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Insights from review and analysis of the Fukushima

Dai-ichi accidentMasashi Hirano

a, Taisuke Yonomoto

a, Masahiro Ishigaki

a, Norio Watanabe

a, Yu

Maruyamaa

, Yasuteru Sibamotoa

, Tadashi Watanabea

& Kiyofumi Moriyamaa

aNuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirane, Shirakata,

Tokai-mura, Naka-gun, Ibaraki, 319-1195, Japan

Version of record first published: 24 Jan 2012.

To cite this article: Masashi Hirano , Taisuke Yonomoto , Masahiro Ishigaki , Norio Watanabe , Yu Maruyama , YasuteruSibamoto , Tadashi Watanabe & Kiyofumi Moriyama (2012): Insights from review and analysis of the Fukushima Dai-ichi

accident, Journal of Nuclear Science and Technology, 49:1, 1-17

To link to this article: http://dx.doi.org/10.1080/18811248.2011.636538

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INVITED REVIEW Fukushima NPP Accident Related

Insights from review and analysis of the Fukushima Dai-ichi accident

Masashi Hirano*, Taisuke Yonomoto, Masahiro Ishigaki, Norio Watanabe, Yu Maruyama, Yasuteru Sibamoto,

Tadashi Watanabe and Kiyofumi Moriyama

Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura,Naka-gun, Ibaraki 319-1195, Japan

(Received 5 September 2011; accepted final version for publication 26 September 2011 )

An unprecedented earthquake and tsunami struck the Fukushima Dai-ichi Nuclear Power Plants on 11

March 2011. Although extensive efforts have been continuing on investigations into the causes andconsequences of the accident, and the Japanese Government has presented a comprehensive report on the

accident in the IAEA Ministerial Conference held in June 2011, there is still much to be clarified on what

happened during the accident and why. This article aims at identifying what should be clarified further

about the progression of the accident at Units 13 through the review and analysis of information released

from Tokyo Electric Power Company and government authorities. It also discusses the safety issues raised

by the accident based on the insights gained, in order to contribute to establishing a new framework that

pursues continuous improvement toward the highest standards of safety that can reasonably be achieved.

Keywords: Fukushima Dai-ichi accident; accident management measures; alternative water injection;

containment integrity; venting; defense in depth; design basis; beyond design basis; external events;

continuous improvement; highest standards of safety

1. Introduction

Approximately three weeks after the accident at the

Fukushima Dai-ichi Nuclear Power Plants (hereinafter,

referred to as F-1NPPs) of Tokyo Electric Power

Company (TEPCO), the Nuclear and Industrial Safety

Agency (NISA) made a presentation at a side event on the

accident during the Fifth Review Meeting of the Conven-

tion of Nuclear Safety, co-sponsored by Japan and the

International Atomic Energy Agency (IAEA), on 4 April

2011 and reported the overall plant behavior during the

accident to that time to the international community [1].

On 25 April 2011, the NISA requested TEPCO to report

the plant records and other relevant information pur-

suant to the Law for the Regulation of Nuclear Source

Material, Nuclear Fuel Material and Reactors (Nuclear

Regulation Law), and Electricity Utilities Industry Law

[2]. On 16 May 2011, TEPCO reported to the NISA [3]

and at the same time released a lot of information

including the plant records on its Webpage at http://

www.tepco.co.jp/en/nu/fukushima-np/index-e.html. On

receiving the report, the NISA further requested

TEPCO to provide additional information which could

supplement the report [4]. On 24 May 2011, TEPCO

submitted the supplemental information to the NISA

[5], which then issued its evaluation report [6].

The Nuclear Emergency Response Headquarters of

the Government presented a comprehensive report on

the accident [7] at the IAEA Ministerial Conference on

Nuclear Safety held atthe IAEA from20 Juneto 24June.

The IAEA fact finding expert mission for the accident

visited Japan from 24 May to 2 June and presented its

report [8] also in this meeting. TEPCO conducted ana-

lyses of plant behaviors at Units 13 with the Modular

Accident Analysis Program (MAAP) codeand the results

were included in the Governments report. On 18 June

2011, TEPCO also released a report that included the

description of operator actions taken to cope with the

accident in its early phase [9].

In the US, the near-term task force of the Nuclear

Regulatory Commission (NRC) issued its report on 12

July 2011 [10]. Although the task force made recom-

mendations especially on the regulatory framework for

the US NRC, the discussion is insightful and part of it

is thought to be commonly applicable to Japan.

In this article, based on the reports mentioned above,

first, the basic information on the damages to the F-

1NPPs by the earthquake and tsunami are summarized.

Then, the plant records during the accident at Units 13

are analyzed to identify key operator actions or physical

phenomena that are not fully understood yet. Finally,

Journal of Nuclear Science and Technology

Volume 49, No. 1, January (2012) pp. 117

http://www.tandfonline.com

*Corresponding author. Email: [email protected]

ISSN 0022-3131 print/ISSN 1881-1248 online

2012 Atomic Energy Society of Japan. All rights reserved.

http://dx.doi.org/10.1080/18811248.2011.636538
http://www.tepco.co.jp/en/nu/fukushima-np/index-e.htmlhttp://www.tepco.co.jp/en/nu/fukushima-np/index-e.htmlhttp://www.tepco.co.jp/en/nu/fukushima-np/index-e.htmlhttp://www.tepco.co.jp/en/nu/fukushima-np/index-e.html


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based on the insights gained from the analysis, some key

safety issues raised by the accident are discussed.

Since this article focuses on behaviors of reactors at

Units 13 during the early phase of the accident when

radioactive materials were released significantly, other

matters such as behaviors of spent fuel pools and on-site and off-site emergency responses are not discussed.

2. Damages by earthquake and tsunami and plant

behavior

2.1. After the earthquake and before the tsunami

A summary of basic plant information on the F-

1NPPs is shown in Table 1 [1,7]. Each of the Units 15

has a Mark-I primary containment vessel (PCV)

consisting of a dry well (D/W) and a suppression

chamber (S/C), whereas Unit-6 has a Mark-II PCV.

The observed seismic acceleration was mostly

smaller than that of the design basis earthquake Ssof the site. However, at Units 2, 3, and 5, the maximum

acceleration observed for the basemat of the reactor

building in the east-west direction was larger than the

maximum response acceleration to Ss, especially in the

frequency range of 0.20.3 s, by approximately 30% at

the highest. Units 13 scrammed during 14:4647 Japan

Standard Time on high seismic acceleration and almost

simultaneously a loss-of-offsite power (LOOP) took

place. At that time, six out of seven offsite power lines

were connected to the site (one line had been taken out of

service for maintenance) and all of them were lost due to

damages of the breakers at switchyards or other causesincluding the collapse of the steel tower (or pylon) of the

transmissionlines to Units 5 and 6 due to a landslide of a

nearby slope. A total of 12 out of 13 emergency diesel

generators (EDGs) automatically started (one at Unit 4

had been taken out of service for periodic inspection).

