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The 9TH European Review Meeting on Severe Accident Research (ERMSAR2019) Log Number: 084
Clarion Congress Hotel, Prague, Czech Republic, March 18-20, 2019
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UPDATED SAMGs IMPACT ON A FUKUSHIMA DAIICHI UNIT 2
ACCIDENT SIMULATION WITH MELCOR 2.1
R. Bocanegra, G. Jiménez, and A. Carlero
Universidad Politécnica de Madrid
r.bocanegra@upm.es
gonzalo.jimenez@upm.es
ABSTRACT
After the Fukushima accident in March 2011, the nuclear community begun to review all the severe accident
management guidelines, also known as SAMGs, to find a way in order to avoid the release of radioactive
material to the environment as happened during these unfortunate days in Japan. This study pretends to
highlight the response of the Fukushima Daiichi Unit 2 nuclear plant if some of these improved strategies
were implemented during the accident. In particular, a new SAMG for boiling water reactors proposed by
Taiwan Power Company, and the actions performed during the accident management in Fukushima Daini
were implemented in a MELCOR model and compared between them and against the registered evolution
of the Fukushima Daiichi Unit 2 accident. The study shows that the strategies proposed are a priori effective
avoiding the vessel failure and the release of radioactive material to the environment, but modifying some
of the hypothesis assumed in the proposed SAMGs, such as the alternative water injection flow rate, and
also the operation pressure range for the RCIC system.
KEYWORDS
Fukushima Accident; Severe Accident; SAMGs; MELCOR
1. INTRODUCTION
On March 11 2011, at 14:41, the most powerful earthquake ever recorded hit Japan. The earthquake
triggered powerful tsunami waves in Miyako (Tōhoku's Iwate Prefecture) and travelled up to 10 km inland
in the Sendai area [1]. The earthquake provoked the loss of almost all electrical lines in the north region of
Japan. Around 50 minutes after the earthquake, a tsunami series hits the north-east coast of Japan. The most
affected NPP was the Fukushima Daiichi NPP, close to the coast line. The plant consists of six Boiling
Water Reactors (BWRs), originally designed by General Electric (GE) and operated by Tokyo Electric
Power Company (TEPCO). Units 2 to 6 were of BWR-4 type, while Unit 1 was an BWR-3 design. At the
time of the earthquake, Reactor 4 had been de-fueled for shroud replacement and refueling operations,
whereas Reactors 5 and 6 were in cold shutdown for planned maintenance. Immediately after the earthquake,
reactor SCRAM was automatically actuated in the remaining units (1-3) due to the earthquake signal. The
Electrical Diesel Generators (EDGs) were automatically activated in order to provide AC power to the safety
systems. With the tsunami arrival, a 14 meters high wave overwhelmed the plant's seawall, which was only
10 meters high. The sea water quickly flooded the low-lying rooms in which the EDGs were housed, which
inevitably failed leading to a Station BlackOut (SBO). Over the coming days, core meltdown occurred and
a significant amount of radioactive materials were released into the environment.
To deal with this accidental situation and prevent or mitigate the impacts of accident in NPPs, the accident
management procedures and guidelines have been developed to help the operator crew to act during an
accident. Depending on the severity of the accident, there are mainly two types. Firstly, there are the
Emergency Operating Procedures (EOPs), which are symptom-based rules used to deal with Design Basis
Accident (DBAs), and some others beyond design basis, where its principles can be consulted in [2]. And
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secondly, the Severe Accident Management (SAM), which were introduced in 1990 [3] with the
development of the Severe Accident Management Guidelines (SAMGs) motivated by the fact that the Three
Mile Island NPP´s EOPs did not avoid the core damage conditions that were reached.
After the Fukushima incident, there was a need to update and improve the SAMGs due to the unexpected
conditions that were reached in the accident. As a consequence, all countries around the globe have been,
or are being, reviewing its SAMGs [4]. And one question that could be interesting to answer is “Would be
effective these improved SAMGs if they were applied in Fukushima Daiichi in 2011?”
Trying to answer this question, in this paper the Fukushima Unit 2 (1F2) accident is analyzed applying two
of the revised SAMG strategies published during the last years: the Ultimate Response Guideline (URG)
proposed by Taiwan Power Company in [5], and the strategy followed by the Fukushima Daini NPP, which
achieved cold shutdown with no major damage [6]. This same exercise had been realized by [7] applied to
a Chinsang NPP TRACE evaluation model and its results will be compared to that obtained in this studies.
