Nuclear Safety
2021

Benchmark Study of theAccident at the Fukushima
Daiichi Nuclear Power Plant
Summary Report

NE A

Nuclear Safety

Benchmark Study of the Accident
at the Fukushima Daiichi Nuclear Power Plant
Summary Report© OECD 2021
NEA No. 7525
NUCLEAR ENERGY AGENCY
ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 38 democracies work together to address the economic,
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NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consists of
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take part in the work of the Agency.
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Cover photos: Fukushima Daiichi Nuclear Power Plant (Creative Commons; Dean Calma/IAEA).

ACKNOWLEDGEMENTS

Acknowledgements

The following representatives from the participating organisations are acknowledged for theircontributions to the project and to the authoring of this report.

Operating agent:
Institute of Applied Energy (IAE-Japan)

Participating organisations and authors:
IAE (Japan): M. Pellegrini
CSN/CIEMAT (Spain): L.E. Herranz, C. López
GRS (Germany): M. Sonnenkalb, M. Sogalla
PSI (Switzerland): T. Lind, L. Fernandez Moguel
JAEA (Japan): Y. Maruyama
NRA(S/NRA/R) (Japan): H. Hoshi
CRIEPI (Japan): Y. Nishi
IRSN (France): D. Jacquemain
IBRAE (Russian Federation): K. Dolganov
KAERI (Korea): J.H. Song, S.I. Kim, T.W. Kim
CEA (France): C. Journeau, G. Ducros, Y. Pontillion, L. Godin-Jacqmin
NRC/DOE/EPRI (United States): N. Andrews (SNL), R. Lee (NRC), D. Peko (DOE), R. Wachowiak (EPRI)
CNL (Canada): A. Morreale
VTT (Finland): T. Sevon

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TABLE OF CONTENTS

Table of contents

List of abbreviations and acronyms .................................................................................................... 9
Executive summary .............................................................................................................................. 11
Chapter 1. Introduction ...................................................................................................................... 15
Chapter 2. Results of the unit 1 analyses ....................................................................................... 17
Overview of the unit 1 accident ..................................................................................... 17
Thermal hydraulics and core degradation.................................................................... 17
Fission product release and behaviour ......................................................................... 20
Chapter 3. Results of the unit 2 analyses ....................................................................................... 25
Overview of the unit 2 accident ..................................................................................... 25
Thermal hydraulics and core degradation.................................................................... 25
Fission product release and behaviour ......................................................................... 29
Chapter 4. Results of the unit 3 analyses ....................................................................................... 35
Overview of the unit 3 accident ..................................................................................... 35
Unit 3 thermal hydraulics and core degradation ......................................................... 35
Fission product release and behaviour ......................................................................... 38
Chapter 5. Atmospheric transport and dispersion: Results for Fukushima Daiichi
units 1, 2 and 3 ................................................................................................................ 43
Unit 1 .................................................................................................................................. 43
Unit 2 .................................................................................................................................. 43
Unit 3 .................................................................................................................................. 43
Comparisons for three units combined ........................................................................ 45
Estimated integral releases ............................................................................................. 47
Chapter 6. Radiological measurements and backward calculations of the source term ....... 49
Dose rate measurements at the plant site and around .............................................. 49
Dose rate measurements/samplings in the Pacific Ocean ......................................... 50
Backwards calculation of source term by WSPEEDI .................................................... 50
Backwards calculation of source term by the GRS method ........................................ 51
Comparison of the GRS and WSPEEDI source term calculation results .................... 53
Chapter 7. Lessons learnt and conclusions from the BSAF ......................................................... 57
Review of major technical findings ............................................................................... 57
Review of major impacts to the BSAF stakeholders .................................................... 60
Impact on industry practice in severe accidents ......................................................... 60
Overall conclusions .......................................................................................................... 61
References ............................................................................................................................................ 63

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TABLE OF CONTENTS

List of figures

6

2.1.

Comparison of analyses results versus measurements for unit 1: Reactor pressure
vessel pressure and dry well pressure...................................................................................... 17

2.2.

Unit 1 in-vessel hydrogen generation....................................................................................... 19

2.3.

Unit 1: Primary containment vessel evolution over 0-500 hour period and alternative
water injection mass flow rate .................................................................................................. 20

2.4.

Unit 1 concrete erosion depth (origin at the bottom of the sump) ....................................... 20

2.5.

Unit 1 mass fraction of fission products released from fuel ................................................. 21

2.6.

Unit 1 mass fraction of fission product deposited on structures in the dry welland solved in water in the suppression chamber ................................................................... 22

2.7.

Comparison of calculation results against measured data: Unit 1 dose rate in the
primary containment vessel ...................................................................................................... 23

2.8.

Unit 1: Comparison of predicted cesium and iodine release [Bq] versus backwards
analyses (GRS and WSPEEDI) ..................................................................................................... 23

3.1.

Comparison of analyses results versus measurements for unit 2 ........................................ 26

3.2.

Three reactor pressure vessel pressure peaks in unit 2 ......................................................... 27

3.3.

Unit 2 primary containment vessel dry well pressure during a longer term ...................... 28

3.4.

Unit 2 mass fraction of fission product released from fuel a) Xe and b) I ........................... 29

3.5.

Unit 2 mass fraction of fission products deposited on structures in the dry welland solved in water in the suppression chamber ................................................................... 30

3.6.

Comparison of analyses results against measured data in the unit 2 dose rate in
the primary containment vessel ............................................................................................... 31

3.7.

Unit 2 mass fraction of fission products released into the environment ............................ 32

3.8.

Unit 2 comparison of predicted cesium and iodine releases versus backwards analyses(GRS and WSPEEDI) and RPV pressure (RPV 3 pressure peaks period until 120 hours) ......... 32

4.1.

Unit 3 reactor pressure vessel pressure and primary containment vessel pressure ......... 36

4.2.

Unit 3 in-vessel hydrogen generation....................................................................................... 36

4.3.

Unit 3: A. containment pressure; B. concrete erosion depth; C. ex-vessel hydrogengeneration..................................................................................................................................... 37

4.4.

Unit 3 mass fraction of fission product released from fuel ................................................... 39

4.5.

Unit 3 mass fraction of fission products deposited on structures in the dry welland solved in water in the suppression chamber ................................................................... 40

4.6.

Comparison of calculation results against measured data for the unit 3 dose ratein the primary containment vessel ........................................................................................... 41

4.7.

Unit 3 mass fraction of fission products released to the atmosphere ................................. 41

4.8.

Unit 3 comparison of predicted cesium and iodine releases [Bq] versus backwards
analyses (GRS and WSPEEDI) ..................................................................................................... 42

5.1.

Unit 1 source terms for Cs-137 predicted by five BSAF participants .................................... 44

5.2.

Unit 2 source terms for Cs-137 predicted by six BSAF participants...................................... 44

5.3.

Unit 3 source terms for Cs-137 predicted by six BSAF participants...................................... 44

5.4.

Combined three-unit source terms for Cs-137 predicted by five BSAF participants .......... 45

5.5.

Cs-137 isopleths for the three-unit source terms, the GRS/backwards calculation
source term and observed ground deposition pattern for Cs-137 ........................................ 46

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TABLE OF CONTENTS

6.1.

Monitoring posts within and around the Fukushima site ..................................................... 49

6.2.

Dose rate measurements performed by TEPCO at some selected monitoring
points and start of severe core degradation phases in units 1-3 .......................................... 50

6.3.

Sampling points in the Pacific Ocean ....................................................................................... 50

6.4.

Release rates from reactor scram on 11 March to 22 March 2011 computed by WSPEEDI ... 51

6.5.

Accumulated release of Cs-137 reconstructed by the GRS for the first two weeksof the accident ............................................................................................................................. 53

6.6.

Comparison of Cs-137 release rates obtained by the GRS (data aggregated tohourly values) and WSPEEDI source term reconstruction methods..................................... 54

6.7.

Comparison of accumulated releases of Cs-137 and I-131 by the GRS andWSPEEDI source term reconstruction methods ...................................................................... 55

List of tables
1.1.

Participants, employed codes and analytical work in Phase II of the BSAF project ........... 16

2.1.

Unit 1 event time of reactor cooling system leak or failure................................................... 18

2.2.

Unit 1 event time of reactor pressure vessel failure and failure mode ................................ 18

3.1.

Unit 2 primary containment vessel and reactor pressure vessel (RPV) lower head
failure and debris mass released from the RPV....................................................................... 28

4.1.

Unit 3 assumed lower head failure time and mode of failure; total debris mass
released from the reactor pressure vessel to the containment ............................................ 37

4.2.

Unit 3 assumed leakages from the reactor pressure vessel into the primarycontainment vessel and timing ................................................................................................. 38

5.1.

Integral releases of Cs-137 (PBq) predicted by BSAF participants ......................................... 47

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LIST OF ABBREVIATIONS AND ACRONYMS

List of abbreviations and acronyms

ADS

Automatic depressurisation system

ATD

Atmospheric transport and dispersion

AWI

Alternative water injection

BSAF

Benchmark Study of the Accident at the Fukushima Daiichi Nuclear
Power Station

BWR

Boiling water reactor

CAMS

Containment air measurement system

CEA

Alternative Energies and Atomic Energy Commission (France)

CNL

Canadian Nuclear Laboratories

CRIEPI

Central Research Institute of Electric Power Industry (Japan)

Cs

CesiumCIEMAT

Research Centre for Energy, Environment and Technology (Spain)

DOE

Department of Energy (United States)

D/W

Dry wellFP

Fission product

GRS

Gesellschaft für Anlagen- und Reaktorsicherheit (Global Research for Safety)

HPCI

High pressure coolant injection

I

IodineIAE

Institute of Applied Energy (Japan)

IBRAE

Nuclear Safety Institute of the Russian Academy of Sciences

IRSN

Radioprotection and Nuclear Safety Institute (France)

JAEA

Japan Atomic Energy Agency

KAERI

Korea Atomic Energy Research Institute

MCCI

Molten core concrete interaction

MP

Monitoring post

MSL

Main steam line

NEA

Nuclear Energy Agency

NG

Noble gas

NRA(S/NRA/R)

Regulatory Standard and Development Department
Secretariat of the Nuclear Regulatory Authority (Japan)

NRC

Nuclear Regulatory Commission (United States)

PCV

Primary containment vessel

PSI

Paul Scherrer Institute (Switzerland)

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LIST OF ABBREVIATIONS AND ACRONYMS

10

PWR

Pressurised water reactor

RCIC

Reactor core isolation cooling

RCS

Reactor cooling system

RPV

Reactor pressure vesselSA

Severe accident

SAMGs

Severe accident management and guidelines

S/C

Suppression chamber (torus)

SRV

Safety relief valve

ST

Source term

SNL

Sandia National Laboratories (United States)

TEPCO

Tokyo Electric Power Company (Japan)

VTT

Technical Research Centre of Finland

WSPEEDI

Worldwide version of System for Prediction of Environmental
Emergency Dose Information

Xe

Xeron

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EXECUTIVE SUMMARY

Executive summary

The Great East Japan earthquake occurred on 11 March 2011 at 2:46 p.m. JST (Japan Standard
Time). At the onset of the earthquake, the operating units at the Tokyo Electric Power
Company’s (TEPCO) Fukushima Daiichi Nuclear Power Plant were automatically shut down.
However, the subsequent tsunami and flooding of critical site areas and equipment produced
conditions that were significantly beyond the design basis of the Fukushima Daiichi Nuclear
Power Plant, resulting in three core meltdown accidents to the operating reactors.
In the aftermath of the event, the Japanese Ministry of Economy, Trade and Industry’s
Agency of Natural Resource and Energy proposed an international programme to the OECD
Nuclear Energy Agency (NEA) to apply severe accident analysis codes to the simulation of the
Fukushima Daiichi events. This initiative became the Benchmark Study of the Accident at the
Fukushima Daiichi Nuclear Power Station (BSAF). Phase I of the project was initiated in 2012 andfocused on the severe accident progression for the first six days. Phase II was established in
April 2015 and extended the analysis to 21 days (500 hours) after the accident, including ananalysis of fission product (FP) behaviour. Two generic objectives were defined:•

to provide information and analysis on the severe accident (SA) progression, FP behaviourand source term (ST) estimation, and to support safe and timely decommissioning of the
Fukushima Daiichi Nuclear Power Plant;•

to contribute to the improvement of methods and models of the SA codes in order toreduce uncertainties in SA analysis.

Fourteen partners from eleven countries joined the project. A number of well-known SA
codes have been applied, such as ASTEC, MAAP, MELCOR, SAMPSON, SOCRAT and THALES.
Phase II of the project gave the participating organisations a chance to revise or consolidate
their interpretation of the accident progression in units 1-3 from Phase I of the project.
The accident progression and associated code predictions for each of the three units are
summarised below. Some conclusions and the main outcomes from the BSAF project are also
included below. The accident transients are characterised by a paucity of data on the actual
progression and a range of operator actions, with varying degrees of success in the face of theenormous challenges posed by the lack of power and reliable information.
During the unit 1 accident: For the first ten hours, the operators were largely blind as to thestate of the reactor, but in time initiated measures to inject water into the reactor pressure
vessel (RPV) to try to avoid or mitigate any fuel damage that might initiate from loss of coolant
after so many hours with no water injection. They were, for the most part, not successful intheir interventions. Attempts to inject emergency cooling water into the core were notsuccessful, with considerable ambiguity about pipe routing within the primary containment
vessel (PCV). Operators were not able to control the RPV pressure, and for the first 24 hours, theyalso could not control the PCV pressure. As a result, core damage is believed to have commenced
early in the accident, 5-12 hours from shutdown. Unit 1 was the first reactor to have suffered
severe damage. In this first 5-12-hour period, a yet to be identified failure or leak of the pressure
boundary depressurised the RPV. As it was not possible to remove the decay heat or injectcooling water, the PCV pressure increased to the point that steam, hydrogen and fission
products leaked to the environment, believed to have occurred in the PCV head flange region.
In the BSAF project analyses, these key events were modelled using various assumptions ofeither leakage at the reactor cooling system (RCS) pressure boundary or a larger failure such asa main steam line (MSL) creep rupture or lower head attack by debris. All calculations predictpressure signatures generally consistent with the sparse RCS and PCV data points in this period.