After the scram, at Unit 1, the isolation condensers

(ICs) shown in Figure 1 [7] were automatically actuated

on high reactor pressure at 14:52 but were shutdown

manually by closing the motor operated valves

(MOVs), MO-3A and MO-3B, at about 15:03. This

was done according to the operating manual to preventthe cooling rate of the reactor pressure vessel (RPV)

from exceeding 55 K/h [7]. After that, the operators

decided to use only Train A and similar automatic

actuation and manual shutdown were supposedly

repeated three times during 15:17 to 15:34. Then, the

tsunami struck and both alternating current (AC) and

direct current (DC) power supplies were lost at 15:37,

resulting in the valves being inoperable. When the

power supplies were lost, the valves on the ICs

remained as is, that means the MO-3A and MO-

3B remained closed and the others remained open.

The high pressure coolant injection (HPCI) system

with a steam turbine driven pump was not actuated atany of the units because the reactor water level did not

decrease to the set point for its actuation (L-2).

At Units 2 and 3, the reactor core isolation cooling

system (RCIC) was started up automatically at 14:50

and 15:05, respectively, on low reactor water level (L-2).

2.2. Just after the tsunami

Approximately 40 min following the earthquake, at

15:27, the first major tsunami arrived and at 15:35 the

second one; this was followed by multiple additional

waves. The ground level of the site is 10 m above sea

level and the tsunami reportedly reached 45 m aboveground level indicating the inundation height was 14

15 m. It should be noted that the ground level of Units

5 and 6 is about 13 m.

Due to the tsunami, a total of 12 out of 13 EDGs

on-site became inoperable resulting in station blackout

Table 1. Specifications summary of units in F-1NPPs [1,7].

Unit Unit 1 Unit 2 Unit 3 Unit 4 Unit 5 Unit 6

Reactor type BWR-3 BWR-4 BWR-4 BWR-4 BWR-4 BWR-5PCV type Mark-I Mark-I Mark-I Mark-I Mark-I Mark-IIPower output

(MWe gross/net)

460/439 784/760 784/760 784/760 784/760 1100/1067

Contractor GE GE, Toshiba Toshiba Hitachi Toshiba GE, ToshibaFirst commercial

operation(year, month)

1971.3 1974.7 1976.3 1978.10 1978.4 1979.10

Max. pressure ofRPV (MPa)

8.24 8.24 8.24 8.24 8.62 8.62

Max. temperatureof RPV (8C)

300 300 300 300 302 302

Max. pressure ofPCV (MPa)

0.43 0.38 0.38 0.38 0.38 0.28

Max. temperatureof PCV (8C)

140 140 140 140 138 171(D/W) 105(S/C)

No. of EDGs (watercooled/air cooled)

2/0 1/1 2/0 1/1 2/0 2*/1**

Plant status priorto accident In operation In operation In operation Refueling outage Refuelingoutage Refueling outage

Note: *One of them is dedicated to the HPCS; **one was functional after the tsunami.

2 M. Hirano et al.


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(SBO) at Units 14. The air cooled one at Unit 6 that

survived the tsunami supplied power to maintain the

cooling capabilities of the cores and the spent fuel

pools at both Units 5 and 6, since the interconnection

of emergency busbars between two units (namely,

Units 5 and 6, as well as Units 1 and 2, and Units 3 and

4) had been implemented as an accident management

(AM) measure to allow the electric power from the

EDG at one unit being used at the other unit.

There were a total of three air-cooled EDGs two

were inoperable while the one at Unit 6 was operable;

these had been additionally installed in 1994 as part of

AM measures [11]. For Units 5 and 6, there were

originally four water-cooled ones among which one at

Unit 6 was dedicated to the high pressure core spray

system (HPCS) and another one had been shared by

both units. These EDGs were located in the basement

of the respective turbine buildings. The air-cooled one

which functioned after the tsunami had been installedfor Unit 6 on the first floor of the EDG building. For

Units 1 and 2, there were originally only three water-

cooled ones and one was shared by both units. This

situation was the same for Units 3 and 4. Then, an air-

cooled one was installed for each of Units 2 and 4 on the

first floor of the building called the Common Pool for

storing spent fuel. They survived the tsunami but could

not supply power because the metal-clad switchgears,

located in the basement level of their respective turbine

buildings, were damaged by the tsunami. It should be

noted that all the metal-clad switchgears were not

functioning except those at Unit 6.

A total of 12 seawater pumps for residual heat

removal (RHR) for six units, installed outside without

walls, were submerged under seawater, resulting in loss

of the ultimate heat sink (UHS) in all the units.

At Unit 1, the ICs were inoperable after the

tsunami because the MO-3A and MO-3B had been

closed as already mentioned. The HPCI also became

inoperable due to loss of DC power [7].

At Units 2 and 3, the RCICs were operatedfor many hours after the tsunami as is discussed in

Section 4.

Figure 1. Schematic of ICs.

Journal of Nuclear Science and Technology, Volume 49, No. 1, January 2012 3


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3. Accident management measures applied

On 28 May 1992, the Nuclear Safety Commission

(NSC) recommended the licensees voluntarily imple-

menting the AM measures [12,13]. Responding to this

recommendation, the licensees had proposed andimplemented the measures, and submitted their report

to the NISA in May 2002, with the results from a

probabilistic safety assessment (PSA) for internal

events for verifying their effectiveness. The NISA

evaluated the submittals from the individual licensees

and compiled the evaluation report in October 2002,

which was reported to the NSC [14].

An example of the measures taken to prevent severe

accidents (SAs) and actually applied in the F-1NPPs

accident is the alternate water injection into the core

shown in Figure 2 [7], where additional pipe lines and

valves were installed to allow the fire protection system

(FPS) and make-up water condensate (MUWC) systemto inject water into the core and PCV through the

emergency core cooling system (ECCS) injection lines.

In addition, a procedure was implemented to reduce

the RPV pressure by opening the safety relief valves

(SRVs), shown in Figure 3 [7], to allow the low pressure

pumps to inject water into the core. Although the use

of fire engines was not part of the measures, they were

actually used in the accident. As is discussed in Section

4, it was difficult to open the SRVs [9], which are air-

operated valves (AOVs), because the operators need

both DC power and compressed air from the instru-

ment air (IA) system.

