Preliminary forensic engineering study on aggravation of radioactive releases during the Fukushima Daiichi accident

In this unit its blowout panel, provided for over-pressure protection against a main steam pipe breach in the secondary confinement building, was inadvertently activated before a leakage of radioactive effluent from the PCV.

Nuclear Engineering and Design 324 (2017) 315–336 Contents lists available at ScienceDirect Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes Preliminary forensic engineering study on aggravation of radioactive releases during the Fukushima Daiichi accident MARK Genn Saji Independent Research and Consultant, Ex-Secretariat of Nuclear Safety Commission (retired), Japan A B S T R A C T Even after years since the Fukushima accident, the exact accident progression for each unit and location of core debris have not been clarified, although solidified low-melting metal debris was identified at the bottom of the 1F3 reactor pressure vessel (RPV) in July 2017 Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside the Primary Containment Vessels (PCV) In spite of their effort, the observed environmental temperature distribution data does not support the existence of a significant radiation heat sources attributable to the molten core at the bottom of the PCV At this point the total decay heat of the core debris should be as large as a few hundred kilowatts after 2000 days since the initiation of the accident as summarized in Annex A In addition, the temperature of the accumulated water inside PCVs is 10–20 °C even with a reduction of the water injection to 3.0 m3/h The RPVs appear to be holding the heating core debris with water, implying that “in-vessel retention” of core debris has been achieved thanks to effective accident management This should help greatly in the retrieval of the core debris by removing the top head of RPV Under these circumstances the author has conducted a forensic engineering study (i.e., different fields of science working together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident Through this study the author identified that a large portion of the land contamination observed at the north-west direction is mostly the result of the accident that occurred at Unit In this unit its blowout panel, provided for over-pressure protection against a main steam pipe breach in the secondary confinement building, was inadvertently activated before a leakage of radioactive effluent from the PCV Its activation is believed due to the hydrogen explosion in Unit which occurred on March 12, next day of the Fukushima accident initiation By losing the confinement function, the radioactive effluent leaked from the 1F2’s PCV and would have been discharged without mitigation This accident scenario explains the series of leakage events identified in two of the 24 monitoring posts which had been installed by the Fukushima Prefectural Government A series of six large releases were repeated between March 15 and 16, behaving like a periodical actuation of the safety valves for the PCV Such multiple release events were very likely induced by the overpressure release of the PCV due to leakage of the dry well flange This leakage should have been induced through discharging steam and hydrogen due to the activation of Safety and Release Valves (SRV) into the suppression pool (SP) water Unfortunately the wet well atmosphere must have been that of air, since there was no nitrogen charge line to the atmosphere of the SP surface water The resultant air-hydrogen mixture resulted in an “internal hydrogen explosion” which should have deformed the flange The recent robotic inspection inside PCV revealed that a gigantic water splash appears to have occurred at the bottom of the PVC dislodging the gratings installed over the platform Next to the series of large releases from Unit 2, Unit also induced two large releases on March 12 Fortunately, these releases left more than orders of magnitude less soil contamination compared with the series of releases from Unit Unit also released a significant amount of radioactive species twice on March 13, resulting in a very small soil contamination The main constituent of the radioactivity is likely radioactive noble gases in this unit E-mail address: sajig@bd5.so-net.ne http://dx.doi.org/10.1016/j.nucengdes.2017.08.002 Received 28 February 2017; Received in revised form 31 July 2017; Accepted August 2017 Available online 21 September 2017 0029-5493/ © 2017 The Author Published by Elsevier B.V This is an open access article under the CC BY license (http://creativecommons.org/licenses/BY/4.0/) Nuclear Engineering and Design 324 (2017) 315–336 G Saji Nomenclature R/B RCIC RHRS RPV SBO S/C S/P SGTS SRV SPDS NPP PCV TEPCO T/H W/W 1F1∼1F4 Fukushima Daiichi Unit 1∼4 BFL basement-floor level DBA/E design basis accident/event DID defense-in-depth D/W drywell ECCS emergency core cooling system FDA Fukushima Daiichi accident HPCI high pressure coolant injection I&C Instrumentation and Control LOCA loss of coolant accident LPCI low pressure coolant injection LUHS loss of ultimate heat sink PCV Primary Containment Vessel reactor building reactor core isolation cooling system residual heat removal system reactor pressure vessel station blackout (loss of all AC power) suppression chamber suppression pool standby gas treatment system Safety/Release Valve safety parameter display system nuclear power plant Primary Containment Vessel Tokyo electric power company Turbine Hall wet well (suppression pool) (LUHS)” and further aggravated through the “loss of I & C (Instrumentation and Control) power.” By losing all safety provisions to control the troubled reactors, a series of environmental release events followed during the active phase of the accident which took place during the course of the first week Unfortunately, the amount of radioactive species and timing of the large environmental releases are still not known, since all of the environmental monitoring stations surrounding the site boundary were wiped out due to the “loss of I & C power” An overall accident scenario is illustrated in Fig in the eventtree formalism Introduction 1.1 Global accident sequence The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in operating history for a commercial nuclear power plant (Nuclear Emergency Response Headquarters, June and September 2011: National