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EXECUTIVE SUMMARY

All analyses predicted almost complete core melt and relocation, debris attack of the lower
head (between about 7 and 15 hours), and near complete debris release to the pedestal region.
Molten corium concrete interaction (MCCI) is predicted in all of the calculations, contributing to
the high containment pressure until the suppression chamber (S/C) was manually vented ataround 24 hours. After the vent, the PCV pressure was predicted generally to increase due to
MCCI gas generation until a leakage in the containment is assumed in all calculations to matchthe observed drop in pressure. At around 270 hours, most of the calculations assumed anincrease in the rate of alternative water injection (taken to be 0 prior to this time) and predicted
an increase of the PCV pressure consistent with recorded data. This suggests that the ex-vesselcore debris was still hot and the MCCI might have been ongoing. In fact, most calculations
indicate that interaction of the molten material with concrete was still ongoing at 500 hours. It
is important to note that there is no evidence as of yet from the unit 1 plant measurements, or
subsequent in-containment investigations, to confirm this code prediction.
The release of fission products (noble gases [NG], cesium [Cs] and iodine [I]) is strongly
linked to the onset of core degradation. Most calculations show a very rapid release of these FPs,
with the total release from damaged fuel to the RPV and PCV reaching a limit generally above
80% of the initial inventory. Note that this predicted behaviour is common to all the units. The
simulated containment FP distribution was influenced by the nature of the assumed RPV leakor failure, which affects the amount of FPs discharged into and scrubbed by the suppression
pool. For most calculations, large amounts of Cs and I were scrubbed by the suppression pool
during the early injection through the safety relief valves (SRVs), and a large amount of Cs wascomputed to be deposited on the upper structures of the RPV. Almost all calculations estimateda first FP release to the environment at around ten hours, when the PCV head flange is predicted
to have commenced leaking. Each code’s ST prediction was compared against estimates froman inverse analysis of atmospheric releases derived using atmospheric transport and dispersion(ATD) codes/methods, such as the WSPEEDI code (Worldwide version of System for Predictionof Environmental Emergency Dose Information) and a GRS method. This comparison shows thatthe integral release after the PCV venting at around 24 hours was fairly well predicted across allcode calculations, with a variability of about 1 order of magnitude relative to the estimation.
The unit 2 accident was characterised by the operation of the reactor core isolation cooling(RCIC) system for around 70 hours in an unanticipated “self-regulating” mode. The mechanism
is not yet fully understood, but it is believed that the uncontrolled full flow RCIC injection
overfilled the RPV up to the steam lines, driving two-phase steam and water into the RCIC
turbine. The rise in the PCV pressure during this period of RCIC operation is lower than expected,
and believed to be the result of stratification of the suppression pool and additional heatremoval due to partial flooding of the torus room by seawater during the tsunami. After around70 hours, the RCIC system failed, prompting operators to depressurise the RPV and attempt lowpressure injection using fire trucks. This action was unsuccessful, and the water level in the
RPV decreased, leading to core damage.
All participants reasonably captured the main trends observed in the plant measurements(i.e. the high RPV pressure and water level and the gradual pressurisation of the PCV) during the
RCIC operation period. The simulated core heat up and degradation phase started at about
75 hours, shortly after operators depressurised the RPV. This phase was characterised by threelarge pressure transients in the RPV and the PCV. These transients are calculated to arise fromhydrogen released by oxidation of core materials that experienced renewed access to steam whileslumping into the lower plenum. Generally, participants that predicted a greater mass of hydrogenproduction (in excess of 800 kg) were able to predict a PCV pressure transient that was in closer
agreement with the measurements. Most calculations predict lower-head failure between 83 and129 hours and consequent release of 16-161 tonnes of debris into the pedestal. Here, the analyses
predicted ongoing transfer of mass from the failed vessel to the reactor cavity and had difficulty
predicting the significant materials remaining in the lower head that is suggested by the TEPCO
muon tomography examinations. Most of the calculations did not predict the onset of MCCI at
unit 2. For those that did, the degree of MCCI as well as the extent of predicted radial and axial
erosion of the reactor pedestal depended strongly on other modelling assumptions.
In unit 2, no leaks were predicted or assumed to connect the RPV to the dry well (D/W) acrossall code calculations prior to RPV failure. Thus, volatile FPs released through core degradation weretransferred to the S/C through the operating SRV and further to the D/W through the operating

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EXECUTIVE SUMMARY

vacuum breakers. Evaluation of the code calculations against the measured S/C dose rate datafrom the continuous air measurement systems (CAMS) showed reasonable agreement with thetiming of increases or decreases in the dose rate. While the majority of code calculations tended
to over-predict the measured dose rate by a factor of 3-4, the agreement is remarkable nonetheless.
A few participants were able to predict the measured D/W dose rate increase at the postulated
time of lower head failure around 95 hours. Estimation by the ST inverse analyses showed a firstpeak release of FPs to the environment at about the time of the first RPV pressure peak, followedby additional peak releases some hours later. A few participants predicted such a first peak release,though the magnitude of the predicted release was considerably lower than the release estimates
based on environmental measurements because the PCV pressure was not large enough to justify
a PCV leak in the calculations. Almost all calculations predicted a large release of FPs to theenvironment later, at the time of the PCV failure (about 89 hours), which entered the environment
through the open reactor building blowout panel by various presumed leakage paths from the PCV.
The variation in the code calculations of the total FP release to the environment was within one
order of magnitude of the estimates by the backwards calculations.
The unit 3 accident was characterised by the normal functioning of heat removal from the
RPV in the first 35 hours of the accident, first by the RCIC system, then by the high pressure
coolant injection (HPCI) system. While these systems performed normally, their operation
replaced water lost from the RPV and moved heat from the reactor fuel to the suppression pooland the PCV atmosphere. The increase in PCV pressure due to steady injection of steam wascontrolled in a later phase of the accident by activation of sprays in the S/C.
During this first accident period, all of the participants predicted the main signatures in the
RPV and PCV reasonably well. After the HPCI was stopped, unit 3 experienced a period of about
ten hours with no coolant injection into the reactor, which is thought to have caused the waterlevel to drop below the bottom of the fuel. Most of the analyses show that core degradation starts
during this time, and was accompanied by hydrogen generation, causing further pressurisation of
the PCV. The reactor is postulated to have depressurised by spurious operation of the automatic
depressurisation system (ADS), from about 42 hours, until the hydrogen explosion in the unit 3
reactor building, at 68 hours. This period was characterised by a number of attempts by operators
to depressurise the containment and inject water into the RPV using fire trucks. As shown by mostof the analyses, core degradation was characterised by a number of core slumping events, leading
to a number of periods of D/W pressure increase mitigated by either containment venting or
leakage out of an impaired D/W upper head. Progressive core damage was predicted by all
calculations to ultimately result in a failure of the RPV lower head between about 43 hours and73 hours. Due to the large debris discharge in the containment, the MCCI was predicted by mostcodes, and persisted in the analyses until 500 hours (the end of the simulation). Even though large
amounts of debris in the reactor pedestal were observed by direct inspection (heights up to
2.5 metres), the occurrence of the MCCI for such a long period of time has not been confirmed bypost-accident investigations.
Comparison of FP data against D/W CAMS dose rate measurements determined that
calculations which assume a moderate leakage from the RPV to the D/W before RPV failure predicta dose rate increase in closer agreement to the onsite and offsite ground and air measurements.
Across all the calculations, large fractions of Cs and I were predicted to be retained in the
suppression pool water and large fractions of Cs were predicted to deposit on the RPV wall. The
main FP release to the atmosphere from unit 3 was calculated to take place from 45 hours to
70 hours during the first two containment vents and at the time of the hydrogen explosion.
Comparisons with the ST inverse analyses show that most calculations predicted releases to theatmosphere of the same order of magnitude as the backwards estimates, with releases of lessthan 0.5% of Cs and less than 2% of I.
In conclusion, from the point of view of the accident progression, the BSAF project has helped
to clarify a large number of uncertainties that were initially identified regarding the operation ofsafety systems, alternative water injection, the period of core degradation and relocation, and themain distribution of debris and FPs within the units. Despite the many clarifications, there remain
several areas specific to the Fukushima accident transients that are worthy of further
investigation, including refinement of core degradation periods, hydrogen generation, MCCI
progression, correlation of CAMS response and details of the source terms, especially the FP
transport by water leakages from the PCV. In addition, key issues identified through this study are
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EXECUTIVE SUMMARY

the debris transported through the core plate, lower head failure mode and debris discharge into
containment, debris retention by structures below the RPV such as control rod drives andinstrumentation tubes, long-term MCCI progression and containment failure modes. By continued
post-accident investigations during the decommissioning of the Fukushima Daiichi reactors,
valuable information could be provided to close some of these issues. In fact, recent plantinspections (i.e. muon tomography and PCV internal assessment) have already provided
observations that present challenges for the current code modelling. In addition, the analyses of
the BSAF have highlighted many uncertainties in the accident boundary conditions, such as theamount and timing of alternative water injection, systems’ leak paths and component offnominal behaviour might affect the results of the analyses regardless of how accurate the in-code
models are. Fission product modelling has also brought up issues that need further improvementin the present codes: pool scrubbing under high temperature and high gas velocities; iodine
chemistry, FP transport by water leakages from the RPV and the PCV, remobilisation of FPs in thelong run and under the effect of some accident management procedures, etc.
Looking beyond these specific conclusions, the BSAF activities have had a significant impact
on the diverse and numerous stakeholders in the nuclear power community. These stakeholders
include the large group of researchers that has directly participated in the BSAF programme, as
well as governmental and regulatory organisations around the world, TEPCO and industry
partners, and the general public.
Perhaps one of the most fundamental contributions of the BSAF project has been to improvethe understanding of the accident sequences. Through the diversity of the modelling codes andapproaches, and the use of parametric studies, it has been possible to identify the more likely
scenarios that can fit with the limited data. As more information becomes available throughdecommissioning, these scenarios can be further down-selected and refined.
This significant understanding of the progression of the accidents among the technicalparticipants has both clarified to analysts the essential details of the accident progression andprovided deep insights into future code and model improvements to advance predictive
capabilities (as described above). The BSAF analyses have served to address many uncertainties
that underlie current regulatory practices with respect to source term characteristics and major
assumptions such as the timing, magnitude and duration of severe accident radiological releases,
as well as pointing to improvements in regulatory oversight and protective measures inemergency response to severe accidents. The industry has used the BSAF analytical findings to
improve accident management procedures, water management and containment protection.
Additionally, a major outcome of the BSAF programme and the resulting improved severe accident
code performance has been to provide information useful to the decades-long decommissioning
activities and to help inform decommissioning technologies with respect to likely distributions ofcore materials and their likely morphology and material composition for optimal design of tools
and extraction approaches. Finally, the public benefits from the understanding of nuclear
accidents and their consequences and the demonstrated competence in explaining what
happened at Fukushima. With the existential threats posed by climate change and global warming,
it is ever more critical that the nuclear industry demonstrates understanding and competency inthe use of nuclear technology as a source of low-carbon energy. Programmes such as the BSAF
can help dispel unwarranted fear and increase understanding of nuclear technology.

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INTRODUCTION

Chapter 1. Introduction

The Tokyo Electric Power Company (TEPCO) shut down its units at the Fukushima Daiichi Nuclear
Power Plant when the Great East Japan earthquake occurred on 11 March 2011 at 2:46 p.m. JST(Japan Standard Time). At first, safety systems were operated successfully to remove decay heat,
driven by diesel generators. However, the subsequent tsunami flooded critical on-site powersystems and resulted in the extended loss of the ultimate heat sink. This produced beyond designbasis conditions in units 1, 2 and 3 and ultimately led to core meltdown in the three units.
Since 2011, TEPCO has successfully implemented efforts at the plant to visualise andquantify the conditions of the reactors through external (i.e. muon tomography) and internal
inspections (i.e. primary containment vessel [PCV] entry). The results of the muon investigation
show that in all reactors, little fuel remains in the core region, confirming the idea of large core
meltdown occurring in all the reactors due to the total or partial loss of core cooling. In addition,
the PCV inspection showed that some degree of debris discharge ex-vessel was experienced in
all of the units.
After the accident, in December 2011, the Japanese government and TEPCO compiled the“Roadmap towards Restoration from the Accident at Fukushima Daiichi Nuclear Power Station”.
The roadmap was revised in June 2013 to accelerate the schedule for removing fuel debris andother restoration measures. As per the roadmap, the Japanese Ministry of Economy, Trade andIndustry’s Agency of Natural Resource and Energy has been promoting the Research andDevelopment Plan towards the Decommissioning of the Fukushima Daiichi Nuclear Power Plant
units 1-4, which includes an analysis of the accident progression at units 1-3 and their current
status.
In a number of the Nuclear Energy Agency (NEA) member countries, severe accident (SA)analysis codes had been developed after the accident at the Three Mile Island unit 2 reactor.
These computer codes were also used to analyse the accident at the Fukushima Daiichi Nuclear
Power Plant, and valuable information and estimations were provided regarding the accident
progression.
Taking these circumstances into account, with the participation of some NEA countries,
working plans were set up to conduct a benchmark study of the accident progression of units 1-3
during the Fukushima Daiichi Nuclear Power Plant accident using computer codes and methodsof analysis. This culminated in the “Benchmark Study of the Accident at the Fukushima Daiichi
Nuclear Power Station” (BSAF), an international programme proposed by the aforementioned
Agency of Natural Resource and Energy. The NEA Committee on the Safety of Nuclear Installations
decided to support the BSAF. After discussion among the related countries and organisations, theBSAF project was launched in November 2012. The project included participants from21 organisations in 11 countries.
Many years will be required to completely reveal the status of the units, including quantitativedebris distribution, and the accident progression that occurred in each of the three units. Thus, aphased approach has been applied, with two generic objectives. The first objective was to provide
information and analysis on the SA progression, fission product (FP) behaviour and source term(ST) release estimation of the Fukushima Daiichi accident, including a comparison to measured
plant data, to support safe and timely decommissioning of the Fukushima Daiichi Nuclear Power
Plant. The second objective was to contribute to the improvement of methods and models of the
SA codes applied by each participating organisation in order to reduce uncertainties in SA analysis
and to validate SA analysis codes using data measured through the decommissioning process.
The BSAF project was conducted in two phases. Phase I began in November 2012 and ran
until December 2014. It focused mainly on the phenomena of the core degradation and debris
relocation inside the reactor pressure vessels and the PCVs of units 1-3 during the first six days.
The project provided valuable insight on the accident progression and improved understanding
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INTRODUCTION

of severe accident phenomena that took place during the accident in order to provide
information on the current status of the units. Nonetheless, Phase I signatories recognised that,
despite the key insights gained into the understanding of accident scenarios, another benchmark
study was needed. Several reasons supported this initiative. Among them was the need tostrengthen some of the insights into the accident so that a better understanding could be gained,
and a desire to extend the time domain of the analysis to three weeks so that the analyses would
encompass the stable and continuous cooling periods in all units, and to capture all potential
melt relocations and main ST releases. This further required the inclusion of the reactor building
as an essential building in the accident progression and the many dose rate measurements thathave been recorded therein. Finally, a detailed review of FP release, transport and emission to
the environment during the Fukushima Daiichi accident was included in Phase II of the study.
In line with this reasoning, a second phase of the project was proposed to build upon the
success of the first phase, with the following major objectives:•

provide information and analysis on the SA progression, FP behaviour, ST estimation andcomparison with measured plant data within the first three weeks of the Fukushima
Daiichi accident (units 1-3) to support a safe and timely decommissioning;•

improve the methods and models of the computer codes aimed at reducing uncertainties
in SA analysis

Phase II began in April 2015 and was finished in March 2018, providing the participatingorganisations with a chance to revise or consolidate their interpretation of the accident
progression in each unit, to continue the comparison with the additional measurement datathat became available later in the accident, and to compare the predicted FP releases with direct
measurements or estimations generated by environmental dispersion codes. Additionally, thelatest information obtained by TEPCO investigations at the plant provided more details on thecurrent condition of the units, which provided guidance to the participating organisations in
the estimation of the accident scenario.
Table 1.1 provides a list of the countries and organisations that participated in Phase II, as
well as the analysis performed and the code(s) used.
This report summarises the progress made in Phase II. It begins with a discussion of thesevere accident codes, with an emphasis on the differences and similarities among theapproaches used by the participants. It then summarises the results of the ST estimations basedon measured plant data and the atmospheric FP transport and dispersion results based on theSTs calculated. Finally, it provides some conclusions and a synthesis of the lessons learnt, as
well as recommendations for potential forthcoming studies of the accidents.