A second example is what is called wet venting

or hardened venting shown in Figure 4 [15], toprevent the PCV failure from overpressurization in

the case of RHR loss, where a pressure-resistant pipe

line was installed to connect the venting line starting

from the top of the S/C to the exhaust line of the

standby gas treatment system (SGTS) inside the

stack. On this line, there are two AOVs (large and

small) in parallel, though only one is shown in

Figure 4, a MOV and a rupture disc to prevent

inadvertent releases of radioactive materials. In the

accident, the MOV was opened by using its

handwheel and the AOVs were opened by connecting

mobile batteries for opening the solenoid valves, and

air cylinders or mobile compressors for compensatingfor the loss of compressed air. Moreover, the

operators had to repeatedly apply the air cylinder

or mobile compressor since the AOV self-returned to

its closed position due to spring force when the

compressed air was lost [9].

It should be noted that electricity is assumed to be

available for opening the SRVs and valves for wet

venting in TEPCOs AM measures; this is certainly

Figure 2. Schematic of alternate water injection implemented as AM measures and actually applied.

4 M. Hirano et al.


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Figure 3. SRVs and SVs for Units 2 and 3.

Figure 4. Schematic of hardened venting implemented as one AM measure for BWR 4.



7/18

one of the main causes for difficulties to open them in a

timely manner during the accident as discussed later.

4. Accident progression at units 13 after the tsunami

4.1. Unit 1

Figure 5 shows the RPV pressures, D/W pressure,

and volumetric flow rate of either freshwater or

seawater injection into the core. The data such as

pressures and reactor water level taken using the

differential pressure transducer should be carefully

interpreted because the water in the pressure sensing

line might have boiled away as the PCV temperature

rose. The first data point of D/W pressure was 0.6 MPa

at 1:05 on 12 March which had already exceeded its

maximum design pressure (Pd). Although the steam

generated in the core was released to the S/C through

the SRVs as shown in Figure 3, steam alone could not

bring about such a large pressure increase in the D/Wsince the water in the S/C still remained subcooled at

this time. On the other hand, if it is assumed that the

ICs were unavailable after the tsunami, the fuel heat-up

and subsequent zirconium-water reaction would have

started to occur within about 2 h. Then, the mixture of

steam and hydrogen would have been transported to

the PCV through SRVs. Therefore, it is reasonable to

presume that the accumulation of hydrogen in the PCV

contributed to this pressure increase. This presumption

is consistent with the fact that a scribble message

remained on the whiteboard in the main control room,

indicating the radiation monitor indication increasedwhen opening the outside door of the air lock at 17:50

[6]. As well, the radiation level in the reactor building

increased at around 22:00 and also in the turbine

building at around 23:00 [9]. These indicate a leakage

from PCV might have taken place already in the early

phase of the accident, although the reason is not

known yet. The D/W pressure at 2:30 on 12 March was

found to be 0.84 MPa; this was almost double of the

Pd, and sufficiently high for causing a leakage as

discussed later.

Figure 6 shows the D/W pressure calculated with

Containment Vessel Balance (CVBAL) [16] along withthe plant records. As briefly described in Appendix A,

CVBAL is a simple computer program developed by

Japan Atomic Energy Agency (JAEA) taking into

account only mass and energy balances within a single

volume. In this calculation, the amounts of hydrogen

and oxidization heat generated by zirconium-water

reaction were given as inputs by assuming that the total

amount of zirconium in the core reacted with steam.

Although structural heat loss was not taken into

account in this calculation, the initial pressure build-

up in the PCV was underestimated. This indicates that

the hydrogen accumulation was not sufficient to lead to

such a large pressure increase, and therefore, addi-tional causes, for example, direct steam release to D/W

through safety valves (SVs) or rupture of SRV line,

may have occurred. A continuous discharge of gaseous

fluid from the containment was assumed to start twice;

one when the D/W pressure reached the peak and the

other when venting was done, in order to obtain good

agreement with plant records. The former simulated

the leakage from the containment and the latter

simulated the venting.

Returning to Figure 5, the RPV pressure was 0.8

MPa at 2:45 on 12 March, while the pressure recorded

before this data point was 6.9 MPa at 20:07 on 11March. This indicates that the depressurization

occurred between these two times. It should be noted

that there is no written indication that the operators

attempted to open the SRVs during this time period [7].

If that was the case, the RPV depressurization was

caused by the damage of the RPV pressure boundary.

Figure 5. RPV pressures, D/W pressure, and injection water flow rate into core at Unit 1.

6 M. Hirano et al.


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In TEPCOs analysis [7], the molten fuels relocated to

the lower plenum and damage to the RPV occurred

15 h after the earthquake, and this caused the RPV

depressurization. The freshwater injection was started

at 5:46 on 12 March using fire engines. This means that

no water had been supplied to the core for 14 h and9 min after the tsunami. Furthermore, if the pump

heads of the fire engines, for which specifications are

not available, were lower than 0.75 MPa, the core

injection would not have started until the containment

was vented at about 14:00. In that case, no water would

have been supplied to the core for about 22 h.

Figure 7 shows the RPV pressure, D/W pressure,

and dose rate measured with the Main Gate Monitor-ing Post. Soon after the D/W pressure reached the peak

at 2:30 on 12 March 2011, the dose rate started to

Figure 7. RPV pressures and D/W pressure for Unit 1 along with site air dose rate.

Figure 6. D/W pressure calculated with CVBAL and plant records for Unit 1.



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increase. This indicates that the containment leakage

might have started due to overpressure at around this

time. The CVBAL analysis in Figure 6 also indicated

that the assumed occurrence of the containment

leakage reproduced the formation of the pressure

plateau at approximately 750 kPa before the actuationof the S/C venting. The hydrogen that had leaked to

the reactor building during this period might have

caused the explosion, namely deflagration or detona-

tion, at 15:36 on 12 March 2011.

Intension is not understandable for some of the

operator actions such as the RPV depressurization and

use of the ICs as discussed in the following.

The operators had started preparation for alter-

native injection within 1 h after the tsunami, and

recognized the necessity of the depressurization of

RPV for the alternate injection [9]. There is, however,

no description indicating they attempted to open the

SRVs in TEPCOs report [9].The ICs could have been used to remove the decay

heat, delay the core damage, and even depressurize the

RPV, if the operators had used them during the

operation after the earthquake and before the tsunami.

According to TEPCO [9], however, they only checked

the condition of one of the ICs at 18:10, almost 3 h

after the tsunami. Furthermore, after they opened the

MO-2A and MO-3A at 18:18 and confirmed the

actuation by checking steam generation in the second-

ary side, they closed the MO-3A at 18:25 and left it

closed for the next 3 h. Then, they opened it again at

21:30. If the core damage started before 21:30 andwater was not sufficiently supplied to the IC secondary

side, opening of the MO-3A at 21:30 might have

caused the containment bypass due to the overheat of

piping.

Regarding the basic design and use of ICs, there

are several things that should be explained further.