In the paper, firstly the 1F2 event sequence is reviewed and emergency equipment and instrumentation
availability determined. Then, the MELCOR Evaluation Model (EM) employed in the analysis will be
briefly described remarking some differences in modeling hypothesis assumed in [7], followed by a
depiction of the two strategies tested. Later on, a comparison between the calculation results from the two
cases and the Chinsang NPP study will be discussed, finishing with some conclusions.
2. THE FUKUSHIMA DAIICHI UNIT 2 ACCIDENT EVENT SEQUENCE
The accident sequence used in this paper has been mainly taken from the latest Fukushima accident reports
released by TEPCO in the information portal [8] and from the IAEA Fukushima report [9].
After the earthquake, the 1F2 reactor was shut down automatically as expected, and due to the loss of the
off-site power, the EDGs started up to assure the availability of the safety systems. The Reactor Core
Isolation Cooling (RCIC) system was switched on (14:50) to keep the reactor cooled but was automatically
shut down one minute later (14:51) due to the activation of the L-8 Reactor Pressure Vessel (RPV) liquid
level signal. A few minutes later, the RCIC was online again (15:02) and kept operating until the arrival of
the first tsunami wave at 15:27. Just one minute later, the operator logs indicate that the RCIC stopped once
more (15:28) due to the high liquid level in the RPV instead of the tsunami flood, unlike it could be expected.
At 15:35, a second tsunami wave arrived and as a consequence, all the AC and most of the DC power were
lost. The RCIC was verified to be online at 15:39 but, with the loss of the batteries, the steam throttle valve
failed at a full open position allowing the RCIC working without any control for almost three days. When
the liquid level reached the Main Steam Lines (MSLs), the RCIC turbine started working in a two phase
condition reducing the system performance.
March 14 at 11:01, the Unit 3 Reactor Building suffered a hydrogen deflagration, and it is believed that the
explosion damaged the Unit 2 reactor building, opening a breach on it. At 13:25, the RCIC automatically
shut down by unclear reasons, and therefore, the RPV pressure started rising. A new plan was carried out to
fix an alternative injection way using a firefighter engine motor pump to inject water into the RPV. But
before it could be done, the RPV had to be depressurized to allow the water injection. At 18:02, the Safety
Relief Valve (SRV) A was manually opened allowing the RPV depressurization. At 19:54, the operators
began to inject sea water into the RPV.
March 14 at 6:14, the Unit 4 explosion damaged the Wet-Well (WW) pressure sensor disconcerting the
operators who believed that the WW was massively failed releasing radioactive material to the environment.
Later on, the WW fail was discarded due to the high Dry-Well (DW) pressure. Moreover, at 8:46, a DW
major leak was produced as a result of the high pressure.
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3. MELCOR EVALUATION MODEL
The Evaluation Model (EM) used for the analysis is an evolution which begins with the State of the Art
Reactor Consequence Analysis (SOARCA) project [10], supported by the U.S. NRC and carried out by
Sandia National Laboratories. Under the cited program, a BWR/4 Mark I EM was developed based on the
Peach Bottom NPP with the MELCOR code [11]. Some years later, T. Sevón, from the Technical Research
Centre of Finland (VTT), using the data from the Fukushima web portal [8] and the Peach Bottom adapted
EM developed by Sandia, released a study referent to the Fukushima Unit 3 accident [12], which included
the MELCOR input file. This EM released by T. Sevón was then adapted for a 1F2 accident analysis with
MELCOR 2.1 by the Nuclear Safety group at the UPM in collaboration with the KIT (Germany) [13]. This
last EM is the basis for the study presented in this paper.
This base EM has been further evolved updating the RPV, Primary Containment Vessel (PCV), and core
following the guidelines shown in the SOARCA report [14]. In relation with the CVH package in MELCOR,
the active core region has been modeled with 40 Control Volumes (CVs) and 83 Flow Lines (FLs), as can
be seen in Figure 1. The core channels are represented with 20 CVs, and the other 20 represents the bypass
region, the flow area between fuel elements. Every FL is set with specific block options depending on the
CVs which are linked: channel-axial channel; bypass-axial bypass; bypass-radial bypass; channel-radial
bypass. This is for representing the coolant flow and the corium movement when the core starts to degrade.