Table 1.1. Participants, employed codes and analytical workin Phase II of the BSAF projectCountry

Partner

CodeAnalytical work

1
2

CanadaFinland

Canadian Nuclear Laboratories (CNL)
Technical Research Centre of Finland (VTT)MELCORMELCOR2.2

Unit 2
Units 1-2-3

3

FranceAlternative Energies and Atomic Energy Commission (CEA)TOLBIAC

Unit 1 (MCCI)

4FranceRadioprotection and Nuclear Safety Institute (IRSN)ASTEC V2.0 rev3 p1

Units 1-2-3

5
6
7

Germany
Japan
Japan

GRS
Central Research Institute of Electric Power Industry (CRIEPI)Japan Atomic Energy Agency (JAEA)GRS Code
MAAP 5.01
THALES 2

Inverse analysis
Unit 3
Units 1-2-3

8

Japan

Institute of Applied Energy (IAE)SAMPSON-B 1.4 beta

Units 1-2-3

9

Japan

Nuclear Regulation Authority NRA(S/NRA/R)

MELCOR 2.1

Units 1-2-3

MELCOR 1.8.6
MELCOR 2.1
SOCRAT/V3

Unit 1
Unit 2
Unit 1MELCOR 2.1-4803

Unit 1MELCOR 2.1_4203
MELCOR 2.1-5864
HYSPLIT

Unit 3
Units 1-2-3
Forward analysis

10 Korea

Korea Atomic Energy Research Institute (KAERI)

11 Russia

Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE)/Rosatom
Nuclear Security Council (CSN)/Research Centre for Energy, Environment andTechnology (CIEMAT)Paul Scherrer Institute (PSI)

12 Spain
13 Switzerland
14 United States

16

Nuclear Regulatory Commission (NRC)/Department of Energy (DOE)/Sandia
National Laboratories (SNL)

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RESULTS OF THE UNIT 1 ANALYSES

Chapter 2. Results of the unit 1 analyses

Overview of the unit 1 accident
As summarised earlier, the unit 1 accident proceeded with essentially no effective intervention by
the operators, unsuccessful attempts to inject emergency cooling water into the core, no operator
control of the reactor pressure vessel (RPV) pressure, and for the first 24 hours, no active control
of primary containment vessel (PCV) pressure. Because of this, core damage is calculated to havecommenced early in the accident, between 5 and 12 hours from scram, and was the first reactor
to have suffered severe damage. In this first 5-12-hour period, the RPV self-depressurised by someyet unidentified failure/leak of the pressure boundary. Because of the inability to reject decay heator inject cooling water, the PCV pressure ultimately became very high, which is believed to have
led to leakage of steam, hydrogen and fission products (FPs) to the environment in the PCV headflange region. In the project analyses, these key events were modelled using various assumptionsof either leakage at the reactor coolant system (RCS) pressure boundary or a larger failure, such asa main steam line (MSL) creep rupture or lower head attack by debris. All calculations predict
signatures generally consistent with the sparse RCS and PCV data points in this period. The
following sections describe the key results from participants’ analyses of each unit.

Thermal hydraulics and core degradation
The RCS pressure in unit 1 was only manually measured twice by operators between 5 and12 hours. These measurements (Figure 2.1a) indicate that an RPV depressurisation event occurred
sometime between 5 and 12 hours. The timing of the RCS leakage or RPV failure in the simulations
varied, depending on the hypothesis made by the modellers (Table 2.1). Most leakages, though,were predicted to occur between three and four hours, regardless of the path chosen. The
differences related to the leakage assumptions (boundary conditions in the codes) during thepressure vessel degradation are visible in the RPV pressure evolution.

Figure 2.1. Comparison of analyses results versus measurements for unit 1:
Reactor pressure vessel pressure and dry well pressure
A. Reactor pressure vessel pressure

B. Dry well pressure

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RESULTS OF THE UNIT 1 ANALYSES

From Table 2.2, two modelling teams (Korea Atomic Energy Research Institute [KAERI] andSandia National Laboratories [SNL]) modelled an MSL failure (at 5 hours and 6.1 hours respectively)
based on high RPV pressures and high core exit gas temperatures, resulting in an immediate
reactor depressurisation following the failure. All other teams assumed an RCS leak occurs at
some penetration(s) (safety relief valve [SRV], source range monitor, transverse instrumentation
probe or primary loop recirculation), producing a gradual loss of the RPV pressure. Some institutes(Research Centre for Energy, Environment and Technology [CIEMAT], Nuclear Safety Institute of
the Russian Academy of Sciences [IBRAE] and the Nuclear Regulation Authority [NRA]) evenassume two leakage pathways. Dry well (D/W) pressurisation is the result of water, steam andhydrogen (generated from zirconium-steam reactions during the core degradation) release fromthe RPV and the absence of any heat rejection from the PCV.

Table 2.1. Unit 1 event time of reactor
cooling system leak or failure
In hours
CIEMAT

IAE

IBRAE

IRSN

JAEA

KAERI

NRA

SNL

VTT

SRM/TIP pipe

1.0

–

3.6

–

2.47

–

4.1

–

–

SRV gasket

4.61

4.1

10.9

3.68

–

–

4.2

–

–

PLR pump leak

–

–

–

–

–

–

–

–

1.0

MSL failure

–

–

–

–

–

5.0

–

6.1

–

Note: SRM: source range monitor; TIP: transverse instrumentation probe; SRV: safety relief valve; PLR: primary loop recirculation;
MSL: main steam line.

There is some variation in the prediction of the RPV failure time and mode between thecalculations (Table 2.2). CIEMAT and the Technical Research Centre of Finland (VTT) predicted
the lower head failure time by melt through or penetration failure respectively, coincident withthe timing of the largest measured PCV pressure at around 11.5 hours. The Radioprotection and
Nuclear Safety Institute (IRSN), the IBRAE and the Institute of Applied Energy (IAE) justified thePCV pressure increase with the assumption that the rise was generated by the slumping of debris
in the lower head and predicted RPV failure a few hours later. The Japan Atomic Energy Agency(JAEA) predicted an earlier vessel failure and associates this pressure rise with the molten coreconcrete interaction (MCCI) progression. KAERI, the NRA and the SNL predicted the pressure riseeither later or before the measured point at the onset of the MCCI phase.

Table 2.2. Unit 1 event time of reactor pressure
vessel failure and failure mode
CEA CIEMAT

IAE

IBRAE

IRSN

JAEA

KAERI

NRA

SNL

VTT

Lower plenum
failure (hours)

–

11.58

14.97

15.10

17.16

7.45

13.50

13.6

12.5

11.42

Failure mode–

Melt Penetration Melt
Creepfailurethroughthrough

Total masses inpedestal (tonnes)

391

129

148

196

2822

User Penetration
Melt Penetration Melt
failurethrough specified
failurethrough
130

136

115

154

111

1. This value includes only the debris mass in two sumps, since the CEA’s calculation with TOLBIAC considers only the sumps. It does notconsider the mass above the pedestal and dry well floor.
2. The IRSN value includes the mass of all in-core materials and the mass of the reactor pressure vessel lower head and is a modelling
hypothesis.

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Figure 2.2 shows the H2 generation partially responsible for the PCV pressure increaseduring the first ten hours. As expected, the different predicted RPV depressurisation behaviours
match the available PCV pressure data in different ways (Figure 2.1b). Those calculations that
did not assume direct steam release into the D/W (KAERI and SNL) show a mild pressurisation
until a peak at the time of the MSL break and, from then on, a subsequent pressure increase aslong as some water remains in the RCS. Assuming the leakage in the RCS as a result of either
the temperature increase (SRV, source range monitor) or from equipment leakage (primary loop
recirculation pump) generally produced a slight depressurisation of the RPV and continuous
containment pressurisation from steam discharging into the D/W. It should be also highlighted
that the data predicted by most participants between 15 hours and the PCV venting time (around24 hours) is consistent with the measurements. The relative steady state during this period isinterpreted in most cases as a natural PCV pressure “self-regulation” balancing the pressure
increase resulting from MCCI with PCV leakage through the flange of the upper head of the D/Winto the reactor building. This is thought to be due to elastic stretching of the PCV head bolts
under high strain from the PCV internal pressure.

Figure 2.2. Unit 1 in-vessel hydrogen generation

The majority of institutes modelled instantaneous corium spreading within the pedestal orin the sump (2.5-3 metres) as a default in the code, and evaluated the erosion from this point.
Only the IAE calculated the process of corium spreading out the pedestal into the D/W. After the
suppression chamber (S/C) vent closure (24.68 hours), the reactor building experienced ahydrogen explosion (24.83 hours) and the continuous gas release from the MCCI drove a pressure
rise in the PCV until 50 hours, when the trend reversed and decreased for more than 200 hours
until it sharply started increasing again at 270 hours from the reactor scram (Figure 2.3a).
All the calculations follow the measurements (Figure 2.3a) by assuming a containment leakof a variable cross-section from 50 hours on, with the exception of the JAEA, which instead
postulated venting from the S/C. Among the institutes that assumed a containment leakage,
three assumed that the leak occurred at a penetration, while the other five assumed that the leak
was the result of MCCI erosion of the D/W liner, consistent with calculated MCCI erosion
behaviour. This assumption is further supported by the Tokyo Electric Power Company (TEPCO)
PCV investigations showing water leaking into the torus room, presumably from the sandcushion region outside of the PCV liner (TEPCO, 2013). Calculations rendered different decreasing
slopes, depending on the gas mass flow rate imbalance between the MCCI and leaking flow rate.
At around 270 hours, a few calculations captured the changing trend by imposing an effective
water injection (into the RPV and ultimately to the cavity floor); however, as shown in Figure 2.3b,they agree on an injection rate lower than 4.5 kg/s, but they differ on the specific injection profile.
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RESULTS OF THE UNIT 1 ANALYSESAt the time of the RPV lower head failure, most calculations predicted a total degraded core mass
released in between 120 tonnes and 190 tonnes (Table 2.2), resulting in a massive amount ofmaterial at very high temperatures being relocated into the pedestal. In addition, despite no
indication to confirm this based on plant inspections, almost all calculations predicted thatinteraction of the molten material with concrete was still ongoing at 500 hours (Figure 2.4).

Figure 2.3. Unit 1: Primary containment vessel evolution over 0-500 hour period
and alternative water injection mass flow rate
A. Primary containment vessel
evolution 0-500 hours

B. Alternative water injection
mass flow rate

Figure 2.4. Unit 1 concrete erosion depth(origin at the bottom of the sump)

Fission product release and behaviour
Codes track the most relevant FPs, either as individual radionuclides or grouped into classes
according to their chemical similarity. However, given the safety significance of cesium (Cs) andiodine (I) isotopes (Cs-137 and I-131, respectively) and the desire for a concise comparison, theanalysis here is focused on these two elements. The initial inventory of every element for unit 1
was specified by the benchmark co-ordinator and in the case of Cs and I, they were 154.0 kg and

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RESULTS OF THE UNIT 1 ANALYSES

11.98 kg, respectively. The results hereafter are presented as an FP fraction based on the initialcore inventory. Figure 2.5 shows the noble gas (NG) and I releases from the fuel in terms offraction from the initial inventory. Given their volatile nature, NGs, Cs and I releases are strongly
linked to core degradation. Most calculations predicted a fast release with or without subsequent
steps according to core degradation progression up to getting an asymptotic value, which in mostcases exceeded 80%.
The calculated results of the FP distribution in the containment were dependent on theassumptions made concerning the leaks from the RPV to the D/W, although in most cases
radionuclides entered the PCV through the SRV discharge into the suppression pool. Most of theCs and I was in aerosol form when it entered the PCV and once they were there, settled downfrom the atmosphere and deposited on D/W structures or were efficiently scrubbed after passing
through the suppression pool. Typically, most analyses predicted less than 10% airborne
concentrations, except in those cases that assumed a direct leakage from the RPV gas phase into
the D/W. It is worth emphasising that significant deposition was predicted on the D/W structuresand some remobilisation was noted in some calculations due to either PCV depressurisation or
heating up of the structures beneath the deposits.

Figure 2.5. Unit 1 mass fraction of fission products released from fuelA. Xeron (Xe)B. Iodine (I)For most calculations, Cs and I were very efficiently scrubbed in the suppression pool during
the early injection through the SRVs. At 35 hours, 10-80%, roughly, of Cs and I was estimated tohave been scrubbed into the S/C water (Figure 2.6c and d). Differences in the calculations were
due to the leak assumptions and the modelling of pool scrubbing geometry. Those calculations
in which most FPs were injected through T-quenchers reached higher values of decontamination
factor than those in which a good fraction of the FPs were transported into the pool through thedown comers, which includes a non-condensable rich flow.
In the Fukushima Daiichi units, the dose rate in the D/W and the S/C of the PCV wasmeasured by the containment air monitoring system (CAMS). Code results are only available for
FP masses: airborne, deposited on structures, and dissolved or dispersed in water. For
comparison with dose rates, a methodology of data conversion was developed by the operating
agent (IAE) within the Phase II project, which relates FP masses to dose rates depending on theirlocation and energy. In general, the conversion factor is larger for elements located either on
the structures or for airborne ones and for FPs that have larger energy or, more precisely, a larger
RHM value (Roentgen per hour per metre). The comparison accounts for decay and includes theeffect of Xe-133, Cs-137, Cs-134 and I-131. Other elements, such as tellurium, are neglected inthis conversion approach, but it is estimated that their consideration would have increased theestimated value by a factor of two. Details are provided in the final report.
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RESULTS OF THE UNIT 1 ANALYSES

Figure 2.6. Unit 1 mass fraction of fission product deposited on structuresin the dry well and solved in water in the suppression chamber
Deposited on the structures in the dry well
A. Cesium (Cs)B. Iodine (I)Solved in water in the suppression chamber
C. Cesium (Cs)

D. Iodine (I)

Figure 2.7 shows a comparison of the dose rate measurements by CAMS and the dose rates
converted based on the analysis results. Both the measurements and analyses show large
variations in the dose rate, but in general, the analyses seem to overestimate the dose rate in
the D/W, especially in the early stages of the accident. The dose rate in the S/C, on the other
hand, was underestimated by most of the analyses. The differences are presumably related tothe assumed transport paths from the RPV to the PCV bypassing the pool, but also different
scrubbing efficiency of fission products (this was not investigated in detail).
During the accident, monitoring posts (MP) recorded dose rates around the plant. In
addition, activity was measured on the land and in the ocean in the following days. Calculations
with atmospheric transport codes were performed by the GRS and the JAEA (Worldwide version
of System for Prediction of Environmental Emergency Dose Information, WSPEEDI code) using
inverse analysis to estimate the time and amount of FP releases based on these measurements.
The method and the results are further explained later in this report in Chapter 6 and in
Sogalla et al. (2019) and Katata et al. (2015).

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RESULTS OF THE UNIT 1 ANALYSES

Figure 2.7. Comparison of calculation results against measured data:
Unit 1 dose rate in the primary containment vessel
A. Dry well CAMS

B. Suppression chamber CAMS

For unit 1, the comparison of the release calculated with the severe accident codes and therelease based on the GRS and WSPEEDI analyses was relatively straightforward at the beginningof the accident because until around 35 hours (the estimated time of the start of unit 3’s coredegradation) there should not have been any other releases than those from unit 1. Figure 2.8
presents a comparison of the cumulative releases against the GRS and WSPEEDI analyses for Cs
and I. First, it should be noted that results of the GRS and WSPEEDI codes tend to reach the samevalue at the time of venting and subsequent hydrogen explosion, although with about an orderof magnitude difference in the case of I. Almost all calculations indicated a first release ataround ten hours, when the PCV head flange was predicted to start leaking and FPs werereleased to the environment through the assumed leakage from the building. Nevertheless,these releases were lower than the value estimated by the GRS and WSPEEDI. The integral
release after the PCV venting at 24.7 hours was predicted reasonably well in terms of the timeand magnitude, within a range of about two orders of magnitude (6·1013 Bq-5·1015 Bq). Most
calculations estimated that less than about 2% of cesium and 5% of iodine might have beenreleased to the environment, although the best fit to the inverse calculations predicts Cs releasebetween 0.5% and 1.0% and I release around 1%.

Figure 2.8. Unit 1: Comparison of predicted cesium and iodine release [Bq]versus backwards analyses (GRS and WSPEEDI)
A. Cesium

B. Iodine

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RESULTS OF THE UNIT 2 ANALYSES

Chapter 3. Results of the unit 2 analyses

Overview of the unit 2 accident
The unit 2 accident is characterised by the protracted and unanticipated operation of the reactor
core isolation cooling (RCIC) system for around 70 hours in a “self-regulating” mode, which is aphenomenon being pursued under separate research activities at the time of this report’s
publication. This situation developed, it is believed, when the uncontrolled full flow RCIC
injection overfilled the reactor pressure vessel (RPV) up to the steam lines, allowing two-phasesteam and water to be ingested into the RCIC turbine. Additionally, it was recognised that thetorus room had become flooded by sea water, which removed heat from the suppression pool.
It is also suspected that the water inside the suppression pool had thermally stratified, further
complicating the analysis of this accident. After around 70 hours, the RCIC system finally failed,prompting operators to depressurise the RPV and attempt low-pressure injection using fire
trucks. This action was ultimately unsuccessful in arresting core damage, owing to excessive
PCV back pressure and eventual loss of water level in the RPV and core damage.