As discussed in Section 2, the operators shut them

down according to the procedures to prevent the

RPV from being overcooled at the rate exceeding 55

K/h. Then, the tsunami rendered them inoperable. It

is stated in the Installment License Application

Document of Unit 1 [17] that the design pressure

of an IC is about 7 MPa and its outlet temperature

is 2868

C. If so, it is hard to understand why such ahigh cooling rate exceeding 55 K/h took place. It is

reasonable to presume that the outlet temperatures

of the ICs were much lower than the design value in

the accident. If this is the case, the basic design

philosophy of the ICs, namely what the originally

intended use of the ICs and how they had actually

been used until this accident occurred, should be

explained.

4.2. Unit 2

Figure 8 shows the RPV pressure, D/W pressure and

S/C pressure, and reactor water level at Unit 2. TheRCIC had been continuously operated during approxi-

mately 3 days and the water level had been maintained

at around the normal value, i.e. approximately

4000 mm above the top of active fuel (TAF). Since

the DC power supply was unavailable, it is understood

that the governor valve to control the steam flow to the

RCIC turbine could not have fulfilled its intended

function (the valve seemed to have been in the open

position due to fail as is per design). Its operation

mode was switched from injection to recirculation from

4:20 to 5:00 on 12 March 2011 and, therefore, both S/C

and D/W pressures increased due to accumulation ofdecay heat.

The RCIC stopped at 13:25 on 14 March 2011, it is

not known why, and then, the water level started to

decrease and the RPV pressure started to increase.

The seawater injection started at 19:54 after the RPV

was depressurized at about 18:00 by opening the

SRVs. The water level recovery could be confirmed

after about 22:00 on 14 March 2011. This means that

Figure 8. RPV pressure, D/W pressure, S/C pressure, and reactor water level for Unit 2.

8 M. Hirano et al.


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it took approximately 4.5 h after termination of the

RCIC to depressurize the RPV, and no water had

been supplied to the core for approximately 6 h and

29 min.

Figure 9 shows the comparison of the recorded

PCV pressures with the calculations by CVBAL. Thebase case without any leakage from the PCV results in

significant overestimation of the D/W and S/C

pressures. In the sensitivity calculation, therefore, a

leakage from the gaseous volume of D/W was assumed

to take place from the beginning of the accident. The

calculation agreed better with the plant record when a

leak diameter of 78 mm-equivalent was applied. The

underestimation of pressure during the early phase in

the sensitivity calculation indicates that further better

agreement could be obtained with assuming delayed

occurrence of leakage. It should be noted that a similar

leakage was also assumed to occur in TEPCOs

analysis [7].

Figure 10 shows the same parameters as those in

Figure 8 in a different time range along with the on-siteair dose rate. From this figure, it can be seen that the

RPV depressurization occurred four times at about

18:30, 21:30, 23:30 on 14 March 2011 and 1:30 on 15

March 2011. The depressurization at the first three

times was probably caused by opening the SRV. The

pressure increase after each depressurization was

caused due to the SRV being closed unintentionally.

When it was opened the second time at about 21:30,

the RPV pressure decreased to about 0.5 MPa and the

Figure 9. D/W pressures calculated with CVBAL along with plant records for Unit 2.

Figure 10. RPV pressure, D/W pressure, S/C pressures, and reactor water level for Unit 2 along with on-site air dose rate.



11/18

water level increased immediately after that, which

indicates that sea water injection was increased by the

depressurization. Concurrently with the second depres-

surization, the dose rate started to increase. This

indicates that a leakage from D/W might have taken

place at this time and also that core damage hadalready started by this time. No pressure increase

after the fourth depressurization indicates either the

SRV was stuck open or the RPV pressure boundary

failed.

It should be noted here that large differences

between the D/W and S/C pressures existed after

about 22:00 on 14 March 2011, which increased to

about 0.4 MPa after about 0:00 on 15 March 2011.

This difference cannot be physically explained con-

sidering the large flow path between the S/C and D/W,

which suggests a measurement error. Since the RPV

pressure after the depressurization was almost the same

as the D/W pressure, the D/W pressure measurement isconsidered to be more reliable. The operators, how-

ever, regarded this pressure difference as the actual

pressure distribution, and decided to use the venting

line from the D/W after they had completed the

preparation of that line from the S/C at 21:00 on 14

March 2011 [9]. Then, they completed the preparation

of that line from the D/W at 0:02 on 15 March 2011.

Several minutes after that, however, they confirmed the

closure of a valve on this line. Consequently, the

venting did not appear to be done due to unintentional

closure of the valves on these lines. No further operator

action has been recorded on venting.An explosive sound was heard at around 6:00 on 15

March 2011. Consistent with this, the dose rate

increased significantly and both D/W and S/C were

depressurized before around 11:00. Since a large

amount of highly contaminated water was identified

in the basement of the turbine building and trench

later, a significant failure might have taken place at the

S/C around this time. These suggest that significant

core damage continued and resulted in depressuriza-

tion of the PCV at around 6:00 to 11:00, although the

amount of injected seawater seems to have been

sufficiently large enough to have cooled the core. This

implies that the actual water injection was not sufficient

because the RPV pressure remained high, approxi-

mately 0.8 MPa. This was because venting could not be

done and the D/W pressure remained high until

around 6:00 to 11:00.

Figure 11 shows the RPV pressure calculated with

the TRAC-BF1 code [18] along with the plant record.

The major models and assumptions applied are

summarized in Appendix B. Figure 12 shows the

sensitivity calculations, where the timing of RPV

depressurization and seawater injection is varied after

termination of the RCIC. The calculation showed that

if the RPV depressurization and seawater injection hadbeen done more than approximately 4 h earlier, core

damage could have been avoided.

4.3. Unit 3

Figure 13 shows the RPV pressure, D/W pressure, S/

C pressure, and reactor water level at Unit 3. At Unit

3, the DC supply was available. The RCIC had been

operated for about 20 h since the tsunami and tripped

at 11:36 on 12 March 2011. Then, the operators started

the HPCI at 12:35 and soon, the RPV pressure started

to decrease to about 10 MPa. TEPCO stated that the

minimum flow line was used to distribute part of the

pump discharge flow to the S/C to ensure the

continuous operation of the HPCI pump without

being tripped-off by the high reactor water level (L8)

according to the operating procedures [19]. Although

the conditions needed for such an operation of the

HPCI are not explained in [19], the reason why the

operating procedures required such an operation

should be clarified because continuous operation of

the HPCI accelerates the consumption of water in thecondensate storage tank (CST) and also causes rather

large depressurization of the RPV due to steam flow to

Figure 11. Numerical results by TRAC-BF1 code (RPVpressure and reactor water level for Unit 2).

10 M. Hirano et al.


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the HPCI turbine. The HPCI suddenly stopped at 2:42

on 13 March 2011. Although this might be caused by

depletion of the batteries according to TEPCOs

report [9], the HPCI turbine would not trip since its

steam stop valve could remain open in case of

battery depletion. As well, a shortage of water in the

CST could be a cause, considering the design flow

rate and the capacity of the tank, but switchover of

the water source from CST to S/C would be

available in case the battery was available. Hence,

the reason why the HPCI tripped should be clarified.