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Clarion Congress Hotel, Prague, Czech Republic, March 18-20, 2019
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Figure 1: 1F2 RPV Scheme with SNAP
The Core (COR) package includes the other options for defining the core behavior during an accident. The
RPV is subdivided in 5 radial rings and 16 axial levels, being 4 radial rings and 10 axial levels for
representing the active core region. Initial masses and geometry have been maintained from the Sevon’s
EM. The B4C control element model has also been included to include the B4C-SS eutectic effect during
the core degradation. The fuel cladding failure is defined using a time-temperature table as recommended
in the SOARCA report. The lower plenum is represented with 7 segments, as it is the minimum required for
the intersections with the COR radial rings and axial levels stablished. Every segment is also subdivided in
10 calculation points, being the node 1 the RPV external side and the node 10 the inner side. The lower head
penetration failure is stablished in 1273.15 K, the default value of MELCOR 2.1. The Heat Transfer
Coefficients (HTCs) are set to default values.
The Reactor Coolant System (RCS) has been modeled with 18 CVs, 26 FLs and 37 Heat Structures (HSs).
The lower plenum is represented with 5 of the CVs employed, 1 for the downcomer, 1 for the jet pumps, 2
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for each recirculation loop, 1 for the steam separator, 1 for the steam dryer 1 for the upper plenum, and 1
for each steam line. The FLs employed also includes the connections for representing the recirculation
pumps and Safety Relief Valves (SRVs) leaks, and an eventual RPV failure.
The containment is composed by the Dry-Well (DW), the Wet-Well (WW), and the connection between
them through the Vent System (VS) and the Vent Line (VL). For representing it, it has been employed 5
CVs for the DW, 8 for the WW, 1 for the VS, and another one for the VL, as can be seen in Figure 2.
Figure 2: 1F2 PCV Scheme with SNAP
The safeguards have been maintained as in [13] except the Reactor Core Isolation Cooling (RCIC) system,
which has been modified to avoid the use of time-dependent flow tables (Figure 3). The RCIC control logic
now includes a pressure operation range (0.8 - 7.170 MPa) for the RCIC turbine following the data from
[15]. But, since the turbine discharge pressure is around 0.35 MPa [16], it becomes difficult to state that the
RCIC turbine could operate at nominal capacity with a differential pressure of 0.4 MPa. Therefore, it was
set a minimum operation pressure of 1 MPa. The steam driven through the turbine is extracted via a mass
external source located in the Steam Line (SL) CV. The energy carried by this steam is also extracted from
such CV making use of an energy external source. The extracted mass, around 2 kg/s at nominal regime
[13], is then added using another external source in the discharge CV, located in the WW. The energy
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introduced in such CV is calculated by the difference of the extracted steam energy and the power developed
by the turbine, around 875 horsepower [17]. The nominal mass flow pumped to the RPV is maintained to
20 kg/s. A RPV liquid level trip is also included to stop the RCIC when it reaches 5.653 m above the Top of
Active Fuel (TAF), which corresponds to the L-8 level in 1F2. The RCIC operation is further reinitiated
when the RPV liquid level becomes lower than TAF + 3.253 m. The switch between the Condensate Storage
Tank (CST) and the WW water suction is configured to be produced when the CST level reach the 20% of
its capacity. The alternative water injection is controlled by a Control Function (CF) which is pressure-
dependent, as in [13].
Figure 3: 1F2 RCIC Scheme with SNAP
It has to be remarked that the alternative water injection was modeled in the 1F2 MELCOR EM using a
specific pump curve for a W.S. Darley & Co. Pump Model 2BE10YDN-N (Figure 4), a motor-pump similar
to that used by the firefighter in Fukushima during the accident. On contrary, in the Chinsang NPP studies
[7], [18], it was assumed that the alternative water was injected at a constant rate of 50 kg/s when the RPV
pressure drops below 0.52 MPa.
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Figure 4: Darley‘s Pump Performance Curve
4. ULTIMATE RESPONSE GUIDELINE (URG)
The Ultimate Response Guideline (URG) was developed by Taiwan Power Company [19] to deal with
situations similar to that occurred in Fukushima in 2011, that is, an event causing a SBO with no water
supply for reactor cooling. The procedure is the next one:
1. Perform a controlled depressurization to bring down steam pressure to 1.45 MPa by regulating the SRVs.
To assure the steam driven passive cooling mechanisms available (RCIC), the system pressure needs to
be maintained higher than 0.8 MPa (approx.) since, as was commented before, the RCIC turbine is
designed to operate at nominal regime between 0.8 - 7.170 MPa, which is actually plant-dependent.
Therefore, the minimum pressure considered in the analysis for the RCIC operation is set to 1.0 MPa.