Thermal hydraulics and core degradation
The RCIC system remained in operation in unit 2 despite the loss of power. However, without
power, it could not be controlled by the operators. When measurements resumed on 11 Marcharound 10:00 p.m., the RPV water level was found to be above the maximum measurable value.
Even though no direct evidence exists, it is suspected that the water level reached the main
steam line (MSL), driving steam and water into the RCIC turbine. Overall, the RCIC worked foraround 70 hours. The PCV pressure rise during this first accident period was found to be much
lower than expected given the cumulative decay heat within the PCV, especially if compared tothe accident progression in unit 3. In order to reproduce the too-low PCV pressure increase,additional phenomena had to be accounted for, such as temperature stratification in the
suppression chamber (S/C) and additional heat losses due to a partial flooding of the torus room
as a consequence of the tsunami. In general, all participants correctly predicted the main trendsof the data, specifically the high RPV water level during the RCIC operation and the RPV pressure
trend (Figure 3.1a). Some differences exist in the timing of the “turn-over point” in the pressure
at around ten hours, which occurs slightly earlier than the time reported when the RCIC system
was switched from taking water from the cooler condensate storage tank over to the warmer
suppression pool water. A physical explanation for earlier-than-expected turn-over is not
currently understood, although the increase in PCV pressure is consistent with the injection of
warmer suppression pool water into the RPV. The PCV pressure increase (Figure 3.1b) was also
captured well by most of the institutes, with a few outliers, such as the Japan Atomic Energy
Agency (JAEA) and the Canadian Nuclear Laboratories (CNL) (early on). The pressure increase is
the result of a balance of conditions driven by early tsunami flooding of the torus room, thermal
stratification in the S/C water and evaporation from the hot surface of the S/C pool, andincomplete condensation of the RCIC water-steam mixture at elevated S/C pool temperatures.
After the failure of the RCIC at around 67 hours, the RPV pressure in unit 2 increased to thepoint where one safety relief valve (SRV) started to open (Figure 3.1), discharging steam to the
S/C and reducing RPV pressure. As no water injection into the RPV was available, the RPV waterlevel decreased further due to the out flow through the cycling SRV. Steam was discharged fromthe SRV into the S/C at a low position and the lower colder layer of water mixed with the hotlayer on top, decreasing the surface temperature and thus condensing the steam and reducing
the pressure. Detailed S/C models are required to represent this well. None of the submitted
results using lumped parameter codes were able to properly match this phenomenon.
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RESULTS OF THE UNIT 2 ANALYSES

Figure 3.1. Comparison of analyses results
versus measurements for unit 2
A. Reactor pressure vessel pressure

B. Dry well pressure

The core heat-up and degradation phase was predicted to start at around 75 hours at highRPV pressure or shortly after RPV depressurisation. There is consensus that core degradation
began at this point and led to the pressure excursions in the RPV and the PCV – the so-called“RPV 3 pressure peaks period”. The specific assumptions taken by the various analysts differ.
However, all participants determined that the core degradation process led to non-condensablegas generation, melt relocation into the lower plenum of the RPV and consequential RPV failure(except for the Technical Research Centre of Finland [VTT]), melt release into the pedestal of thePCV, and in parallel fission product (FP) release from fuel/melt and transport into the PCV andfurther into the environment.
The first RPV pressure peak was attributed to steam and hydrogen production (Figure 3.2)
during core heat-up. The hydrogen generation was predicted to be stronger in the next phase
before the second RPV pressure peak due to more steam being available after reopening one SRVor due to material relocation within the RPV into a water pool. The PCV pressure trend until theend of the first peak was predicted well, even though the absolute values deviate by -0.1 MPa to+0.2 MPa from the measured values. The specific influence on the simulations of the different
S/C nodalisations could not be separated out from other factors. Theoretically, the generation of
steam and hydrogen has the same effect on the pressurisation of the RPV; however, hydrogen
has a very different impact on the containment response, as steam released into a subcooled S/C
may condense. At the time of SRV reopening (around 78.5 hours), the containment pressure
showed a sharp increase due to gas/hydrogen released from the RPV (Figure 3.2).
During the second pressure peak, the RPV pressure increased dramatically (Figure 3.2a) and
the PCV pressure followed consequently (Figure 3.2b). In order to replicate this simultaneouspressure rise, the calculations assume a partial SRV closure rather than a direct leakage from the
RPV into the containment. Only four institutes (the Central Research Institute of Electric Power
Industry [CRIEPI], the Institute of Applied Energy (IAE) and the Sandia National Laboratories (SNL),and to some extent the JAEA) show an RPV pressure trend similar to what was measured(Figure 3.2a), while the PCV pressurisation (Figure 3.2b) in general was not well predicted, except
by the IAE and the JAEA. The pressure evolution during this period is assumed to be the result of
direct slumping of core debris from the core into the lower plenum (for the IAE, JAEA and SNL).
The pressurisation calculated by CRIEPI appears to be the result of water reflooding into the coreregion, which started early. At this time, CRIEPI computes a comparatively intact core, and steam
is generated through quenching of the fuel rods. At the end of the second pressure peak, the PCV
pressure reached its highest value, close to 0.8 MPa (Figure 3.2b), which was only reproduced wellby the IAE and the JAEA. The maximum value, with a different timing, was reproduced by the
Nuclear Regulation Authority (NRA) and the Radioprotection and Nuclear Safety Institute (IRSN).
The high containment pressure remained for almost ten hours and, according to the analyses,

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RESULTS OF THE UNIT 2 ANALYSES

was caused by large amounts of hydrogen released into the PCV (Figure 3.2c). Three institutes(the IAE, the IRSN and the NRA) predicted greater than 800 kg of hydrogen and were able to comeclose to predicting the measured PCV pressurisation. The other institutes that predicted lower
hydrogen generation did not predict the large pressurisation of the PCV shown in themeasurements. It was not analysed whether other phenomena may have caused such high PCV
pressure during this period of the accident; however, direct leakages from the RPV into the drywell (D/W) of the PCV may be one possible phenomenon.

Figure 3.2. Three reactor pressure vessel pressure peaks in unit 2
A. Reactor pressure vessel pressure during coredegradation

B. Primary containment vessel dry well pressureduring core degradation

C. Hydrogen mass generated during core degradation

Note: Legend reported in Panel c.

During the third RPV pressure peak, complete closure of the SRV was assumed, as the PCV
pressure did not increase in parallel. Almost all calculations reproduced the RPV pressure risewell, with the exception of the JAEA. The third pressure peak was likely caused by debris
quenching in the lower RPV head associated with alternative water injection (AWI), whichbecame available in unit 2 at about 75 hours at the earliest, according to measured data. In theanalyses, the AWI data (timing, amount) have been widely varied to best match the measured
plant parameters. The debris masses contained in the RPV lower head after this time periodspan a large range, from a few tonnes up to 160 tonnes; the RPV itself is still predicted to be
intact at this point shortly after the three RPV peaks.
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RESULTS OF THE UNIT 2 ANALYSES

Since the time when the PCV pressure reached its highest value (at around 80.7 hours), theoperators tried unsuccessfully to vent the PCV. A fast PCV pressure decrease was measured atabout 89 hours (Figure 3.3), although the exact timing is unknown, due to a significant gap in data
recording between the maximum value and the next low value. All institutes assumed that thispressure drop was due to a permanent failure/leakage occurring in the containment at 89 hours(typically at the head flange, but also at other locations as discussed in Table 3.1). The PCV
failure/leakage reduced the pressure quickly (Figure 3.3) and in the long term, the PCV pressure
was predicted to remain at a low level. A single PCV pressure measurement at about 118 hours
indicates another possible peak, but this measurement was not considered in the analyses, as it
seemed likely to be erroneous. In addition to the PCV failure at the head flange, two institutes(the CNL and the JAEA) also model an S/C failure in the pool region. The S/C failure is predictedbased on evidence obtained at the plant from measurements and investigations that confirmed
the water level in the torus room and in the torus itself to be nearly identical (TEPCO, 2014).

Figure 3.3. Unit 2 primary containment vessel dry
well pressure during a longer term

Table 3.1. Unit 2 primary containment vessel and reactor pressure vessel (RPV)
lower head failure and debris mass released from the RPV

LH failure
time (hours)
LH failuremode
PCV failure(hours)
PCV failure
mode

CNL

CRIEPI

IAE

IRSN

JAEA

KAERI

NRA

SNL

VTT

104.57

129.0

93.36

95.32

92.0

441.19

85.8

83.0

–

RPV
creep

Penetration
failurePenetration
failure

88.55

88.57

87.0//89.2

Penetration Penetration
failureejection
89.0

88.54

Head flange Head flange
leakageleakage

Penetration Penetration Penetration
failurefailurefailure
88.89

Head
X-6
X-6
Head flange
flange penetration penetration/
leakageleakageleakagehead flangeNo
failure

N/A

N/A

89.0

N/A

N/A

Head
flangeleakage

S/C water
leakage(hours)

89.0

N/A

N/A

N/A

88.0

N/A

N/A

N/A

N/A

Total masses
in pedestal(tonnes)

78

3991

35

16

N/A 0

186

157

111

N/A 0

1. Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia,
LLC, a wholly owned subsidiary of Honeywell International Inc., for the US Department of Energy’s National Nuclear Security Administration undercontract DE-NA0003525. SAND2018-8356 R.
Note: LH: lower head; PCV: primary containment vessel; S/C: suppression chamber.

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Due to the large mass of debris in the lower head and the relatively low AWI into the reactor
vessel, the lower head was predicted to fail between 83.0 hours and 129.0 hours for most
calculations, except for Korea Atomic Energy Research Institute (KAERI), with a very late failure
at 441 hours, and the VTT, without RPV failure (Table 3.1). The debris mass released into thepedestal for all calculations except KAERI varies from 16 tonnes for the IRSN up to 161 tonnes for
the NRA. The molten core concrete interaction (MCCI) radial/axial erosion of the cavity/pedestal,and continuous release of non-condensable gases depend strongly on the modeling assumptions.
After the RPV failure and melt release, CRIEPI, the NRA and the SNL predicted unterminatedconcrete erosion; KAERI also predicted this condition, but at a much later point in time. None ofthe other calculations predicted an MCCI. The prediction of little or no MCCI seems to beconsistent with evidence found in Tokyo Electric Power Company’s (TEPCO) plant investigationsof unit 2.

Fission product release and behaviour
In unit 2, a large release from the damaged fuel of noble gas (NG) (Xe), Cs and I, in general above
80% of the initial mass, was predicted by the calculations during the early core degradation phaseprior to the first RPV pressure peak. Most calculations predicted that fission product (FP) releasefrom fuel started at around 75 hours and was almost complete at the end of the second RPVpressure peak (around 82 hours). Only KAERI predicted a much slower release of NG and volatile
FPs from the fuel. The predicted FP release from fuel for Xe and I are provided in Figure 3.4.

Figure 3.4. Unit 2 mass fraction of fission productreleased from fuel a) Xe and b) I
A. Xeron (Xe)B. Iodine (I)Because FP release from fuel was predicted to start after the RCIC failed and no leak path
was assumed connecting the RPV to the D/W, then from the point of RCIC failure up until RPV
failure, the SRVs were the only modelled connection path from the RPV to the PCV. Volatile FPs
were predicted to have been transported from the RPV through an SRV and its sparger into the
S/C. Predictions of FP behaviour in the containment were calculated for the D/W and S/C anddivided into airborne FPs (not presented in this report), deposited FPs on structures (wall, piping
and floor), and FPs in the water.
NGs were released to the S/C through the operating SRV during the RPV 3 pressure peaks
period and partially transferred to the D/W through operating S/C vacuum breaker(s). Of majorinterest to the FP behaviour in containment are the deposition of Cs and I on D/W structures
and the amount of Cs and I solved in the S/C water (Figure 3.5). The predicted Cs and I depositionon the D/W structures (Figure 3.5a and b) for most calculations is a final deposition of less than
3%, with some exceptions. The IAE predicted about 23% of I deposited on structures and the CNL
predicted about 7%. The NRA predicted a strong initial peak in the deposition of both Cs and I,

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RESULTS OF THE UNIT 2 ANALYSES

but subsequently calculated that this deposited mass was revolatilised within the next 40 hours.
The predictions for Cs and I retention in the S/C water pool by pool scrubbing (Figure 3.5c and d)
were very different across the calculations, from less than 10% (KAERI) to about 75% (IRSN) in
the long term. The JAEA showed an even larger peak value, about 90%, during core degradation,
which decreased due to leakage assumed at the bottom of the S/C. The variations in thepredictions were caused by differences in the retention of Cs and I in the RPV, and by the factthat some calculations modelled water leakages from the S/C to the torus room (the CNL and
the JAEA). The calculations by the CNL, the JAEA and the NRA showed a continuously decreasing
mass of Cs and I retained in the S/C water in the longer term due to the water leakages from the
S/C. Additionally, less contaminated water injected by AWI into the RPV flows over from theD/W, which diluted the S/C water inventory.

Figure 3.5. Unit 2 mass fraction of fission products deposited on structuresin the dry well and solved in water in the suppression chamber
Deposited on structures in the dry well
A. Cesium (Cs)B. Iodine (I)Solved in water in the suppression chamber
C. Cesium (Cs)

D. Iodine (I)In unit 2, the analyses predicted that the main FP release into the PCV occurred during the
RPV 3 pressure peaks period and later into the environment, when the PCV depressurisation
started, at about 89 hours. The release path of airborne FPs was typically assumed through a

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RESULTS OF THE UNIT 2 ANALYSES

leakage of the PCV head flange, into the reactor well and further through the fifth floor of theintact reactor building upper part and out the failed blowout panel into the environment (seeTable 3.1).
Unit 2 had relatively detailed containment air monitoring system (CAMS) measurement
data in the D/W and S/C (two systems A and B each) available during the accident. During the
expected core degradation period, starting about 75 hours after scram, dose rate measurements
increased quickly, from 10 Sv/h and 3 Sv/h to around 100 Sv/h and 10 Sv/h for the D/W CAMS
(Figure 3.6a) and S/C CAMS respectively (Figure 3.6b). The FP release from fuel predicted by thecalculations started at about the same time as the CAMS data, for both the D/W and S/C, showed
this first large increase. This is seen as a confirmation of the analyses. The first FP releases were
through the operating SRV into the S/C. The S/C CAMS measured a peak dose rate at about
80 hours (Figure 3.6b). The peak appears to be overpredicted by most of the calculations by four
to five times, but this is still within the uncertainty of the conversion methodology (estimating
an equivalent dose from the predicted FPs in the D/W or S/C). With regard to the D/W CAMS
measurements, the predicted results for most calculations are in better agreement with themeasurements, but overpredictions are also seen (Figure 3.6a). Dose in the D/W before the RPV
failure is assumed to be caused by the transport of NGs and aerosols from the S/C through thevacuum breakers. The dose measured in the D/W increases by another peak at ~95 hours, whichmight indicate RPV failure.