Then, the RPV depressurization and the venting were

carried out at 9:08 and 9:20, respectively, andfreshwater injection was commenced at 9:25, which

was changed to seawater injection at 13:12. This

means that it took approximately 6.5 h after the

HPCI trip to depressurize the RPV and no water

injection into the core had been done for approxi-

mately 6 h and 43 min.

Figure 14 shows the same parameters as those in

Figure 13 in a different time range, compared with the

on-site dose rates measured at the main gate as well as

at the Monitoring Post No. 4 (MP 4) located near Unit

6. Before venting was done at 9:20, the on-site dose

rates started to increase, when the D/W pressure was

more than Pd. This indicates that core damage had

already started and also a leakage from D/W had

started to occur by that time. The first peak in D/Wpressure was caused by rapid steam and hydrogen

ingress through the SRVs.

Figure 12. Sensitivity calculations for Unit 2: Maximum clad temperature as a function of recovery action (depressurization andinjection) timing. Water injection starts after RPV pressure falls to 0.6 MPa.

Figure 13. RPV pressures, D/W pressure, S/C pressures, and reactor water level for Unit 3.



13/18

The vent valve (large) closed before 11:17 due to

loss of compressed air and then the D/W pressure

increased rapidly again. Then, it started to decrease at

about 12:30 when the valve was opened by replacing

the air cylinder with a new one. The second peak ofdose rate at around 14:00 on 13 March 2011 might

have been caused by this venting. At about 20:10, the

vent valve (small) was confirmed open, and consistent

with this, the D/W pressure started to decrease from

around that time. The third peak of the on-site dose

rate at about 2:00 on 14 March 2011 might have been

caused by this venting.

After about 0:00 on 14 March 2011, the D/W

pressure started to increase again, which implies the

vent valves again closed unintentionally (due to lack of

compressed air). Then, it reached more than 0.5 MPa

exceeding Pd, and leakage might have taken place due

to overpressure as indicated by the dose rate measured

at MP 4. Hydrogen leaked during this time period

should have largely contributed to the explosion at the

reactor building of Unit 3 at about 11:00. All the

pressures decreased suddenly immediately after the

hydrogen explosion. The reason should be clarified.

5. Major safety issues raised by the F-1NPPs accident

5.1. Basic understanding

In the Governments report [7], a total of 28 lessons

learned have been identified based on the information

available before June 2011. Most of the issues relevantto these lessons learned, however, are not new and

some of them, for instance, the design basis tsunami,

have been discussed since long before the accident,

although they were not sufficiently addressed to the

extent expected from the state-of-the-art technical

knowledge and findings, and international practices.

For some of the issues such as the beyond design basisevents, discussion had been started rather recently

before the accident. Those insufficiently addressed

issues clearly indicate that appropriate framework

needs to be established to pursue continuous improve-

ment toward the highest standards of safety discussed

as the safety objective in the IAEA Safety Funda-

mentals No. SF-1 [20]. Hereafter, the major safety

issues raised by the accident are discussed from the

international perspective.

5.2. Design basis tsunami

In 2002, TEPCO evaluated the design basis tsunami

height based on the method developed by the Japan

Society of Civil Engineers (JSCE) [21], which is also

reflected in IAEA Tsunami Hazard Guide DS417 [22],

and voluntarily revised it from 3.1 m to a range of 5.4

5.7 m [7]. The IAEA mission pointed out that the

increased design basis tsunami height and additional

measures taken in 2002 were insufficient; moreover,

they were not reviewed and approved by the regulatory

authority.

One of the focal issues here may be that the tsunami

recurrence period is not identified in the method of

JSCE [7]. As mentioned in the IAEA mission report,the design basis usually excludes very remote events or

combination of events those typically with a lesser

Figure 14. RPV pressures, D/W pressure, S/C pressures, and reactor water level for Unit 3 along with site air dose rates.

12 M. Hirano et al.


14/18

probability of occurrence than 1 in 10,000 to 100,000

per year. As stated in the Governments report, some

experts pointed out in 2009 the existence of tsunami

that accompanied the Jyogan Earthquake in 869. This

implies that even if that tsunami would have been

taken into consideration, the occurrence probabilitywould be an order of 1 in 1000 per year.

It should be noted here that the NSC revised the

Seismic Design Review Guide in 2006 [23] and it was

newly required to take into account tsunami as one of

the events caused by earthquake. However, neither the

specific requirements nor guides to deal with these

events have been developed yet.

Considering the uncertainties in predicting earth-

quake magnitudes and subsequent tsunami, it is

insufficient to rely just on the breakwater that protects

the facility against the design basis tsunami. Safety

measures based on the defense-in-depth concept should

have been provided in order to reduce the risksassociated with earthquakes and subsequent tsunami

beyond their respective design bases.

5.3. AM measures

The AM measures implemented in the F-1NPPs

were insufficient as discussed in the previous section.

Although their effectiveness had been assessed by

using PSA, it was only for internal events and did not

include external events such as earthquakes and

flooding. For this reason, the resultant core damage

frequency (CDF), for example, for SBO becamesufficiently low by taking the interconnection of

emergency busbars between units and recovery of

off-site and on-site power supplies into account and,

therefore, it was concluded that any additional mobile

EDGs and batteries were not effective to further

reduce the CDF. This was detrimental because if these

external events had been taken into consideration,

such additional measures would have been effective to

reduce the total CDF since the seismic risk is

apparently dominant in Japan. If PSA methods for

external events were judged to be immatured at that

time, a simplified method or even engineering judg-

ments could have been applied to develop the AM

measures for common cause failures (CCFs) caused

by such events.

From the accident, a lack of consideration was

revealed on concurrent SAs in multiple units in a site,

loss of spent fuel cooling, and hydrogen explosion

outside containment. It is also important to emphasize

the fact that the deteriorated on-site environment caused

by the hydrogen explosion hindered the operator actions

to cope with the accident. The deteriorated off-site

environment and infrastructure by the earthquake and

tsunami hindered transportation of heavy items for

support to the site and communication.One of the weaknesses in the current AM measures

might be the use of plant specific systems. The IC is one

of them for Unit 1. Considering the capability of the IC

which operates without being heavily dependent on the

power supply, AM measures by using ICs could have

been prepared.

The Governments report states: In addition,

accident management measures are basically regardedas voluntary efforts by operators, not legal require-

ments, and so the development of these measures

lacked strictness. Moreover, the guideline for accident

management has not been reviewed since its develop-

ment in 1992, and has not been strengthened or

improved. Since the prime responsibility for safety

rests with the licensees, however, they should have

borne in mind the recognition that best efforts should

be made to ensure the safety of their facilities even on a

voluntary basis.

