caused the opening of the pressurizer PORV by a high pressure opening set point of 15.5 MPa.Te reactor was tripped by a high pressurizer pressure signal.Te pressurizer PORV was not closed when the pressurizer pressure was reached 15.2 MPa, which was the closed set point.Unknown to the operators, the pressurizer PORV failed to close and started the SBLOCA.Te break location was the top of the pressurizer.Te 1 (of 3) makeup pump 1B was operated at 41 s.Te HPI (high pressure injection) was operated by high pressurizer pressure signal at 122 s.However, this pump was stopped by the operator because of a misreading of the high level of the pressurizer at 278 s.Te auxiliary feedwater was startup at 480 s.Boiling occurred in the core at 552 s.Te Bloop and A-loop RCP (reactor coolant pump) s were stopped at 4,440 s and 6,000 s, respectively.Te core was uncovered at 6,184 s, which resulted in an increase of the fuel cladding temperature.Te fuel cladding oxidation begun at 7,442 s.Te fuel cladding was failed by overstrain at 7,719 s.Te operator closed the block valve in PORV pipe line at 8,340 s, which meant the end of the SBLOCA.However, the fuel was melted at 9,014 s.One B-loop RCP was operated to supply the coolant into the core at 10,440 s.Tis pump was stopped at 11,580 s.Te feed and bleed operation of the primary system was started to cooldown the core at 12,000 s.However, the melted core material relocated to the lower head at 13,440 s because of the molten pool formation with crust formation in the core.Te general emergency declared at 18,000 s.
Figure 2 shows the end state of TMI-2 severe accident [26].Approximately, 62 tons of the all core material was melted in the core and 19 tons of the corium of the melted core material was relocated to the lower head in the reactor vessel during the TMI-2 severe accident.However, the reactor vessel failure did not occur due to the operation of the HPI.

CINEMA Input Model and Steady State Results
Te thermal power of the TMI-2 PWR was 2,772 MW t .Te TMI-2 core had 177 fuel assemblies, which had approximately 93 tons of fuel and 23 tons of the fuel cladding.Te RCS of the TMI-2 included the reactor pressure vessel, two hot legs, the pressurizer including the surge line, two onethrough steam generators, four RCPs, and four cold legs.Te high pressure of the RCS in the TMI-2 was controlled by the pressurizer PORV. Figure 3 shows an input CSPACE model of the CINEMA computer code for the TMI-2.All RCS including the reactor vessel, two hot legs, the pressurizer, two steam generators, four RCPs and clod legs, pressurizer PORV, and HPI are modeled as shown in Figure 3. Tree radial channels of central, middle, and outer for 177 fuel assemblies in the TMI-2, and 5 axial nodes of 1 to 5 are used in CSPACE core input model.Fuel and control rods for  A steady state simulation for approximately 2,000 s was performed to verify the input model of CINEMA computer code for TMI-2 severe accident.Table 2 shows a comparison of the CINEMA simulation results with the TMI-2 operating condition [13].Te steady state results of the CINEMA computer code for a selected set of parameters were in very good agreement with the TMI-2 operating conditions.Te steady state results obtained from the CINEMA simulation were used as initial conditions for the TMI-2 transient calculation.

. CINEMA Results and Discussion
Figure 4 shows the CINEMA results with the TMI-2 data on the pressurizer pressure.Te total loss of feed water caused a dryout of the secondary side of the steam generators.Tis loss of heat sink capability of the RCS resulted in the coolant to expand and initially increased to/of the pressurizer pressure to opening setting point of the pressurizer PORV, which was 15.7 MPa.After the pressurizer PORV opened, the PORV failed to close as the pressurizer pressure decreased which resulted in RCS coolant loss through the PORV pipeline and initiated the SBLOCA.Te HPI was operated by the SI (safety injection) signal.However, this actuation was stopped by operators who thought that the pressurizer water level indicated a nearly full of the RCS, while coolant continued to be lost through PORV pipeline.Te pressurizer pressure remained at the saturation pressure of approximately 7 MPa for the primary system, after an initial decrease of the RCS pressure.After the fuel cladding temperature is higher than the saturation temperature, the pressurizer pressure increases by coolant boiling in the core.Increase of heat transfer from the fuel cladding to the coolant by decay and oxidation heats resulted in a coolant boiling and the pressurizer pressure to increase rapidly.Te CINEMA results on the pressurizer pressure are very similar to the TMI-2 data in general.Moreover, the CINEMA results after molten pool formation and quenching at 10,440 s is not similar to the TMI-2 data 4 Science and Technology of Nuclear Installations because of complex quenching phenomena in the core by restart of the RCP and HPI operation.
Figure 5 shows the CINEMA results on mass fow rate through the pressurizer PORV in the TMI-2 severe accident.Te opening of the pressurizer PORV at 3 s resulted in the mass fow rate through the PORV pipe line.In addition, the PORV failed to close led to loss of coolant through the PORV at approximately 10-30 kg/s of the coolant mass fow rate.After operator acted the close of the PORV line block valve at 8,340 s, the coolant mass fow rate reached to 0 kg/s.After the start of a feed and bleed operation for the primary system at 12,000 s, the coolant mass fow rate showed periodically peaks by the opening and closing of the PORV pipeline.
Figure 6 shows the CSPACE results of the CINEMA computer code on the collapsed water level for the pressurizer in the TMI-2 severe accident.When the pressurizer PORV opened, the collapsed water level increased to the top of the pressurizer because the coolant in the RCS moved to the pressurizer through the surge line.Te collapsed water level decreased because of coolant loss through the PORV pipeline until 8,340 s.Te pressurizer water drained to the RCS after the PORV block valve was closed, which resulted in the rapid decrease of the water level.After the start of HPI actuation by the feed and bleed operation of the primary system at 12,000 s, the water level increased.Figure 7 shows the CSPACE results of the CINEMA computer code on the collapsed water level for the core in the TMI-2 severe accident.After no PORV valve closing at 13 s, the water level decreased.Te water level increased for short time after shutdown of Aloop RCP at 6,000 s.After restart, one B-loop RCP at 10,440 s and HPI operation at 12,000 s, the water level increased.