Figure 3.6. Comparison of analyses results against measured datain the unit 2 dose rate in the primary containment vessel
A. Dry well CAMS

B. Suppression chamber CAMS

Differences in the calculations were pronounced for the airborne release of Cs and I into theenvironment (Figure 3.7). The final predicted values for Cs release were lower than 0.1% (theIRSN and the IAE), about 0.1% (the VTT and the JAEA) and 1.5-5% (all others). For I, the split of
the results is different: less than 0.5% (the IRSN and the NRA), about 2-5% (the JAE, KAERI and
the SNL), 9% (the CNL), and about 17% (the IAE). Almost all NGs were released through PCV
leakages into the environment within about 10 hours after the PCV failure, at about 89 hours.
The predicted FP release source term from the severe accident (SA) code calculations wascompared against inverse analyses provided by the GRS and WSPEEDI (Worldwide version of
System for Prediction of Environmental Emergency Dose Information). During the RPV 3
pressure peaks period and the strong core degradation in unit 2, the wind was blowing towards
the land. The strong releases (dose rates) measured during this time were attributed to releases
from unit 2 (Sogalla et al., 2019; Katata et al., 2015), as the main core degradation periods in theother two units occurred much earlier. The mechanisms or release paths from the unit 2
containment are still unclear. In general, airborne releases from unit 2 are expected to occur
when the PCV pressure is high (likely through PCV head flange leakage). Unit 2 PCV venting was
attempted, but reported unsuccessful.
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Figure 3.7. Unit 2 mass fraction of fission products
released into the environment
A. Cesium (Cs)B. Iodine (I)

Figure 3.8 shows a comparison of Cs releases predicted by the calculations versus the GRS
and WSPEEDI backwards analysis during and after the three RPV pressure peaks period (up until
120 hours); the RPV pressure is also included in the figure for comparison (green triangles). The
integrated release data of the GRS and WSPEEDI analyses is only available starting at around75 hours; earlier releases are not considered.

Figure 3.8. Unit 2 comparison of predicted cesium and iodine releases [Bq]versus backwards analyses (GRS and WSPEEDI) and
RPV pressure (RPV 3 pressure peaks period until 120 hours)
A. Cesium (Cs)B. Iodine (I)Even though all institutes predicted the start of the FP release from the fuel at the time of
the first RPV pressure peak, only four (the CNL, the IRSN, the NRA and the VTT) predicted anincrease in the radio nuclide release to the environment at this time as well, which is suggestiveof a small leakage from PCV through the reactor building upper part (likely the head flange) tothe environment. These predicted results are still much lower than the values estimated by the
GRS and WSPEEDI analyses.

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For the initial release (at ~77 hours), the NRA and the IRSN predicted release values around
1.0 E+11 – 1.0 E+12 Bq for Cs and one order of magnitude more for I (the IRSN only), while the CNL
and the VTT showed releases starting at about 82 hours, but at several orders of magnitude lower.
Spikes in Cs release were also predicted by the JAEA (~1.0E+13 at 82 hours) and the SNL (~1.0 E+11
at 84 hours). The IRSN and NRA calculations predicted a large PCV pressure at ~79 hours
coinciding with the end of the first RPV pressure peak, much earlier than measured. This might
be the reason for the predicted early FP release from the PCV. All institutes showed significant
FP releases from ~89 hours on when the PCV depressurisation starts and PCV head flange leakage
was assumed (see Table 3.1). These release predictions were almost in accordance with anothersignificant increase of the integral releases estimated by the GRS backwards calculation, while
the WSPEEDI calculation showed this increase later, between ~95 and 105 hours. After the PCV
had completely depressurised at ~95 hours, results by the CNL, the JAEA, KAERI, the NRA and the
SNL showed a continuous increase of the releases (though with decreasing slope), while in allother calculations, the releases stopped. At ~120 hours the agreement between the estimations
by the GRS and the WSPEEDI analyses and the predicted releases from the calculations is nottotal. For Cs, about half of the calculation results were still significantly below the values that the
GRS and WSPEEDI estimated, while others were in agreement (the CNL and KAERI) or slightly
above (the NRA and the SNL). The result was similar for I releases, with most calculations yielding
results a few orders of magnitude below the GRS and WSPEEDI estimates, one in agreement (JAEA)
and three slightly above (the CNL, KAERI and the SNL).

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Chapter 4. Results of the unit 3 analyses

Overview of the unit 3 accident
The unit 3 accident is characterised by normally functioning heat removal from the reactor
pressure vessel (RPV) in the first 35 hours of the accident, first by the reactor core isolation cooling(RCIC) coolant injection system, then subsequentially by the much higher capacity high pressure
coolant injection (HPCI). While these systems performed normally, their operation could only
serve to replace water lost from the RPV and to move heat from the reactor fuel to the suppression
chamber (S/C) water and the primary containment vessel (PCV) atmosphere. The PCV pressure
increase due to steady steam injection was controlled in a later accident phase by spray activationin the S/C. During this first accident period, all of the participants predicted the main signaturesin the RPV and PCV reasonably well. After the HPCI was stopped, unit 3 experienced a period of
about ten hours with no coolant injection into the reactor, which is thought to have caused thereactor water level to drop below the bottom of active fuel. Most of the analyses show that coredegradation started during this time, and was accompanied with hydrogen generation, causing
further pressurisation of the PCV. The period after reactor depressurisation is postulated to haveoccurred by spurious operation automatic depressurisation system (ADS), at about 42 hours, up
until the hydrogen explosion in the unit 3 reactor building, at 68 hours, was characterised by anumber of attempts by operators to depressurise the containment and inject water into the RPV
using fire trucks.

Unit 3 thermal hydraulics and core degradation
Unit 3 had direct current (DC) power available after the tsunami, and consequently, a large
amount of measured data is available (e.g. water level values, RPV pressure, PCV pressure andlower head temperature) for long periods of time. Several containment vent actuations wereattempted as well as coolant injections by means of the RCIC, the HPCI and the alternative water
injection, with some interruptions. The timings of the coolant injection to the reactor as well as
containment vent attempts were recorded by the operators and used by the analysts as boundary
conditions, depending on their interpretation of the accident. Even though the approximate
timing of the venting and coolant injections are known, the actual success of these operations isstill considered uncertain.
For the first 20 hours after the accident initiation, the unit 3 core was cooled by the RCIC,
pressure in the RPV was regulated by the safety relief valves (SRVs; see Figure 4.1a) and thewater level in the reactor stayed relatively constant at a high level above the top of active fuel.
During this time, the containment pressure (Figure 4.1b), increased continuously, which isassumed to be due to thermal stratification in the suppression pool, leading to high pool surface
temperature. After about 20 hours, the RCIC stopped automatically due to high pressure in the
suppression pool. As a result, the water level in the reactor started to decrease. The HPCI systemstarted about one hour later due to the low water level in the reactor. After the HPCI operation
started, the water level in the reactor increased again while the pressure in the reactor
decreased sharply due to large amounts of steam extraction by the large capacity HPCI turbine.
The calculations assumed that the HPCI performance started to degrade at around 30 hours.
The HPCI was manually stopped at 36 hours by the operators, after which they attempted todepressurise the RPV by opening the SRVs, which eventually failed. Most of the calculations
could reproduce the RPV and PCV pressure trends in a satisfactory way during this time.
After coolant injection by the HPCI stopped, there was a period of some ten hours with no
coolant injection into the reactor. During this time, the water level in the reactor dropped below
the bottom of active fuel and the reactor pressure increased rapidly. Most of the calculations
predicted that major core degradation started during this time, accompanied by hydrogen
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generation (Figure 4.2), resulting in an increase in the dry well (D/W) pressure up to 0.4 MPa. The
reactor pressure reached the set point of the SRVs, and after several hours of high RPV pressure,
the reactor is hypothesised to have become depressurised by the ADS at 42 hours. The ADS wasthought to have accidentally activated due to high D/W pressure, which is believed to be theresult of the continuous hydrogen and steam generation in the RPV. The ADS included
simultaneous activation of 6 SRVs producing a PCV pressure increase up to 0.6 MPa andconfirming that this first RPV vent activation was successful. The containment pressure wasdecreased by venting from the S/C gas space soon after the ADS activation at around 42.5 hours.

Figure 4.1. Unit 3 reactor pressure vessel pressure
and primary containment vessel pressure
A. Reactor pressure vessel

B. Primary containment vessel pressure

Figure 4.2. Unit 3 in-vessel hydrogen generation

The period after reactor depressurisation at 42 hours until the hydrogen explosion that took
place in the unit 3 reactor building at 68 hours involved repeated attempts by the operators to
depressurise the containment and inject water by means of fire trucks. As presented in the latest
Tokyo Electric Power Company (TEPCO) reports, the only other confirmed vent occurred at
45 hours (TEPCO, 2017a). Most of the calculations predicted major core degradation and coreslumping during this time, eventually leading to the failure of the RPV. The timing and mode of
the RPV failure assumed/predicted by different calculations are shown in Table 4.1. The timing of
the vessel failure varies some depending on the boundary conditions and codes used, with mostparticipants within a range of 43-58 hours, with two later predictions at 73 hours (the Paul Scherrer
Institute – PSI) and 102 hours (Central Research Institute of Electric Power Industry – CRIEPI).

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Table 4.1. Unit 3 assumed lower head failure time and mode of failure; total debris
mass released from the reactor pressure vessel to the containment
Time of failure
Mode of failure
Mass in pedestal(tonnes)

CRIEPI

IAE

IRSN

JAEA

101.95
Penetration
failure

55.20
Creep
failure

55.37
Creep
failure46.48
Vessel
melt

244

105

51

188

NRA

PSI

49.40
73.05
Penetration Penetration
failurefailure
65

21

SNL

VTT

58.03
User
specified

43.30
Penetration
failure205

224

After the hydrogen explosion at 68 hours, the containment pressure remained high until
about 130 hours (Figure 4.3). This is partly due to further gas generation by the oxidation of corium
and metallic structures in the containment, as predicted by several calculations, and partly due to
steam generation from the continuous water injection. Due to the coolant injection, several
calculations predicted that the water level in the PCV reached the main steam line penetration in
the dry well by the end of the calculation. The possibility of this has been observed in the unit 3
plant inspections, as a water leakage has been observed at the level of the main steam line. Mostof the calculations reproduced the containment pressure trend during this time relatively
accurately.

Figure 4.3. Unit 3: A. containment pressure; B. concrete erosion depth;
C. ex-vessel hydrogen generation

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From the time of vessel failure to the end of the simulation (~500 hours), the majority of thecalculations predicted a large mass of debris discharged into the containment, followed bycontinuous molten core concrete interaction (MCCI) (Table 4.1). A large amount of debris on thecontainment floor has been confirmed through unit 3 plant inspections with video cameras
mounted on submersible robots (TEPCO, 2017b). Three calculations show smaller amounts ofmaterial released to the containment. The variation in the results of the calculations was large
for both the timing and the magnitude of the corium release from the reactor pressure vessel tothe containment. TEPCO’s latest investigations of unit 3 containment (TEPCO, 2017b) indicatethat the debris mass in the containment is likely closer to the higher values given by the
analyses than the lower ones. All of the calculations except for one (the PSI) predict that the
MCCI started once the corium was released to the containment floor. However, the MCCI in theunit 3 containment has not been confirmed nor disproved by plant investigations.

Fission product release and behaviour
The core degradation accompanied by fission product (FP) release from the fuel was predicted
to have initiated prior to the RPV depressurisation by the ADS in most of the analyses. The
volatile FP release from the fuel was predicted to progress rapidly once the core degradation
began (Figure 4.4). Cs and I (Figure 4.4b) release was very similar to that of Xe (Figure 4.4a). Mostof the calculations predicted a similar profile of a fast release, sometimes in steps following thecore degradation progression, up to an asymptotic high value bracketed between 90% and 100%of the element inventory. A slightly lower total release of about 80% was calculated by the PSI
and the SNL for I.
Before the ADS, FPs were transported to the suppression pool through SRV spargers andwere assumed to be scrubbed efficiently. As a consequence, large amounts of fission products
accumulated in the suppression pool. Noble gases (NGs) were transported through the pool andremained partly in the gas space of the S/C, and were partly transported to the dry well as the
S/C pressure increased.
The distribution of the FPs within the PCV was influenced by the assumptions of the analysts
in modeling RPV leakages to the D/W. As presented in Table 4.2, some of the calculations assumed
direct leakages from the RPV to the D/W (such as primary loop recirculation or recirculation pumpseal leakage), which allowed FPs to be transported from the RPV to the containment without being
scrubbed in the suppression pool. Two calculations (the JAEA and the Technical Research Centreof Finland [VTT]) assumed early leakage from the RPV through a pump seal. Other calculations
predicted direct leakages from the RPV to the D/W at around the time core degradation started in
Unit 3. The SNL calculation, as in its unit 1 analysis, predicted a main steam line failure aroundthe time core degradation starts and subsequent large releases of FPs to the dry well.

Table 4.2. Unit 3 assumed leakages from the reactor pressure vessel
into the primary containment vessel and timing
Hours
CRIEPI

IAE

IRSN

JAEA

NRA

RPV leak (MSL creep)
RPV leak (SRV leak)

PSI

SNL

VTT

42.30
43.00

42.20

RPV leak (pump seal)

5.00

RPV leak (TIP leak)

39.75

6.33
41.90

Notes: RPV: reactor pressure vessel; MSL: main steam line; SRV: safety relief valve; TIP: transverse instrumentation probe.

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Once in the D/W, the FPs may be transported into the reactor building if the containment
integrity is compromised. All the calculations assumed that once the containment pressure
increased to a certain level, the dry well head would lift due to elastic stretching of the headbolts, opening a gap at the D/W head flange and producing leakage from the D/W into thereactor well. The reactor building was not designed as a pressure-tight structure and, as such,releases to the reactor building were followed by a minor release to the atmosphere throughbuilding leakage. Once the hydrogen explosion in unit 3 occurred (~68 hours), a large direct
leakage path from the reactor building to the environment was created.

Figure 4.4. Unit 3 mass fraction of fission
product released from fuelA. Xeron (Xe)B. Iodine (I)Large fractions of Cs and I were predicted to be retained in the suppression pool water(Figure 4.5) in all of the calculations. Several calculations showed a large fraction of Cs in the
RPV due to deposition of Cs compounds on the reactor walls, either by chemisorption or aerosol
deposition. A fraction of FPs was also calculated to be deposited on the walls of the dry well
(Figure 4.5a and b). Note that the results of one analysis (the SNL) are above the scale of y-axis
shown in Figure 4.5a and b.
The dose rates in the D/W and S/C of the containment were measured during the accident
by the containment air monitoring system (CAMS). In unit 3, CAMS measurement data wereavailable around the time of the hydrogen explosion (60-70 hours), and from 150 hours to theend of the simulation time (500 hours). The timing and magnitude of the CAMS measured dose
rates were compared with those predicted by the code calculations. For the comparison, theconcentration of the different radionuclides in the containment as calculated by the severe
accident codes was converted to a dose rate considering the specific geometry of the CAMS
measurement (Uchida et al., 2021). Calculations that assumed an early and large leakage fromthe RPV to the D/W and subsequent large deposition of fission products on the D/W structures
tend to overpredict the dose rate in the dry well significantly. The other calculations, whichassumed a smaller early leakage from the RPV to the D/W, seemed to predict the increase in thedose rate in the D/W too early; however, given the lack of dose rate measurements before
60 hours, nothing conclusive can be determined. The calculations that did not assume any
direct release of FPs from the RPV to the D/W before 60 hours underestimate the dose rate in
the D/W by a large extent. Based on the dose rate measurements at around 60 hours, about 5%
of Cs and I was in the D/W at this time, mainly as a result of a direct transport of Cs and I fromthe reactor vessel to the D/W.

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Figure 4.5. Unit 3 mass fraction of fission products deposited on structuresin the dry well and solved in water in the suppression chamber
Deposited on structures in the dry well
A. Cesium (Cs)B. Iodine (I)Solved in water in the suppression chamber
C. Cesium (Cs)

D. Iodine (I)Comparison of the calculation results with the S/C CAMS measurements shows that almost
all calculations overestimate the dose rate in the suppression chamber. Note that some of theresults are not visible in Figure 4.6 because their values are higher than the y-axis scale. Possible
reasons for the overestimation may be a different S/C water level than assumed by the conversion
method or other phenomena, such as wall washing, that increase FP remobilisation from thestructures to the pool that are not accounted for in the code calculations. Overestimation may also
be due to unrealistically high calculated pool scrubbing efficiency of the FPs in the suppression
pool.
The calculations predicted that the main FP release to the atmosphere in unit 3 took place
during the first two containment vents and at the time of the hydrogen explosion. In addition,
one calculation (the SNL) showed a continuous release of Cs and I through a D/W head flange
leak after the hydrogen explosion, and two calculations (the JAEA and the NRA) showed aconsiderable release at around 220 hours in connection with the pressure increase in thecontainment at that time (Figure 4.7).