This regime is maintained until the RCIC failure, and assures that when it occurs, an eventual emergency
depressurization does not flash all the coolant wherein the RPV. It is estimated that if a depressurization
occurs at around 7 MPa, the coolant mass flow through the SRVs will be around 100 kg/s. On contrary,
if depressurization occurs at 1.45 MPa, the coolant mass flow though the SRV will be around 20 kg/s.
This measure could avoid the core uncovering during the depressurization process.
2. The URG suggest to makes available the alternative water supply which might include raw or sea water
powered by gravity, or any non-designed pumps, within the first hour.
Therefore, according to the 1F2 events, the alternative water injection available was the water supplied
by the fire engines motor-pumps, which will be considered online 1 hour later the SCRAM occurred.
3. To equalize the RPV and containment pressures, the Automatic Depressurization System (ADS) is
activated, before the RCIC becomes inoperable, or when the support equipment for injecting alternative
water becomes placed.
In case of 1F2, the ADS will be represented as the instantaneous opening of all the SRVs at the same
time. The emergency depressurization will be produced 1 hour later than the beginning of the SBO status,
when it is assumed that the fire fighters arrived to the location.
4. Inject raw or sea water into the reactor after the system fully depressurizes.
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5. Perform containment direct venting if containment pressure is beyond design to maintain containment
integrity.
It estimates a minimum time required to be sustained by the plant of 36 h. After that, support equipment will
be considered available, which in fact didn’t happened in Fukushima during the accident.
5. FUKUSHIMA DAINI NPP STRATEGY
During the Fukushima Daini NPP accident, the strategy was a series of controlled depressurizations every
1 hour while the RCIC was in operation [6]. The SRV opening was maintained until the cladding temperature
dropped 55 K to avoid RPV damage due to fast cooling. Then, the operators waited 1 hour and proceeded
to the following SRV opening. The plan was the injection of alternative water by using the condensate water
transfer pumps, which had a discharge pressure estimated in around 0.7 MPa and a nominal mass flow of
50 kg/s.
For 1F2 analysis, the SRV logic was modified in the MELCOR EM to adapt it to the strategy followed in
Fukushima Daini NPP. The SRV-A is configured to be opened each hour from the SBO status, and to be
closed when the vessel temperature drops 55 K. This process is repeated three times, maintaining the SRV
opened indefinitely after the third sequence.
The alternative water injection (fire fighter motor-pumps) is assumed to be available 1 hour after the SBO
status.
It has to be accounted that during the Fukushima Daini incident, one of the EDGs remained operative,
allowing to set an AC line to power some systems to the different units.
6. RESULTS AND DISCUSSION
Both strategies above depicted are analyzed and compared between them. The RPV pressure evolution for
the case, where the URG is applied, is shown in Figure 5. The first depressurization keeps the pressure
around 1.5 MPa to maintain the RCIC working until the alternative water line is ready. When it happens, a
second depressurization using all the SRVs is performed to equals the RPV pressure to containment pressure
(around 0.42 MPa). It allows the firefighters injecting water with a motor-driven pumps. As can be seen,
the pressure curve is identical to that obtained by Wu et al. [7] with the unique difference that for the
Chinsang NPP analysis, the pressure ends at almost 0 MPa, probably due to the containment pressure set in
the TRACE code, which is not able to correctly model the containment building.
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Figure 5: URG RPV Pressure & Injected Water
In the Figure 6 it is shown the RPV liquid level. The Top of Active Fuel (TAF) is lightly uncovered during
a short time period, and then the liquid level is rapidly recovered due to the alternative water injection. The
liquid levels (L-2 and L-8) sensors which control the water injection within the RPV are assumed offline,
as happened during the 1F2 accident. On contrary, in order to control the RCIC mass flow injected in the
RPV, the RCIC turbine throttle valve was assumed available, as well as the SRV batteries for maintaining
the pressure at around 1.5 MPa. These are in fact strong hypotheses that lead to a controlled reactor state,
which could be cooled indefinitely while the alternative water injection availability is assured.
Figure 6: URG RPV Liquid Level & Injected Water
The cladding temperature drops drastically with the first manual depressurization, as can be seen in Figure
7. Then, it is maintained around 480 K, showing an oscillatory behavior due to the SRV operation to keep
the pressure as constant as possible. During the emergency depressurization, clad temperature drops again
due to the RPV water evaporation, and then keeps dropping slowly as a consequence of the alternative water
injection.
