updated samgs impact on a fukushima daiichi unit 2 ... 2019 - final papers/84. bocane… · power...

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

[email protected]

[email protected]

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

mailto:[email protected]

mailto:[email protected]

https://en.wikipedia.org/wiki/Miyako,_Iwate

https://en.wikipedia.org/wiki/Boiling_water_reactor

https://en.wikipedia.org/wiki/Boiling_water_reactor

https://en.wikipedia.org/wiki/Earthquake

https://en.wikipedia.org/wiki/Shutdown_(nuclear_reactor)

https://en.wikipedia.org/wiki/SCRAM



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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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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.

0

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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

Daini Alt. Water



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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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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.

REFERENCES

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[16] “Dresser Rand. Steam Turbine, Motor and Generator D. Terry Nuclear Drive Turbines. E-mail

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[19] K. S. Liang, S. C. Chiang, Y. F. Hsu, H. J. Young, B. S. Pei, and L. C. Wang, “The ultimate

emergency measures to secure a NPP under an accidental condition with no designed power or

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updated samgs impact on a fukushima daiichi unit 2 ... 2019 - final papers/84. bocane… · power...

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