5.4. Regulatory treatment of beyond design basisevents

In 1992, the NSC recommended the licensees

implement the AM measures on a voluntary basis.

The reason is understood as follows [12]: One of the

legal requirements as criteria for the license stipulated

in Article 24 of the Regulation law is no hindrance to

the prevention of the hazard and compliance with this

requirement is strictly reviewed by, for example,

assuming occurrence of design basis events (DBEs).

For this reason, the risk is considered to be suppressed

to a sufficiently low value and the AM measures to

address SAs, namely beyond DBEs, are to reduce therisk further: therefore, no additional regulatory action

is required.

It is stated in the US NRCs task force report

frequently, the concept of DBEs has been equated to

adequate protection, and it has been equated to beyond

adequate protection (i.e. safety enhancements). This

situation seems to be similar to that mentioned above.

However, the US NRC has made regulatory rules that

address beyond DBEs (BDBEs) such as those regard-

ing SBO and anticipated transient without scram

(ATWS). The SBO rule [24], for example, was issued

in 1988.

Contrary to this, in Japan, such rules have never

been made. Regarding SBO, for example, Criterion 27

in the regulatory guide for reviewing the safety design

of light water nuclear power reactor facilities requires

design measures to cope with a short time SBO to

ensure safe shutdown and decay heat removal [25]. In

other words, the extended loss of off-site power did not

need to be considered since it was generally recognized,

at that time, that the reliabilities of both off-site power

and EDGs in Japan were higher than those in the US.

As well, Criterion 27 was intended to assure that the

turbine-driven pump systems, such as RCIC and HPCI

of BWRs and auxiliary feedwater system (AFW) ofPWRs, could cope with the short-term SBO situations.

The SBO rule in the US, on the other hand, requires



15/18

them to cope with a long-term SBO considering

external hazards such as heavy snow fall, tornados,

and hurricane winds.

The IAEA draft Safety Requirements DS414 [26]

for revision of Safety Standards Series No. NS-R-1

requires the design extension conditions (DECs) forbeyond design basis accidents. Similarly, the US NRC

task force concluded that the application of defense-in-

depth should be strengthened by formally establishing,

in the regulations, an appropriate level of defense-in-

depth to address requirements for extended design-

basis events; many of the elements of such regulations

already exist in the form of the SBO rule, ATWS rule,

and others.

There may be several reasons why such rules have

never been established in Japan. One of them seems to

be that the operating experience feedback has been

insufficient. As mentioned above, in the US, the SBO

rule was developed in the 1980s and in someEuropean countries such as Sweden, additional power

generators such as mobile gas turbine generators are

installed on-site in order to cope with the SBO.

However, there has been no discussion on the SBO

since 1994 in Japan.

Another reason might be related to so-called

backfitting. In the US, the backfitting rule [27] provides

bases for determining whether or not backfitting is

required at individual facilities and the NRC requires

the backfitting of a facility only when it determines,

based on the analysis, that there is a substantial

increase in the overall protection of the public healthand safety or the common defense and security to be

derived from the backfit and that the direct and

indirect costs of implementation for that facility are

justified in view of this increased protection.

In European countries, the backfitting is being dealt

with in a 10-yearly safety reassessment referred to as a

periodic safety review (PSR) [28] which includes an

assessment of plant design and operation against

current safety standards and practices, with the

objective of ensuring a high level of safety throughout

the plant operating lifetime.

The Governments report stated: the Government

will clarify technical requirements based on new laws

and regulations or on new findings and knowledge for

facilities that have already been approved and licensed,

in other words, it will clarify the status of retrofitting in

the context of the legal and regulatory framework.

Here, it seems that retrofitting corresponds to

backfitting.

The introduction of AM measures on a voluntary

basis in 1992 may be warranted in the beginning,

considering immaturity of the understanding of SAs

and PSA results showing very low occurrence frequen-

cies of SAs without any AM measure at that time. The

fact that the regulatory position on the AM measureshas never been improved since then, however, clearly

shows that a framework is needed to pursue

continuous improvement toward the highest standards

of safety on both regulatory and industry sides.

5.5. Framework that pursues the highest standards of

safety

In order to establish a framework that pursues the

highest standards of safety that can reasonably be

achieved, it is obvious that the commitments of the

licensees are necessary because the prime responsibility

of safety rests with them.

The regulatory bodies, however, also play important

roles, as well. One of them is to implement clear

regulatory requirements including review guides, stan-

dards, and other relevant documents based on the state-

of-the-art technical knowledge and findings. Such

requirements should, in principle, have legal bases.

Although it may take time to develop such requirements

to resolve a certain issue, the regulatory bodies shouldrecommend or require the licensees to take measures

regardless of a legal or a voluntary basis, if the measures

have an impact on safety. Even on a voluntary basis, as

well, the regulatory bodies may examine the licensees

commitment to implement some of them for enhancing

safety through the review process.

It is of prime importance, furthermore, to continue

to reflect national and international operating experi-

ence, results from safety research worldwide, and

evolution of international safety standards to such

requirements and recommendations.

6. Concluding remarks

Japan needs to establish a framework toward

continuous improvement to pursue the highest stan-

dards of safety that can reasonably be achieved as

described in the IAEA Safety Fundamentals No. SF-1

[20]. Since the prime responsibility for safety rests with

the operating organizations, the licensees need to make

their best effort for continuously improving safety on a

voluntary basis. The major role of the regulatory body

is to have oversight on the licensees efforts and to

revise and backfit the regulatory requirements based on

the state-of-the-art technical knowledge and findings in

accordance with an appropriate rule, if a substantial

increase is clearly expected in the overall protection of

the public health and safety, and the environment.

In order to achieve continuous improvements, all

the entities relevant to nuclear safety such as industries,

regulatory bodies, research organizations, and univer-

sities have to discuss ways to seek the effective frame-

work for this purpose, thoroughly taking the publics

involvement into account.

Furthermore, external stimulation plays a vital role

as a driving force to make the framework actually

functional toward the highest level of safety. Thisexternal stimulation might include pursuing compli-

ance with IAEA Safety Standards and also

14 M. Hirano et al.


16/18

contributing to development of such Standards. Peer

reviews within the frameworks of the Convention on

Nuclear Safety (CNS) [29] and IAEA services such as

the IAEA Integrated Regulatory Review Service

(IRRS) [30] may be effective as an external stimulation.

The contracting parties of the CNS have alreadyproposed to hold a dedicated meeting in 2012 on the F-

1NPPs accident aiming at sharing the lessons learned

and at reviewing the effectiveness and, if necessary, the

continued suitability of the provisions of the CNS [31].