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Figure 7: URG RPV Liquid Level & Injected Water
The DW pressure (Figure 8) experiments a drastic increase due to the RPV depressurization, reaching
around 650 kPa. It is plotted along with the TEPCOs DW measurements during the accident for comparison
reasons. The design pressure of these type of containments is around 392 kPa [20], but it does not mean that
the containment will fail if this pressure is surpassed. In fact, containment failure modes applied in this
MELCOR evaluation model are the same than that applied in [13], so it means that if pressure reaches 710
kPa, the DW head flange will start leaking through a 0.023 cm2 aperture, and if WW pressure surpasses 1.2
MPa, a break of around 2 cm2 will occurs. Since neither of these criteria has been accomplished when the
URG is applied, it is considered that 1F2 containment would hold such pressure excursion as a consequence
of the RPV depressurizations. However, during these 24 hours of transient analyzed, the DW pressure
slightly increases in a constant manner, because of the alternative water injection, which removes the core
residual heat, and it ends in the WW. Therefore, a longer analysis should be performed to study the
containment behavior, but it is not irrational to anticipate that if there are no means for removing the residual
heat from the containment, it will probably fail at the end.
Figure 8: URG Drywell Pressure
When the Fukushima Daini strategy is applied to the 1F2 accident, the first RPV depressurization is
performed by opening the SRVA until the clad temperature drops 55 K. When this temperature decrease is
reached, the SRV is closed during one hour and then opened again to until the clad temperature drops another
55 K. This sequence is performed three times, but as can be seen the RPV pressure (Figure 9) does not drop
as low as it happened in the URG case. From the second SRV depressurization, the pressure drops to levels
where the RCIC is not capable of working, since it was estimated to works between a steam pressure range
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of 1-7.5 MPa, but it is still high enough to not allow the alternative water pump injecting water into the
RPV.
Figure 9: Daini RPV Pressure & Injected Water
As can be seen in Figure 10, the RPV pressure does not decrease enough to allow the alternative water
pump injecting with its nominal flow rate. Consequently, a partial core uncovering is produced, lately than
the URG case, but the liquid level is never recovered. That means that perhaps, only one SRV is not enough
to depressurize the RPV when this strategy is employed, at least for the last depressurization, since the RPV
pressure does not drop enough to allow the firefighter motor-pumps injecting alternative water. In
Fukushima Daini were employed the condensate water transfer pumps, which were capable of injecting
water at 0.7 MPa with a mass flow rate of 50 kg/s. However, in Fukushima Daiichi, the pumps available
were the less powerful firefighter motor-driven pumps, and not the condensate water transfer pumps as in
Fukushima Daini, which led to a different result.
Figure 10: Daini RPV Liquid Level & Injected Water
The clad temperature behavior (Figure 11) shows the expected stepped decrease in 55 K intervals,
corresponding in time with the RPV depressurization operation. When the last SRV opening is performed,
the clad temperature drop becomes much less effective than in the previous ones since the evaporation is
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Daini RCIC
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considerable reduced due to the low water level in the RPV. In fact, after 5.5 hours, the clad temperature
start increasing again due to the uncovering, and at the end, after 17 hours, the core partially meltdown.
Figure 11: Daini Clad Temperature
The Daini-case containment pressure (Figure 12) follows a trend that results quite similar to that observed
in the URG case (Figure 8) during the first 6 hours, but that, after that time, it raises until reaching the
failure criteria assumed in the 1F2 MELCOR evaluation model.
Figure 12: PCV Pressure
7. CONCLUSIONS
The Ultimate Response Guideline, a SAMG for boiling water reactors proposed by Taiwan Power Company,
and also the actions performed during the accident management in Fukushima Daini in March 11, 2011,
were implemented in a Fukushima Daiichi Unit 2 MELCOR evaluation model. The response of the plant
was analyzed for both cases, accounting for the events occurred in Fukushima Daiichi, with the exception
that the DC power for governing the RCIC turbine and the SRV operation was assumed available. Results
have shown that the strategy of reducing the RPV pressure is very effective for limiting the loss of core
cooling. It allowed the operation of the RCIC system until the alternative water injection was ready.
However, this system is dependent on the availability of DC power, which in fact it was not available during
the Fukushima Daiichi Unit 2 accident. On contrary, the Fukushima Daini strategy does not result as
effective as the URG. The opening of only one SRV resulted not enough to neither, drop the RPV pressure
567
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Time (h)
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Daini
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to levels where the motor- pumps employed by the firefighters in Fukushima Daiichi were able to inject
water, nor to keep operating the RCIC system.
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