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RESULTS OF THE UNIT 3 ANALYSES

Figure 4.6. Comparison of calculation results against measured data
for the unit 3 dose rate in the primary containment vessel
A. Dry well CAMS

B. Suppression chamber CAMS

Figure 4.7. Unit 3 mass fraction of fission
products released to the atmosphere
A. Cesium (Cs)B. Iodine (I)About 80-100% of the NGs were released to the atmosphere up until the hydrogen explosion
at 68 hours. The calculations vary in their predicted timing of the release, depending on theaccident progression and the assumed transport path for the FPs. Two calculations (the JAEA andthe NRA) showed continued release of NGs after this time. Differences in the calculations are
more pronounced for the release of Cs and I. Three calculations (the IRSN, the NRA and the VTT)show a fast release of 3-5% of Cs to the atmosphere during the first containment vent that
occurred shortly after the ADS actuation. The majority of the calculations assumed transport ofCs from the RPV to the containment through SRV cycling with efficient scrubbing of Cs in the
suppression pool, and a subsequent release of less than 0.5% Cs until the hydrogen explosion.
Iodine release was similar, with the release fraction being on average slightly higher than that of
Cs. One calculation (the IRSN) predicted a fast, high release of I at the first containment vent,reaching a total of 13%. The other calculations fell into two groups: three calculations (the NRA,
the SNL and the VTT) predicting I release to the environment of 4-9%; the remaining four
calculations predicting about 2% or less.

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Unit 3 is a special case, as a fraction of the gas in the unit 3 containment was transported tothe unit 4 reactor building through the common stack during attempts to vent unit 3. A hydrogen
explosion took place in the unit 4 reactor building about 19 hours after the explosion in unit 3.
TEPCO analysis shows that the hydrogen that caused the explosion in unit 4 was transported fromunit 3 through the ventilation channel during venting of the containment of unit 3 (Uchida et al.,2021). According to TEPCO’s analysis, approximately 20-35% of the vented gas could have been
diverted to the unit 4 reactor building during the vent actions. This transport path is not accounted
for in the Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF)
calculations and would reduce the release to the atmosphere due to the deposition of FPs in theventilation lines and in the unit 4 reactor building.
Figure 4.8 shows a comparison of cumulative releases of Cs and I as predicted by the severe
accident code calculations and the releases estimated by the GRS and WSPEEDI (Worldwide
version of System for Prediction of Environmental Emergency Dose Information) backwards
analysis based on environmental measurements and distribution in the atmosphere (Sogalla
et al., 2019; Katata et al., 2015). The comparison covers the time period from 40 hours to 75 hours
after the accident initiation, which is when the major contributions to the FPs releases are
believed to have come from unit 3.

Figure 4.8. Unit 3 comparison of predicted cesium and iodine releases [Bq]versus backwards analyses (GRS and WSPEEDI)
A. Cesium (Cs)B. Iodine (I)The comparison shows that the calculations with a large early release of Cs and I tend to
significantly overpredict the integral release estimated by the GRS and WSPEEDI analyses. The
rest of the calculations predicted a similar order of magnitude as the GRS and WSPEEDI analysis,
indicating that the releases to the environment up until the hydrogen explosion should havebeen less than 0.5% of Cs and less than 2% of iodine. However, during the timeframe of the main
release events in unit 3 (i.e. the first containment vents and the hydrogen explosion), thedominant wind direction was towards the ocean, thereby carrying the released FPs away fromthe land. This introduces additional uncertainty, as the releases calculated by the inverse
methods during this time period rely on the activity measurements of samples of the ocean water.

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ATMOSPHERIC TRANSPORT AND DISPERSION: RESULTS FOR FUKUSHIMA DAIICHI UNITS 1, 2 AND 3

Chapter 5. Atmospheric transport and dispersion:
Results for Fukushima Daiichi units 1, 2 and 3

Many of the Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station(BSAF) Phase II participants provided source terms (STs) to be evaluated by Sandia National
Laboratories (SNL) 1 by applying HYSPLIT (Draxler and Hess, 1998; Draxler, 1999; Air Resources
Laboratory, 2016, 2017) to treat atmospheric transport and dispersion. The objective was toestimate the deposition pattern that would have resulted from the predicted ST. For theparticipants who provided results for all three units, the overall deposition pattern can becompared with the observed deposition pattern. For the participants who submitted STs for one
or two units, the results can only be compared with each other. Atmospheric transport
calculations were performed for a single isotope, Cs-137. It is the primary isotope of concern for
long-term contamination and is relatively easy to measure due to the strong gamma signal
produced from its short-lived decay product, Ba-137m. All of the atmospheric transport
calculations used the actual location of each of the three units; the releases were not presumed
to emanate from the same location. Also, when they were provided, release energies were
accounted for in the analysis, so plume lofting was considered. Finally, aerosol size distribution
data were considered for the purposes of estimating deposition. In some cases, aerosol size
distribution can significantly influence deposition patterns.

Unit 1Five of the organisations participating in Phase II of the BSAF (CIEMAT, the Institute of Applied
Energy [IAE], the Radioprotection and Nuclear Safety Institute [IRSN], Korea Atomic Energy
Research Institute [KAERI] and the SNL) submitted a unit 1 source term for atmospherictransport analysis. A comparison of these five source terms is shown in Figure 5.1. Of the fiveSTs, three predicted a significant deposition over land (CIEMAT, KAERI and the SNL). The othertwo (the IAE and the IRSN) predicted nearly all the deposition over the ocean.

Unit 2Six of the organisations participating in Phase II of the BSAF project (the Canadian Nuclear
Laboratories (CNL), Central Research Institute of Electric Power Industry (CRIEPI), the IAE, the
IRSN, KAERI and the SNL) submitted a unit 2 source term for atmospheric transport analysis.
A comparison of these six source terms is shown in Figure 5.2. All except those from the IAE
and the IRSN predicted significant deposition over land.

Unit 3Six of the organisations participating in BSAF Phase II (CRIEPI, the IAE, the IRSN, KAERI, the Paul
Scherrer Institute [PSI] and the SNL) submitted a unit 3 ST for atmospheric transport analysis.
A comparison of these six source terms is shown in Figure 5.3. Three of the STs for unit 3
predicted significant deposition over land (the IRSN, KAERI and the SNL), while the other three(CRIEPI, the IAE and the PSI) predicted low-level deposition over land as well as over the ocean.

1.

Sandia National Laboratories is a multi-mission laboratory managed and operated by National
Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International
Inc., for the US Department of Energy’s National Nuclear Security Administration under contract DENA0003525. SAND2018-8356 R.

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ATMOSPHERIC TRANSPORT AND DISPERSION: RESULTS FOR FUKUSHIMA DAIICHI UNITS 1, 2 AND 3

Figure 5.1. Unit 1 source terms for Cs-137 predicted by five BSAF participants

Figure 5.2. Unit 2 source terms for Cs-137 predicted by six BSAF participants

Figure 5.3. Unit 3 source terms for Cs-137 predicted by six BSAF participants

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ATMOSPHERIC TRANSPORT AND DISPERSION: RESULTS FOR FUKUSHIMA DAIICHI UNITS 1, 2 AND 3

Comparisons for three units combined
Four of the organisations participating in Phase II of the BSAF (the IAE, the IRSN, KAERI and SNL)
submitted a source term for all three units for atmospheric transport analysis. A comparison ofthese four STs is shown in Figure 5.4. Note that the land/ocean border is indicated by the light
blue line that runs roughly north and south through the centre of the isopleth plots. The GRS
used observational data from locations near the Fukushima site to reconstruct a source term.
The observational data were interpreted using a newly developed method for the OECD BSAF
Phase II project (see Chapter 6 of this report for a description of the method). As only land-based
observations were available to the project, the ST calculated by the GRS has been augmented bypublished source term data based on the Japanese WSPEEDI model system for phases whenradioactive releases are dispersed over the ocean. The deposition patterns predicted for thesevere accident calculation STs are depicted in Figure 5.5, along with the calculated GRS source
term (Figure 5.5e). For comparison, Figure 5.5f shows the measured Cs-137 isopleths using thesame distance scale and using similar contour shading to facilitate the comparisons. However,note that the dark red contour level shown for the other results is not used there. The IAE source
term predicted significant deposition over the ocean but very little over land, while the otherthree STs produced significant deposition over land. Each of these has unique characteristics,
but all three have some features in common. These include: significant deposition (>3 MBq/m2)
out to 10-20 km towards the northwest; and a peninsula of deposition southward along thecoastline. The deposition pattern for the GRS source term does not show the distinct peninsulasof deposition to the northwest or south along the seacoast that are observed in the severe
accident calculations.
The Cs deposition to the northwest was observed, but such a high level of onshore
deposition to the south was not. However, it has been noted that the sand along the seacoast to
the south is heavily contaminated, which may indicate that significant deposition occurred
slightly to the east (over the ocean) and resulted in contamination of the sand through wave
and tidal interactions with the seashore (Katata et al., 2015). Thus, a peninsula of deposition to
the south near the seashore is plausible.

Figure 5.4. Combined three-unit source terms for Cs-137
predicted by five BSAF participants

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Figure 5.5. Cs-137 isopleths for the three-unit source terms, the GRS/backwards
calculation source term and observed ground deposition pattern for Cs-137

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A. IAE

B. IRSN

C. KAERI

D. SNLE. GRS/backwards calculation source term

F. Observed ground deposition pattern for Cs-137

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Estimated integral releases
Table 5.1 shows the integral releases of Cs-137 (in PBq) predicted by each of the BSAF contributors
to this chapter and the combined releases for all three units for organisations that evaluated all
the units. In the case of the GRS, individual units were not evaluated; instead, observational data
were used to reconstruct a source term for the combination of the three units. In comparing theseresults to code analyses, the contribution of each unit was roughly estimated assuming that therelease from the first failed unit (unit 1) stops when the release from the second failed unit starts(unit 3) and that this stops when the third unit (unit 2) fails.
Estimated releases from all three units range from 5.5 PBq to 43.2 PBq of Cs-137. Backwards
calculation results are in the order of 13.1-14.5 PBq.

Table 5.1. Integral releases of Cs-137 (PBq)
predicted by BSAF participants
CodeCountry

Unit 1

Unit 2

Unit 3

Combined

CIEMAT

MELCOR 2.1-4803

Spain

5.6

CNL

MELCOR 2.1-6342

Canada8.1

CRIEPI

MAAP 5.01

Japan

6.7

0.8

GRS

Backward calculation

Germany

0.79

10.7

1.6

13.1

IAE

SAMPSON-B 1.4 beta

Japan

4.4

0.0*

1.1

5.5

IRSN

ASTEC V2.0 rev3 p1

France

0.1

0.1

43.1

43.2

JAEA

WSPEEDI

Japan

KAERI

MELCOR 1.8.6

Korea

18.9

5.8

10.9

35.5

NRC/DOE/SNL

MELCOR 2.1-5864

United States

2.8

12.3

10.6

25.7

PSI

MELCOR 2.1-4206

Switzerland

14.5**

0.2* Little release predicted from the reactor building due to modelling deficit.
** See Katata et al., 2015.

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Chapter 6. Radiological measurements and backward
calculations of the source term

Dose rate measurements at the plant site and around
During the accident, the local dose rate was measured by the monitoring points (MPs) locatedat the plant site and around in the Fukushima prefecture (Figure 6.1). The available data arecollected as Gy/h (or Sv/h). Since the measured value depends not only on the release time but
also on the wind direction, meteorological conditions and deposition patterns, these data need
to be analysed by atmospheric dispersion codes in order to confirm the actual release from theplant in Bq/h.
Examples of dose rate measurements performed by the Tokyo Electric Power Company(TEPCO) at some selected MPs are shown in Figure 6.2.
The beginning of the severe core degradation phases in units 1-3 is marked there as well.
Those data have been made available through the Benchmark Study of the Accident at the
Fukushima Daiichi Nuclear Power Station (BSAF) project, Phase II, and have been used forbackwards calculation of the source term as discussed below.

Figure 6.1. Monitoring posts within and aroundthe Fukushima site
A. Within the Fukushima site

B. Around the Fukushima site

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Figure 6.2. Dose rate measurements performed by TEPCO at some selected monitoring
points and start of severe core degradation phases in units 1-3

Dose rate measurements/samplings in the Pacific Ocean
Measurements of radionuclides by samples have been taken in the Pacific Ocean at various
locations and time (Figure 6.3). Those measurements are relevant for several release periods ofthe accident, when the wind was blowing east towards the ocean. In these conditions, eventhough the radioactive plume was released by either venting or building leak, it would beundetectable by the MP or post measurements on land. The measurements of the ocean water
samples were not available within the BSAF Phase II, but have been used in the source term (ST)estimation by the WSPEEDI code (Katata et al., 2015) described below.

Figure 6.3. Sampling points in the Pacific Ocean

Source: Katata et al., 2015.

Backwards calculation of source term by WSPEEDIThe results of a detailed ST estimation based on the coupling of the WSPEEDI-II model to oceanic
dispersion and sophisticated deposition models SEA-GEARN-FDM (Finite difference oceanic
Dispersion Model) are described in Katata et al. (2015) (henceforth referenced as “WSPEEDI” for thesake of brevity). They relate observed features in the measured local dose rate to the activity

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release and selected available plant parameters during the accident progression. The investigation
period covers more than the first month of the accident. The results have been used in Phase II of
the BSAF for comparison against severe accident analyses.
The following brief sketch of the methodology for conducting the WSPEEDI analyses isprovided as follows. Source term estimation was based on the reverse calculation of fission
product (FP) releases which were related to radiological observations in the environment byatmospheric dispersion modelling. The preferred observational parameter measured over land forsource term estimation was air activity concentration. When this parameter was not available, airdose rate measurements combined with samples of deposited nuclides were used. The distanceof the accident site to radiological observations over land served as a basis for ST reconstruction
and ranged from 4 km to over 80 km. The local dose rate measurements of the MPs at the
Fukushima site (cf. MP-1-MP-8 in Figure 6.1a) were not included in the quantification of FP releases.
During periods when released FP was mainly dispersed over the ocean, seawater concentration
measurements in the Pacific Ocean at distances from a few hundred up to several thousand
kilometres were used to correct land-based ST estimations (see Figure 6.3). For the distinction ofdifferent release phases, plant-based information and parameters were used. Active release
phases were related to information on containment venting and hydrogen explosions in thebuildings as well as on RPV and PCV pressure curves in units 1-3. Thus, the identification of active
release phases was not independent from plant-related data assumptions. Between the distinct
active release phases, a continuous release was assumed. The subdivision of these continuous
release phases was mainly based on variations in the available radiological observations and was
not linked to events in the plant. Within each release phase, the reverse calculation method
applied yielded a constant release rate. If necessary, subjective corrections based on “expert
judgement” (e.g. rotations of the modelled FP plumes) were applied to the modelled air
concentration and dose rate fields to reduce discrepancies with the corresponding measurements.
The results of WSPEEDI source term calculations are summarised in Katata et al. (2015).
Release rates until noon on 22 March 2011 are shown in Figure 6.4 as an example.

Figure 6.4. Release rates from reactor scram on 11 March to 22 March 2011
computed by WSPEEDI

Source: Katata et al., 2015.