Science and Technology of Nuclear Installations
Figure 8 shows the CINEMA results on fuel cladding surface temperature for the central channel.Position of nodes 1, 2, 3, 4, and 5 are 0.36 m, 1.09 m, 1.82 m, 2.55 m, and 3.28 m from the bottom of the fuel rod, respectively.Te fuel cladding surface temperature is a little higher than the coolant temperature.Fuel cladding surface temperature at Node 5 rises when a core uncover occurs at the top of the core.Te fuel cladding surface temperature rapidly increases by no quench of the core, which results in the damage of the fuel cladding and hydrogen production by the fuel cladding oxidation.When the fuel cladding temperature reaches approximately 1,000 K, fuel cladding oxidation begins.Following this time, the fuel cladding temperature rises abruptly due to the heat generation from oxidation.Te fuel cladding surface temperature rapidly increases by no quench of the core, which results in the damage of the fuel cladding.Te fuel cladding is quenched rapidly by water injection into the core at 10,440 s, which was occurred by RCP operation.Because of water in the bottom of the core as shown in Figure 7, the fuel cladding surface temperature at Node 1 did not increase.Figure 9 shows CINEMA results on the fuel cladding mass for the central channel in the core.Fuel cladding mass of only one fuel rod is in this Figure .Te fuel cladding surface temperature rapidly increases by oxidation heat when the temperature reaches at approximately 1,500 K. Top of the fuel cladding (Node 4, 5) is melted and relocated to the middle part of the core (Node 2, 3).Fuel cladding surface temperature and relocation afect the hydrogen generation from the oxidation.Te lowest value of the fuel cladding temperature after approximately 9,000 s means that all cladding is melted and relocated to the lower part.
Figures 10 and 11 show the CINEMA results on fuel of UO 2 temperature and mass for the central channel, respectively.Positions of nodes 1, 2, 3, 4, and 5 are same for fuel cladding cases.Te fuel temperature rapidly increases by no quench of the core, which results in the damage of the fuel.Te fuel is quenched rapidly by water injection into the core at 10,440 s, which was occurred by RCP operation.Fuel mass of only one fuel rod is in this fgure.Te fuel surface temperature rapidly increases by cladding oxidation heat.Top of the fuel (Nodes 4 and 5) is melted and relocated to the middle part of the core (Nodes 2 and 3).Zero value of the fuel mass after approximately 9,500 s means that in Node 5 all fuel is melted and relocated to the middle part.Te upper part fuel rod of nodes 4 and 5 melts and relocates to the core middle part of the nodes 2 and 3, which is very similar to the melt pool formation in the real TMI-2 data, as shown in Figure 2.
Figure 12 shows a comparison of CINEMA results with TMI-2 data on the generated hydrogen mass.Te calculated hydrogen generation mass is in general agreement with the hydrogen production estimated from the date base of the TMI-2 severe accident.Te estimated hydrogen generation mass at the startup of the RCP was approximately 300 kg 6 Science and Technology of Nuclear Installations [13].At the startup of the RCP, the CINEMA calculations of cumulative hydrogen production are higher than 300 kg.However, the CINEMA results and measured total hydrogen generation mass are very similar, which were approximately 465 kg and 460 kg [13], respectively.Hydrogen production did not occur after 10,500 s because the intact fuel rods and some metallic cladding were quenched by the HPI water.

Conclusions
Te severe accident progression in TMI-2 has been analyzed as a part of a validation of the CINEMA computer code, which is capable of simulating core melt progression with thermal hydraulic analysis of the RCS and severe accident analysis of the containment.Te CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data.Te CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data in general.Te upper part of the fuel rod melts and relocates to the middle part of the core, which is very similar to the melt pool formation in the real TMI-2 data.Te CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively.Te CINEMA results after molten pool formation and quenching at 10,440 s is not similar to the TMI-2 data because of complex quenching phenomena in the core.For this reason, further CINEMA model development and analysis for a molten pool formation with crust formation and quenching process in the core and lower plenum of the reactor vessel are necessary to simulate all TMI-2 severe accident progression.

Figure 1 :
Figure 1: Structure of CINEMA computer code.

Figure 4 :
Figure 4: Comparison of CINEMA results with TMI-2 data on the pressurizer pressure.

Figure 5 :
Figure 5: CINEMA results on mass fow rate through pressurizer PORV of the TMI-2 severe accident.

Figure 6 :
Figure 6: CINEMA results on collapsed water level in pressurizer of the TMI-2 severe accident.

Figure 7 :Figure 8 :Figure 9 :
Figure 7: CINEMA results on collapsed water level in core of the TMI-2 severe accident.

Figure 12 :
Figure 12: Comparison of CINEMA results with TMI-2 data on the generated hydrogen mass.

Table 1 :
Detailed main events in TMI-2 severe accident.

Table 2 :
Comparison of CINEMA results with the TMI-2 operating condition.