References

[1] NISA, JNES, The 2011 off the Pacific Coast of TohokuPacific Earthquake and the Seismic Damage to the NPPs,NISA and JNES (2011). Available at http://www.nisa.meti.go.jp/english/files/en20110406-1.html.

[2] NISA, NISA News Release, 26 April, NISA (2011).Available at http://www.nisa.meti.go.jp/english/press/

index.html.[3] TEPCO, TEPCO Press Release, May 16, TEPCO

(2011). Available at http://www.tepco.co.jp/en/press/corp-com/release/.

[4] NISA, NISA Press Release, May 17, NISA (2011) [inJapanese]. Available at http://www.nisa.meti.go.jp/earthquake_index.html.

[5] TEPCO, TEPCO Press Release, May 24, TEPCO,Tokyo, Japan, 2011.

[6] NISA, NISA News Releases, May 24 and June 6, NISA,2011.

[7] Nuclear Emergency Response Headquarters Govern-ment of Japan, Report of the Japanese Government to theIAEA ministerial conference on nuclear safety, Govern-ment of Japan, 2011.

[8] IAEA, IAEA International Fact Finding Expert Mission ofthe Fukushima Dai-ichi NPP Accident Following the GreatEast Japan Earthquake and Tsunami, IAEA, Vienna,Austria, 2011.

[9] TEPCO, TEPCO Press Release, June 18, TEPCO,Tokyo, Japan, 2011.

[10] C. Miller, A. Cubbage, D. Dorman, J. Grobe, G.Holahan, and N. Sanfilippo, Recommendations forEnhancing Reactor Safety in the 21st Century, USNRC, Maryland, United States, 2011.

[11] Subcommittee on Safety Design Review Guide [inJapanese], Doc.No. 2-1, August 3, NSC (2011). Avail-able at http://www.nsc.go.jp/senmon/shidai/anzen_sekkei/anzen_sekkei2/siryo2-1.pdf.

[12] NSC, On Accident Management as Measures against

Sever Accidents in Light Water Reactor Facilities [inJapanese], approved by NSC on May 28 (1992).Available at http://www.nsc.go.jp/shinsashishin/pdf/1/ho016.pdf.

[13] NSC, Resolution by NSC [in Japanese], May 28, NSC,1992.

[14] Government of Japan, Convention on Nuclear SafetyNational Report of Japan for the Fifth Review Meeting,Tokyo, Japan, Government of Japan, 2010.

[15] TEPCO, TEPCO Press Release [in Japanese], March 26,TEPCO, Tokyo, Japan, 2004.

[16] Y. Sibamoto, K. Moriyama, and H. Nakamura,Examination of the containment vessel conditions duringthe Fukushima Daiichi NPP accident by simple heat andmass balance model [in Japanese], Transactions of 2011

Fall Meeting of the Atomic Energy Society of Japan,1922 September 2011, Kita-Kyushu, Japan. AESJ:Tokyo, Japan, p. 42, (2011).

[17] TEPCO, The Installment License Application Documentof Unit 1 of Fukushima Dai-ichi Nuclear Power Plant[in Japanese], TEPCO, Tokyo, Japan, 1966.

[18] M.M. Giles, G.A. Jayne, S.Z. Rouhani, et al., TRAC-BF1/MOD1: An Advanced Best-Estimate ComputerProgram for BWR Accident Analysis, NUREG/CR-4356, US NRC, Maryland, United States, 1992.

[19] TEPCO, TEPCO Handouts at Press Conference, July 28,TEPCO (2011). Available at http://www.tepco.co.jp/en/nu/fukushima-np/handouts/index-e.html.

[20] IAEA, Fundamental Safety Principles, Safety Funda-mentals No. SF-1, IAEA, Vienna, Austria, 2006.

[21] Tsunami Evaluation Subcommittee, Nuclear CivilEngineering Committee, Tsunami Assessment Method

for Nuclear Power Plants in Japan, JSCE, Tokyo, Japan,2002.

[22] IAEA, Meteorological and Hydrological Hazards in SiteEvaluation for Nuclear Installations, Draft Safety GuideDS417, IAEA, Vienna, Austria, 2009.

[23] NSC, Regulatory Guide for Reviewing Seismic Design ofNuclear Power Reactor Facilities, NSC, Tokyo, Japan,2006.

[24] US NRC Regulations, Title 10 Code of FederalRegulations, Section 50.63 Loss of All AlternatingCurrent Power, 53 FR 23215, US NRC, Maryland,United States, 1988.

[25] NSC, Regulatory Guide for Reviewing Safety Design ofLight Water Nuclear Power Reactor Facilities, NSCRG:L-DS-1.0, NSC, Tokyo, Japan, 1990.

[26] IAEA, Safety of Nuclear Power Plants: Design, DraftSafety Requirements DS414, Revision of Safety Stan-dards Series No. NS-R-1, IAEA, Vienna, Austria, 2010.

[27] US NRC Regulations, Title 10, Code of FederalRegulations, Section 50.109 Backfitting, 53 FR 20610,US NRC, Maryland, United States, 1988.

[28] IAEA, Periodic Safety Review of Nuclear Power Plants,

Safety Guide No. NS-G-2.10, IAEA, Vienna, Austria,2003.[29] IAEA, Convention on Nuclear Safety, INFCIRC/449,

IAEA, Vienna, Austria, 1994.[30] IAEA, Integrated Regulatory Review Service, IAEA.

Available at http://www-ns.iaea.org/reviews/rs-reviews.asp.

[31] L. Ganjie, SummaryReport of the 5thReviewMeeting of theContracting Parties to the Convention on the Safety onNuclear Safety, CNS/RM/2011/6/Final, IAEA, Vienna,Austria, 2011.

[32] K. Moriyama, Y. Maruyama, and H. Nakamura, SteamExplosion Simulation Code JASMINE v.3 Users Guide,JAEA-Data/Code 2008-014, Japan Atomic EnergyAgency, Ibaraki, Japan, 2008.

[33] T. Watanabe, M. Ishigaki, A. Sato, and H. Nakamura,Analysis of BWR station blackout accident[in Japanese], J. At. Energy Soc. Jpn. 10(4) (2011), pp.240244.