Backwards calculation of source term by the GRS method
Source term reconstruction at the GRS has been performed for the period from 12 March 2011
00:00 to 26 March 2011 00:00. The reconstruction method aimed at an optimised use of available
radiological measurements at or nearby the Fukushima site (see Figure 6.1). It thus focused on
the evaluation of the numerous local dose rate measurements on site or near the plant, while
the nuclide composition had to be estimated from a limited number of available soil samples.
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In contrast to the WSPEEDI analysis, a deliberately “blind” approach was followed, which omits
the use of any plant information for identification/definition of release phases or quantificationof releases. Measurement data of samplings from the Pacific Ocean at various locations andtimes, as presented in Figure 6.3, could not be used. The reconstruction scheme was based on
the following steps.
Calculation of surface contamination from local dose rate and specific soil activity: For thisstep, measured local dose rates at each of the included MPs were divided into cloud shine causedby airborne radioactivity and ground shine caused by surface contamination. Then, nuclidespecific surface contamination was estimated by relating ground shine to the nuclide compositionof the deposited nuclides, which in turn was determined from available samples of specific soil
activity.
Calculation of air activity concentration from surface contamination and information on
precipitation: Air activity concentration during cloud phases were calculated from the differencein surface contamination before and after the respective cloud phase. The respective deposition
mechanism (dry or wet) was taken into account based on available information on precipitation.
Deposition rates were assumed proportional to the strength of the cloud shine signal. This method
yields estimates for the temporal development of air concentration of aerosols and elementary
iodine, but not for possible contributions from noble gases.
Calculation of radioactive releases from local dose rate and air concentration and modelled
dispersion: These calculations were performed with the Lagrangian dispersion model ARTM
(Atmospheric Radionuclide Transport Model), which was developed by the GRS (Bundesamt fürStrahlenschutz, 2018). Atmospheric transport dispersion parameters and gamma submersion
factors were computed, which can be directly linked to calculated cloud shine at each MP. The
quantity of radionuclides released was then calculated by an appropriate backward calculation
method. For this purpose, an optimal solution for radioactive releases is sought by minimising thedifference between the observed and calculated cloud shine that would result from the release
estimate.
The database for ST reconstruction consists of local dose rate observations, soil activity
concentration samples and meteorological measurements. The suitability of local dose ratemeasurements obtained at the MP depends on the integral length of observation, but also on thedata coverage within the record. A suitable set of on-site MP at Fukushima Daiichi Nuclear Power
Plant exists (see Figure 6.1a). The set of MPs for reconstruction was completed by local dose ratemeasurements in the surroundings, namely at the Fukushima Daini Nuclear Power Plant (MP-4)
and MPs at Kiyohashi, Namie, Yamada, Oono, Yonomori, Shoukan and Yamadoaka (seeFigure 6.1b). In addition, the MP at Kooriyama, Shinzan, Kamihatori, Minamidai, Mukaihata,
Ottozawa, Shigeoka and Futatsunuma were employed for validation of the method.
Samples of specific soil activity are available at the Fukushima Daiichi Nuclear Power Plant
from 21 March 2011. By the time of the first sample, possible releases of short-lived nuclides
with half-lives of a few hours like I-132 would no longer be detectable in the deposits if they had
occurred in the first few days of the accident. However, longer lived nuclides can be comparedin the samples. For this purpose, all samples are decay-corrected to a common reference date
(21 March at 00:00).
Based on this knowledge, I-132 was chosen as representative for short-lived nuclides which
contributed to surface contamination. The amount of I-132 initially to be added to the basic
nuclide composition to suitably explain observed local dose rates for each ground phase wascalculated by a least square fit approach. This inclusion of I-132 in the nuclide vector especiallyenabled the proper consideration of the MPs at the site for ST reconstruction.
The computation of nuclide composition served as the basis for determining air
concentration during cloud phases (Step 2) and performing the backward calculations (Step 3)
for each MP. For Step 3, measurements of wind direction and velocity as well as precipitation
information at the Fukushima Daiichi Nuclear Power Plant were used. Until the afternoon of
16 March 2011, additional weather data were available at the Oono MP, which were combinedwith measurements at the Fukushima Daiichi Nuclear Power Plant for use in the dispersion
calculations.

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Naturally, the observation of radioactive releases at each MP was confined to situations
where the wind disperses the release approximately in the direction of the respective MP.
Likewise, uncertainties in the quantification of the source term grow when the respective MP is
only marginally affected by the dispersing radioactive cloud. These limitations and uncertainties
can be reduced by the combination of backward calculation results for the ensemble of 14 abovementioned MPs. For this purpose, the observed releases at each MP were weighted with themagnitude of dispersion parameters. The weighted average from each observation represents
the common ST estimate based on this ensemble.
As observations over the ocean were not available for the BSAF project for our reconstruction,
source term reconstruction was confined to time phases when releases were dispersed over land.
Nevertheless, an observational coverage of about 50% was obtained for the first phase of theaccident. During the occurrence of the largest measured peaks in local dose rate between theevening of 14 March 2011 and the afternoon of 16 March 2011 (Figure 6.2), observational coverageamounted to about 75%.
Reconstruction results are illustrated for the accumulated release of Cs-137 in Figure 6.5. It
is evident that the weighted average provides a reasonable estimate of the Cs-137 release within
the ensemble and that convergence between minimum and maximum estimates within theensemble increases with time.

Figure 6.5. Accumulated release of Cs-137 reconstructedby the GRS for the first two weeks of the accident

Note: Blue: weighted ensemble average; red: minimum estimate within ensemble of monitoring points; green: maximum estimate
within ensemble; shaded: no observation by monitoring point ensemble possible; dark grey: number of observing monitoring points.

Comparison of the GRS and WSPEEDI source term calculation results
This section gives a comparison between the WSPEEDI source term estimation results and the
GRS source term calculation results. The objective of this comparison is to take advantage ofthe higher temporal resolution and independent identification of active release phases of the
GRS data set. By this, the validity of release phase subdivision within the WSPEEDI data set canbe assessed and insights into possible links between active release phases and available
information about accident progression can be broadened.
A clear advantage of the WSPEEDI method is the coverage of the whole investigation period,
including those periods when the radioactive releases were dispersed over the ocean. On theother hand, through the participation in BSAF, the use of local dose rate measurements on the
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Fukushima Daiichi Nuclear Power Plant site very close to the location of release was possible
for the GRS. This enhanced the temporal resolution of the ST reconstruction and reduced theinfluence of errors and uncertainties on the dispersion calculations.
Figure 6.6 illustrates the calculated temporal development of radioactive releases obtainedby both methods for Cs-137. For the sake of a clear visualisation, the GRS data were aggregatedfrom a higher temporal resolution of ten minutes to hourly values. During most of the periods
covered by both the GRS and WSPEEDI data (white background), the agreement between the
curves is remarkably good, with some differences in the details of temporal development. During
periods which were solely covered by WSPEEDI data (grey background), calculated release rates
show mostly a uniform behaviour, with few variations. This lack of variation is presumably aneffect caused by the lower temporal resolution achievable in these periods rather than by therelease dynamics. This lower temporal distribution is caused by the reconstruction method,which relies on data based on samples from the ocean, which only provide aggregate deposition.

Figure 6.6. Comparison of Cs-137 release rates obtained by the GRS (data aggregatedto hourly values) and WSPEEDI source term reconstruction methods

Note: Shaded: no observation by monitoring points ensemble used for the GRS backwards calculation possible.

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The agreement between the results of the GRS and WSPEEDI calculations is pronounced in
the comparison between accumulated releases of I-131 and Cs-137.

Figure 6.7. Comparison of accumulated releases of Cs-137 and I-131
by the GRS and WSPEEDI source term reconstruction methods

Notes: Blue: Cs-137; red: I-131; GRS (dashed); WSPEEDI (dash-dotted); shaded: no observation by monitoring point ensemble
used for the GRS backwards calculation possible; solid lines: accumulated releases for combined GRS + WSPEEDI ST (GRS data
gaps filled by WSPEEDI data).

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Because of the striking agreement between the GRS and WSPEEDI source term results during
periods covered by both methods, it seems reasonable to combine the ST results in the following
manner. For periods covered by both the GRS and WSPEEDI results, the GRS data were taken as
ST data because of their higher temporal resolution, larger number of nuclides considered andindependent treatment of I-132. For periods not covered by the GRS data (wind direction towards
the ocean), GRS data were completed by WSPEEDI data with respect to releases of I-131, Te-132,
Cs-134 and Cs-137. This procedure combines the advantages of both data sets. The agreement
between this combined “GRS-ST + WSPEEDI-ST” data set and the original WSPEEDI accumulated
releases is nearly perfect, with maximum relative deviations of about 12% for iodine and less than
1% for cesium.
The results of this backwards calculation of the source term have been used for comparisonof severe accident analyses results within the BSAF project and for an improved understanding of
the accident progression. The calculated accumulated release amounts are nearly identical by theafternoon of 16 March 2011. Thereafter, the GRS calculation predicted lower integral releases than
WSPEEDI, due to the increasing fraction of observational gaps in the GRS calculations, time periods
where the wind was blowing towards the ocean. Nevertheless, after the first two weeks of theaccident, the accumulated releases calculated by the GRS summed up to 60% of the WSPEEDI
results for iodine and about 45% for Cs and tellurium (Te) isotopes. Differences in the calculated
release amounts can thus be mainly attributed to the observational gaps in the GRS data.
Because of the striking agreement between the GRS and WSPEEDI source term results
during the periods covered by both methods, it seems reasonable to combine the source term
results in the following manner. For periods covered by both the GRS and WSPEEDI results, theGRS data were taken as source term data because of their higher temporal resolution, larger
number of nuclides considered and independent treatment of I-132, whereas, for periods notcovered by the GRS data (wind direction towards the ocean), the GRS data were completed by
WSPEEDI data with respect to releases of I-131, Te-132, Cs-134 and Cs-137.

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Chapter 7. Lessons learnt and conclusions from the BSAF

The current severe accident analyses have been carried out by various well-known severe accident
codes and the results have been compared against major plant events and measurement datataken during the accident, such as the timing of the hydrogen explosions, reactor pressure vessel(RPV) water level, RPV pressure and primary containment vessel (PCV) pressure (suppression
chamber and dry well). Phase II of the Benchmark Study of the Accident at the Fukushima Daiichi
Nuclear Power Station (BSAF), thanks to the recent findings obtained during plant inspections(mainly muon and PCV internal assessment), allowed uncertainties regarding the current plantstate to be drastically reduced. Comparisons with real evidence highlight some of the limitationsof existing codes and identify areas that could benefit from novel experimental programmes andfuture code improvements.

Review of major technical findings
In the BSAF project, the severe accident analysis codes were used in two ways: 1) a forward
approach, where the analysts allowed the code to predict the accident behaviour according to
the current state of modelling; and 2) a forensic approach, where the analysts forced the code
and/or boundary conditions to comply with observed events. Uncertainties in the measured
data and observed events must be taken into account, too. During the calculations, the analysts
may have needed to force certain events in their simulations in order to synchronise thepredictions with the observed data. This might suggest that specific code models needrefinement or improvement and validation, but it may also simply be a reflection of inherent
natural and stochastic randomness and variability in the exact sequence and timing of key
events in core degradation behaviour.
Core degradation: Major plant measurements, such as RPV and PCV pressure, or radiologicalmeasurements such as containment air measurement system (CAMS), may give indications of
the occurrence of certain events, such as core relocation within the RPV, RPV failure, corium
relocation from the RPV onto the cavity floor and containment failure.
This information can help to reduce uncertainties in predicting the timing and magnitudeof key events. Uncertainties still exist regarding boundary conditions, such as the water
injection during the reactor core isolation cooling (RCIC), high pressure coolant injection (HPCI)and external water addition, despite the development of new models that create more realistic
representations. In addition, the state of the valves and rupture discs for containment venting
system activations also remain a significant source of uncertainty. The magnitude of events
such as pressure peaks are also uncertain. However, findings on how to adjust the simulationin order to reproduce the measurements can inform how core relocation models should be
adjusted to better reflect the observations.
One typical source of scatter in the results of the calculations arises in the late stage of coredegradation with the accumulation of debris on the core plate. In the boiling water reactor (BWR),
this plate has a special design with many openings. It is uncertain if/how debris accumulates on
the core plate or if the debris can independently relocate from each assembly to the lower
plenum, either in between the control rod guide tubes or into it. This could affect the morphologyof the debris in the lower plenum during the quenching period and affect the mode of lower head
failure. Another possible method of corium relocation into the lower plenum is linked to molten
pool formation, radial spreading of the molten pool above the core plate and ablation of the
shroud wall; a scenario often assumed, especially in pressurised water reactors and heavily
reflective of best understanding of the Three Mile Island 2 accident progression. This scenario
represents additional uncertainty for the modelling of corium relocation from the core region.
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The behaviour of control blades during the accident progression is also a major uncertainty.
It is not known to what extent the boron carbide in the control blades was exposed to steam
flow, producing volatile boric acid. The location and condition of residual peripheral structuresin the units may provide some insight on the control rod behaviour.
In addition, modern severe accident (SA) analyses tend to predict early in-vessel melt
progression as a slower process than previously thought two decades ago, driven largely by codesmodelling more heat transfer pathways and accounting for more heat sink availability in the RPV
and containment. In BWRs, the extensive upper internal structures in the RPV have the potential
to melt and add steel to the core debris produced during accident progression. A major uncertainty
related to this process is the heat transfer upward to these structures from the degrading BWR
core. This also affects the heat transfer to gases from a degraded core configuration, which cancontribute to possible failure of the main steam line due to creep; representing one of the existing
uncertainties in the accident progression.
Core-concrete interaction: One striking result obtained in the BSAF activity is the extensive
progression of the molten core concrete interaction (MCCI) predicted by all the codes (except
dedicated MCCI codes such as TOLBIAC) for the unit 1 analyses and to some extent for unit 3,
which continues without termination until 500 hours (the end of the simulation), independentof the water injection onto the debris. This continuous MCCI is inconsistent with internal
inspection of the units, which show that the pedestal walls and structure are still mostly intact(i.e. the structure has not collapsed). Thus, the code predictions of the MCCI erosion seem to beoverestimated and unrealistic, at least with respect to lateral erosion and pedestal wall attack.
Fission product release and transport: Regarding fission products (FPs), there are still several
major uncertainties to be understood, including:As suggested by the reverse FP analyses and some of the computer analyses, releases
from damaged reactors likely persisted in between the readily observable transient burst
releases associated with venting operations or containment failure/leakage events. Some
codes (e.g. MELCOR) model revaporisation of volatile FP species such as CsI and Cs2MoO4 thatare driven off of deposition surfaces by increasing structure temperatures, and other codes(e.g. Thales/Kirche) that consider ongoing evolution of gaseous iodine species such as I2 andCH3I, both of which lead to ongoing low level releases throughout the weeks following theaccident.
While not generally modelled in the BSAF analyses, the transport of aqueous borne FPs thatleak from the suppression pool or the PCV, or through the known underground water infiltration
into and out of the reactor building lower compartments and turbine hall is known to exist, but
not well characterised. Furthermore, leaching of FPs from submerged fuel debris into the PCV
water pools is a persistent source term (ST) that can participate in the leakage of FPs from thePCV from water infiltration as well. These transport pathways could be better modelled in the
SA codes if deemed important for predicting long-term FP releases. Finally, certain unusual FP
forms have been observed, such as Cs silicate particles, where clear understanding of the originof these forms is not clear and improvement in this area of ST characterisation may be useful.
User effects and uncertainty: It was also recognised during the project that a “user effect” is amajor player in obtaining consistent results in the analysis of complicated transients. Variability
in user-selected code input can lead to variability in the computed outputs, and likely reflect theuncertainty in predictive technologies in this area. Additionally, SA codes contain various pointsof occasional numerical instability that can contribute to excessive run-time code execution
failures. The project has shown how scattering of the results from the various participants
decreased from Phase I given the time for the analysts to become more and more experienced
with BWR systems and modelling. The whole BSAF project also allowed participants to shareexperiences of multiple codes and consider how to improve models based on the maturity of othercodes, thus strengthening the capabilities of all of the codes involved.
Fission product transport: Atmospheric transport and dispersion (ATD) of radionuclides fromthe Fukushima Daiichi site played a role in two different BSAF Phase II activities: 1) the estimationof deposition pattern that would have resulted from the predicted STs by the partners; and 2) the
backwards calculation of the ST based on measured data such as local dose rates on-site or near
the plant, samplings taken in the Pacific Ocean, and weather data. The phases when the wind was

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blowing either towards the Pacific Ocean or towards the land are known, and as such, the resultsof the calculated STs can be checked (e.g. the timing of the releases by the ATD calculations andthe resulting overall deposition profiles). Larger uncertainties still have to be considered, as
explained below.
Different weather data sets have been used by the Sandia National Laboratories (SNL) in theestimation of the deposition pattern based on source term data from the BSAF participants,
providing significantly different results. The weather data set that seemed the most appropriate
was selected. For the backwards calculation of the ST, other weather data sets were used by the
GRS and the Japanese organisations using WSPEEDI. In conclusion, it is not entirely obvious
which of the weather data sets best matches the observations.
Deposition patterns in the ATD analyses were only calculated for a single isotope, Cs-137, forthe purposes of comparison to measurements of ground contamination by Cs-134 and Cs-137,
while the backwards calculation provided results for other nuclides. Additionally, release energies
associated with the source term data was accounted for in the analysis using the consideration ofplume lofting. Finally, aerosol size distribution data were considered for the purpose of estimating
deposition as, in some cases, aerosol size distribution can significantly influence deposition
patterns.
It is well known that the results of both the deposition pattern and the backwards calculated
STs were significantly influenced by the weather data set. Further analyses would therefore beneeded to determine why the ST provided by the GRS backwards calculation, which is very similar
to the Japanese WSPEEDI results, does not result in a similar deposition pattern as measured. It is
recommended that future activities include a discussion and possibly a harmonisation of theweather data used, the deposition velocities for dry and wet deposition, and the simulated rainfall
pattern in both activities. Furthermore, uncertainty bands linked to both activities should bedetermined. Nevertheless, these calculations provide important contributions to assess thecalculated STs.
The results in Phase II can be considered to be much more robust than those in Phase I, as
the understanding of the accident and its progression, and the experience and skills of theparticipants improved over the course of the project, and additionally, many plant inspection
results have since been provided by the Tokyo Electric Power Company (TEPCO).
The first advancement from the Phase I efforts regards the information about the plantinternals and current status of the units. Muon inspections of the RPVs and robot inspection of
the PCVs yielded significant information in order to discern between realistic and not realistic
scenarios. The muon imaging investigation showed that very little debris still exists in the coreregion of all the units. It also showed very little debris in the lower head of unit 1, while somehigh-density debris was identified in units 2 and 3. The PCV internal investigations in all theunits revealed ex-vessel debris in all the units. For units 2 and 3, whose status was uncertain in
Phase I, the ex-vessel nature of the accidents became obvious through the camera inspectionsof the PCV, which showed evidence of debris through the videos.
The second advance from Phase I was the identification of the critical parts of the accidents
that, thanks to TEPCO’s support, were stressed throughout the project. In particular, the focus on
the period of core degradation (e.g. three peak periods for unit 2, or the time before the modelledautomatic depressurisation system (ADS) activation in unit 3). This approach stimulated theparticipants to make a particular effort in specific parts of the transient, which were considered
critical for the explanation of the accident and identification of debris distribution.
Finally, it needs to be recognised that the understanding of the nature of a severe accident
evolved throughout the project. Historically, the most studied example of a severe accident is
the Three Mile Island 2 accident. However, this accident was very different from the Fukushima
accidents and the pressurised water reactor (PWR) reactor design involved in Three Mile Island 2
is quite different from the Fukushima Daiichi BWRs, which also included a different accident
for each of the three units. Therefore, participants required time to become familiar with the
Fukushima Daiichi design in particular, because it involved very detailed knowledge of the plant,
which had generally not been addressed in previous studies (e.g. difference between control rod
and control rod guide tube, structures below the lower plenum, details of RCIC/HPCI, details of
valves, etc.).

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Review of major impacts to the BSAF stakeholders
Research programmes such as the BSAF1 and BSAF2 projects, because of the large stakeholder
group size, have the potential to produce impact on the nuclear community at large. Some of
the important impacts of the BSAF studies on the diverse stakeholders are summarised below.
Stakeholders include the international research community of scientists and engineers,
government and regulatory bodies, industry partners including TEPCO, and the general public.
First, for code developers and accident analysts, the BSAF efforts provided a unique
opportunity to apply codes to real-world accident conditions. In the process, the BSAF projects
have resulted in the development of a skilled collective of human knowledge that is now moving
to apply the knowledge gained from BSAF improved code models for melt progression and more
confident application of severe accident codes to both existing plants and emerging nextgeneration plants. This collective will also be extremely useful to the future Fukushima-relatedOECD efforts, as well as ongoing activities related to Fukushima site decommissioning. A review
of some of these accomplishments, gained knowledge and areas for code model improvements
follows.
Impact on code developers and severe accident analysts: The BSAF analytical efforts have led
to greater confidence in the predicted accident progression as well as awareness of the inherent
variability and uncertainty in the code results. It also emphasised the inherent differences
between BWR core melt progression and those in PWRs. The close integration of TEPCO insights
and emerging information about the state of each reactor has allowed identification of potential
code and model improvements in a number of complex areas, such as suppression pool thermal
stratification, material interactions (especially zirconium interactions with fuel and RPV
structures), the mode of vessel head failure and release of core materials to the cavity region, andthe degree of core debris attack on the cavity concrete. These are a few of the focus areas that will
be addressed by code developers and analysts in ongoing investigations of the Fukushima
accidents.
Impact on regulatory practice: The BSAF project has contributed to a good understanding of
the detailed progression of each of the Fukushima accidents and has thereby improved regulatory
confidence of oversight of severe accidents, containment performance and their potential source
terms to the environment. This additionally improves public confidence in that regulatory
oversight. Increased regulatory competency in severe accidents will help ensure that regulations
are as effective as possible and that adequate safety margins exist in the current regulations.
Regulatory considerations of severe accident issues have been improved by the BSAF with respect
to understanding what happened in the accidents, and regulator/public confidence in nuclear
technology and its regulation for safe societal use have also improved. Industry has also beenstrongly affected by the project’s results through improved understanding of what happened in
the accidents and by extension, how accidents can be better managed. Some of the industry
impacts of the project are discussed below.

Impact on industry practice in severe accidents

60•

Severe accident management and guidelines (SAMGs) – The Electric Power ResearchInstitute (EPRI) and the BWR and PWR Owners groups have followed the BSAF programme
and applied the lessons learnt to the refinement and improvement of both the severe
accident guidelines recommended by the community and the education and competencyof the utility industry in the phenomena and impacts of severe accidents on power plants.

•

Vessel breach signatures used in navigating the SAMGs are likely problematic in thatsignificant fractions of debris are likely to remain in the vessel even after initial failureof the lower head.

•

Severe Accident Water Addition (SAWA)/Severe Accident Water Management (SAWM) isa recent SAMG development arising from the understanding of the Fukushima
containment performance and improved containment protection by water addition(SAWA) and in the optimum use or “management” of water (SAWM) in US BWRs. Realworld evidence from Fukushima and the BSAF efforts have provided the technical basis
for these accident management procedures.
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With fuel debris likely clumping in the reactor pedestal initially on steel structures below
the vessel, the impact on a dry well-only injection point requires further evaluation tounderstand the extent to which cooling will reach the ex-vessel debris bed.

•Containment protection and venting: Thinking is evolving regarding the potential forcontainment protection from the over-pressurisation seen in the Fukushima accidents,
where early venting before significant accumulation of fission products in the PCV can bea strategy to prevent radioactive releases later as containment challenges become severe.

•Liner melt-through appears to be a less likely outcome, at least in the timeframe
following lower head failure, based on the analytical findings of many BSAF analyses of
the Fukushima accidents.

•

Risk-informing flex equipment requirements in the long run (long-term maintenance cost):
This is greatly facilitated by increased confidence in the application of severe accident
codes that have successfully captured the essential signatures of the Fukushima accidents.

A special case of industry impacts can be identified for both TEPCO as well as the Japaneseorganisations tasked with decommissioning the Fukushima site. Some additional impacts forJapanese industry follow.

Impact on Fukushima site decommissioning and Japanese industry stakeholders
Collaboration between TEPCO and the BSAF programme has significantly improved TEPCO’s andthe Japanese nuclear industry’s understanding of the accidents and, in time, will improve publicconfidence in nuclear plant safe operations across their fleet. The confidence gained in codeanalysis of these complex accidents also will allow the codes to be used to inform
decommissioning activities where predicted damaged fuel distribution within the RPV and PCV
allow for comprehensive assessment of potential points of entry into the PCV to remove fueldebris, and estimates of material composition of relocated fuel debris allow evaluation of potential
material cutting requirements through the testing of unirradiated material surrogates informedby analysis. Finally, the decommissioning activities will no doubt reveal other phenomena of
potential significance to code modelling capability. Further scientific understanding and code
improvements can be realised by examining and characterising materials recovered from thedamaged reactors – the BSAF foundations provide an informed basis for efficiently and withminimal additional costs identifying areas where scientific understanding can be advanced
through inspections and sample analyses obtained during decommissioning.
These aspects of BSAF support for decommissioning activities have been one of the
programme’s focuses from the onset, and the body of knowledge and technical competency will
greatly improve TEPCO’s success in decommissioning activities that will be ongoing for decades.
Next, we turn to the impacts on public perception.

Impact on public confidence in the nuclear industry
Public confidence regarding the nuclear industry is understandably tenuous, and accidents suchas Three Mile Island 2, Chernobyl and Fukushima present powerful reminders that nuclear
technology must be handled with care and the highest levels of safety. As the existential threats
posed by climate change and global warming put increased pressure on traditional carbon-basedenergy sources, it is ever more critical that the nuclear industry demonstrates understanding andcompetency in the use of nuclear technology for low-carbon energy in the future. The BSAF effort
contributes to this by advancing public knowledge and confidence and dispelling unwarranted
fears. This is a worthy ambition.

Overall conclusions
The BSAF project has been the most significant assessment of severe accident code capabilities
to date against real-world accidents, providing a benchmark of several major integral severe
accident analysis codes. In a way similar to the other benchmark study, Phase I was conducted
as a sort of blind test due to lack of the accident progression information and end state of the

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Fukushima Daiichi accident plant. At the same time as Phase II started, considerable
information inside the RPV and PCV was obtained through direct and indirect investigation.
Therefore, Phase II of the project was conducted with well-organised information provided by
the operating agent with support from the Tokyo Electric Power Company (TEPCO). After twophases and six years of continued analysis, the BSAF project has led to a significant increase in
the state of knowledge of the progression and associated consequences for boiling water reactor(BWR) severe accidents. As information has become available through forensic evaluations andengineering analysis, the state of the art of severe accident modelling practices has beensignificantly improved. This can be seen in the development of new phenomenological models
and code performance enhancements, which enabled Phase II participants to model a full three
weeks following the initiation of the accident, a feat not possible at the beginning of the project.
The project has established a firm basis for future analysis and has also highlighted key
phenomenological gaps in our understanding of severe accidents. These include component(e.g. reactor core isolation cooling – RCIC) behaviour, core damage progression in Zr/SS-richdebris (BWR specific), lower head attack and failure mode, persistence and aggressiveness ofthe molten core concrete interaction (MCCI), and radionuclide revolatilisation. The participatingorganisations look forward to continuing these activities in the coming years and projects.

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REFERENCES

References

Air Resources Laboratory (2017), “Weather data archive”, National Oceanic and Atmospheric
Administration, Silver Spring, MD, ftp://arlftp.arlhq.noaa.gov/archives.
Air Resources Laboratory (2016), “HYSPLIT user’s guide”, National Oceanic and Atmospheric
Administration, Silver Spring, MD, www.arl.noaa.gov/documents/reports/hysplit_user_guide.pdf.
Bundesamt für Strahlenschutz (2018), “Atmospheric Radionuclide Transport Model (ARTM) and
its dose module (DARTM)”, Federal Office for Radiation Protection, Salzgitter, Germany,www.bfs.de/EN/topics/ion/environment/air-soil/emission-monitoring/artm.html.
Draxler, R.R. (1999), “HYSPLIT4 user’s guide”, NOAA Technical Memorandum ERL ARL-230,
National Oceanic and Atmospheric Administration, Silver Spring, MD, www.arl.noaa.gov/
wp_arl/wp-content/uploads/documents/reports/arl-230.pdf.
Draxler, R.R. and G.D. Hess (1998), “An overview of the HYSPLIT_4 modeling system oftrajectories, dispersion, and deposition”, Australian Meteorological Magazine, Vol. 47/4,
pp. 295-308, www.arl.noaa.gov/documents/reports/MetMag.pdf.
Katata, G. et al. (2015), “Detailed source term estimation of the atmospheric release for the
Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an
atmospheric dispersion model with an improved deposition scheme and oceanic dispersion
model”, Atmospheric Chemistry and Physics, Vol. 15/2, pp. 1029-1070, https://doi.org/10.5194/
acp-15-1029-2015.
Richter, C., H. Thielen and M. Sogalla (2015), Ergänzende Untersuchungen zur Validierung desAtmosphärischen Radionuklid-Transport-Modells (ARTM), GRS-394, GRS, www.grs.de/sites/
default/files/pdf/grs-394.pdf.
Sogalla, M. et al. (2019), “Reconstruction of the Fukushima source term from nearby radiologicalmeasurements”, paper provided for NURETH-18, Portland, OR.
TEPCO (2017a), “Evaluation of the situation of cores and containment vessels of Fukushima
Daiichi Nuclear Power Station units 1 to 3 and examination into unsolved issues in theaccident progression”, Progress Report No. 5, Attachment 3-8, “Leaks from the Unit-3 PCV
and steam release in a large amount”, Tokyo Electric Power Company,
www.tepco.co.jp/en/press/corp-com/release/betu17_e/images/171225e0240.pdf.
TEPCO (2017b), “Locating fuel debris inside the Unit 3 reactor using a muon measurement
technology at Fukushima Daiichi Nuclear Power Station”, Tokyo Electric Power Company,
www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170928_01-e.pdf.
TEPCO (2014), “Results of water level measuring inside the suppression chamber (S/C) at theUnit 2 in Fukushima Daiichi NPS”, Tokyo Electric Power Company, www.tepco.co.jp/
en/nu/fukushima-np/handouts/2014/images/handouts_140121_04-e.pdf.
TEPCO (2013), “Results of investigation around lower parts of Unit 1 vent pipes at Fukushima
Daiichi NPS (second day)”, Tokyo Electric Power Company, www.tepco.co.jp/en/nu/
fukushima-np/handouts/2013/images/handouts_131114_05-e.pdf.
Uchida, S. et al. (2021), “An approach toward evaluation of long-term fission productdistributions in the Fukushima Daiichi Nuclear Power Plant after the severe accident”,
Nuclear Engineering and Design, Elsevier, https://doi.org/10.1016/j.nucengdes.2021.111256.

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Benchmark Study of the Accident
at the Fukushima Daiichi Nuclear
Power Plant

Understanding the accident at the Fukushima Daiichi Nuclear Power Plant is important for safe and timely
decommissioning of the reactors. This objective, together with the development of better computer codes
for analysis of severe accidents, was the aim of the benchmark study conducted under the auspices of theOECD Nuclear Energy Agency. Through the diversity of the modelling codes and approaches, and the useof parametric studies, it has been possible to identify the more likely scenarios that can fit with the limited
data available from the accident. The insights gained from the project will help guide research into severe
accident behaviour, improve severe accident computer codes, develop accident mitigation and response
at nuclear power plants, support regulatory oversight related to severe accidents, and inform policies on
the development and deployment of nuclear technology.

Nuclear Energy Agency (NEA)
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NEA No. 7525