Appendix A

A model, CVBAL, illustrated in Figure A1 was developedfor estimation of accident conditions inside the PCV fromobserved pressure and temperature trends based on roughassumptions. It often happens that an application of moresophisticated codes needs much more time and labor for thepreparation of input data and running the calculation. Also,they are not necessarily suitable for handling low pressuresubcooled conditions. Thus, such a simple program with a

simple method may be highly useful.A set of conservation equations, exit conditions, and

numerical solution method are summarized below. The

http://www.nisa.meti.go.jp/english/files/en20110406-1http://www.nisa.meti.go.jp/english/files/en20110406-1http://www.nisa.meti.go.jp/english/press/indexhttp://www.nisa.meti.go.jp/english/press/indexhttp://www.tepco.co.jp/en/press/corp-com/release/http://www.tepco.co.jp/en/press/corp-com/release/http://www.nisa.meti.go.jp/earthquake_indexhttp://www.nisa.meti.go.jp/earthquake_indexhttp://www.tepco.co.jp/en/nu/fukushima-np/handouts/index-ehttp://www.tepco.co.jp/en/nu/fukushima-np/handouts/index-ehttp://www-ns.iaea.org/reviews/rs-reviews.asphttp://www-ns.iaea.org/reviews/rs-reviews.asphttp://www-ns.iaea.org/reviews/rs-reviews.asphttp://www-ns.iaea.org/reviews/rs-reviews.asphttp://www.tepco.co.jp/en/nu/fukushima-np/handouts/index-ehttp://www.tepco.co.jp/en/nu/fukushima-np/handouts/index-ehttp://www.nisa.meti.go.jp/earthquake_indexhttp://www.nisa.meti.go.jp/earthquake_indexhttp://www.tepco.co.jp/en/press/corp-com/release/http://www.tepco.co.jp/en/press/corp-com/release/http://www.nisa.meti.go.jp/english/press/indexhttp://www.nisa.meti.go.jp/english/press/indexhttp://www.nisa.meti.go.jp/english/files/en20110406-1http://www.nisa.meti.go.jp/english/files/en20110406-1


17/18

water-vapor state is calculated by a fast running wide rangesteam table program developed for JASMINE code [32].The ideal gas equation was used for non-condensable gases(e.g. N2, H2) with the zero point of the energy placed at273 K.

Mass conservation equations

Mass conservation for component k (k v,l,n,ni,lc) isexpressed by the following equations,

dmv

dt _miv _mov _mev; A1

dml

dt _m0il _m

0ol _m

0ev _m

0lex; A2

dmlc

dt _milc _molc _mlex; A3

dmn

dt _min _mon; A4

dmni

dt _mini _moni; A5

mv rvVg;ml rlVl; mlc plcVlc; mn rnVg; mni rniVg;A6

Vt Vg Vl Vlc const:; A7

where,

mk: mass of component k,_mik, _mok: mass flow rates of flow-in (ik) and flow-out (ok)

of component k,_mev: evaporation rate of water,_mlex: mixing rate of the subcooled water into the

saturated water,Vt, Vg, Vl, Vlc: total, gas and water volumes,rk: density of component k.

Subscripts v, l, lc, n, and ni for component k indicate,

v: vapor,l: water (saturated),lc: water (subcooled),n: non-condensable gas injected from outside (e.g. N2),ni: non-condensable gas produced inside the volume (e.g.

H2).

Energy conservation equations

Energy conservation equations are expressed as,

E mvev mlel mlcelc mnen mnieni; A8

dE

dt Wgi Wgo pt

_mil

ril

_milc

rilc

pt

_mol

rl

_molc

rlc

_miveiv _mileil _minein _minieini _milceilc

_movev _molel _monen _monieni _molcelc

X

k

QK

A9

where,E: total internal energy in the system,ek: internal energy of component k,rik: densities of flow-in fluids (at the total pressure and

the temperature of each flow-in fluid),pt: total pressure of the gas phase (pvpnpni).

For the terms in the right-hand side of Equation (A9),the first line means the work at the inlet and outlet, thesecond and third lines mean flow-in and flow-out of theinternal energy, respectively, and the last one means the heatinput.

Exit conditions

The flow rate is expressed by the following, with the holearea (A), density, and the velocity (u),

_mok Aru: A10

The velocities of gas components are assumed common(uniform mixture).

The Bernoullis law for incompressible fluids is used forthe liquid phase,

u ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi

2p p0=rkp

; A11

where, p0: outside pressure.The generalized Bernoullis law with compressibility

is used for the gas phase with consideration on the criticalflow,

u

ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi2gg1

ptrt

p2=gr p

g1=gr pr > pc

qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi

ptrtg

2g1

g1g1

rpr pc

8>: A12

pr p0=pt;pc 2

g 1

gg1

; A13

where,rt: total density of the gas phase (rv rn rni),g: specific heat ratio of the gas mixture,

pc: critical pressure.

Figure A1. Mass and energy balance model for theevaluation of pressure and temperature based onassumptions of PCV situation during the accident.

16 M. Hirano et al.


18/18

Numerical solution method

(1) Integrate the mass conservation equations fromEquations (A1)(A5) and energy conservation equa-tion of Equation (A9) with an explicit integrationscheme in terms of time, obtain the mass of eachcomponent and the total internal energy at the new

time step, except that the phase change term,_mevDt Dmev, in Equations (A1) and (A2) is left

unsolved with implicit scheme as follows,

mvn1 Dmn1ev m

v m

nv Dt _miv _mov;

A14

mn1l Dm

n1ev m

l m

nl Dt _mil _mol _mlex:

A15

Superscripts (n), (n+1) denote old and new timestep variables, * denotes the intermediate values of

mass changes except the phase change.(2) Total internal energy at the new time step is expressedas the sum of specific internal energies of componentsthat are unknown.

En1 mv Dmn1ev

en1v m

l Dm

n1ev e

n1l

mn1lc e

n1lc m

n1n e

n1n m

n1ni e

n1ni :

A16

The known masses except Dmev are substituted inthis equation.

(3) Solve the conservation and state equations at the newtime step by the Newton method and obtain thephase change mass, pressures, and volumes.

Appendix B

The analysis of a long-term station blackout accident of aBWR has been performed using the TRAC-BF1 code, andthe results were compared with the observed data at the F-1NPP Unit 2 [33]. As shown in the noding diagram in FigureB1, the RPV thermal hydraulics from the feed water line tothe steam line were modeled. Although Unit 2 was a BWR-4

with 780 MW output, the analysis was based on a BWR-5with 1100 MW by scaling the feed water flow rate with thepower ratio. Since the power-to-volume-ratio and the relativevolume distribution in the RPV are almost the same betweenthe two reactors, thermal hydraulic responses are expected tobe conserved. The reactor scram due to the earthquakeacceleration and the station blackout sequence were assumedto occur. The RCIC was actuated under the assumption thatthe steam flow rate to the RCIC turbine and the injectionflow rate from the RCIC pump were both balanced with thereactor power. The steam line and the feed water line shownin Figure B1 were used for the RCIC. The RCIC wasterminated at 250,000 s, and depressurization using the SRVwas performed at 270,000 s, according to the event sequenceat Unit 2. Although the reactor type and the power weredifferent, the reactor pressure and the core liquid level were ingood agreement with the observed data [33]. This mayindicate the appropriateness of the above-mentioned scalingconsideration.

Figure B1. TRAC-BF1 noding diagram.


insights from review and analysis of the fukushima daiichi accident

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