Diet of Japan, 2012; (National Government's) Investigation Committee, 2012; TEPCO's Investigation Report, 2012; Atomic Energy Society of Japan, 2015; RJIF, 2014) The author has also published Post Accident Safety Analysis Report of the Fukushima Accident – Future Direction of Evacuation: Lessons Learned (Saji, 2013) The tsunami, which arrived at 15:37 on March 11, 2011 brought the plants into an unprecedented severe accident status of prolonged SBO (NUREG/CR-5850, 1994) combined with the “loss of ultimate heat sink 1.2 Breakdown of fundamental safety approach – Significance of the Fukushima severe accident Assuming a LOCA as its DBA following the well-accepted hypothesis from the 1970s, the fundamental approach for safety assurance of the Fig Simplified accident sequence for Fukushima Daiichi Unit to Unit Note: The disaster was triggered by the gigantic earthquake which induced the loss of offsite power Although DGs started as designed, they all failed 51 later due to the tsunami, which submerged seawater pumps disrupting the discharge of residual heat to the ocean and induced LUHS Resultant degradation of core cooling and loss of the hydrogen removal function induced a hydrogen explosion which devastated the reactor buildings 316 Nuclear Engineering and Design 324 (2017) 315–336 G Saji determine progressive causes of the event Since many law suits are being raised after the Fukushima Daiichi Accident (FDA) against TEPCO, the Government Investigation Committee reports should have considered such forensic engineering investigation Instead, the testimony collected during such investigation is not being disclosed in order to weigh more on an investigation of the truth of the accident details For example, nine months after the accident, the Fukushima Nuclear Accident Independent Investigation Commission was established by a unanimous resolution of both the House of Representatives and the House of Councilors of the National Diet, which represent the people of Japan (National Diet of Japan, 2012) In the preface of the final report, Dr K Kurokawa explains the objectives of the committee which includes the following statement “To investigate what was at the center of this accident, we could not but touch upon the root of the problems of the former regulators and their relationship structure with the operators.” In line with the committee’s indication the new regulatory framework (i.e., Nuclear Regulatory Agency (NRA)) was established by abolishing the previous Nuclear Safety Commission (NSC) Note that the forensic engineering studies are not in the basis for the investigation committee’s survey However Dr J.O Henrie’s approach in investigating the TMI accident is impossible to apply in the study on FDA, since the process data are not available unlike in the case of TMI accident In addition, Dr Henrie focused on H2 generated by the reaction of zirconium with water, by stating that “H2 generated by radiolysis was probably insignificant” The author has theoretically investigated the root cause of hydrogen generation during the FDA and found that the “radiationinduced electrolysis” is more likely than radiolysis (Saji, 2016) It clarifies that the hydrogen generation may not have been from the high temperature zirconium– steam reaction A short overview of the hydrogen generation mechanisms is summarized in Annex B Rather it is more likely due to “radiation-induced electrolysis” occurring with a “different radiation cell” configuration (Saji, 2017) With this mechanism, a large amount of hydrogen is generated before the loss of the reactor water level It also explains the root cause of the hydrogen explosion that occurred in Unit 4, where all of the fuel assemblies in the reactor core were evacuated to the spent fuel pool for special maintenance2 at the time of the accident Since there is a high possibility that hydrogen generation may neither be through the zirconium-steam reaction nor radiolysis, more robust evidence of the footprints for the accident is necessary In view of this, diverse evidence were collected and integrated in this forensic engineering study, although the author does not intend to go into the legal implications Typical data included: Fukushima Daiichi has been deployed In this approach, the DBA-LOCA should envelope a spectrum of accidents induced by the malfunction of equipment and human errors This practice was imported by referring to the many plants with a firm construction and performance record developed in the Mid- to Eastern US where the seismic events are not dominant “Design Basis Events (DBE) are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions.“ (NRC 50.49) Historically “defense-in-depth (DID)” is the basic approach for prevention of the occurrence of DBEs, as well as for protection from the further evolution of the events and mitigation of their consequences The concept of the DID has been fortified by incorporating lessons learned from accidents and operational experiences One of the respected textbooks on DID is the IAEA’s INSAG-10 which classifies five levels of defense as extracted below (IAEA, 1996): Level 1: Prevention of abnormal operation and failures Level 2: Control of abnormal operation and detection of failures Level 3: Control of accidents within the design basis Level 4: Control of severe conditions including prevention of accident progression and mitigation of the consequences of a severe accident Level 5: Mitigation of the radiological consequences of significant external releases of radioactive materials In referring to this ranking, the first layers related to the DBE were wiped out upon the tsunami’s arrival Nevertheless there should have been some means to control severe accidents and/or to mitigate their consequences in relation to Level of DID This is the main objective of this report It is to provide feedback as to why the TEPCO’s commendable accident management effort could not sufficiently prevent gross release of radioactive species to the environment This induced prolonged off-site counter measures (Level 5) resulting in prolonged evacuation of the general public1 and more than 2000 “disaster-related premature early deaths” due to stresses and deterioration of Quality of Life of the evacuees, although there were no radiological health death toll Due to the difficulties of predicting natural phenomena, occurrence frequency and prediction of consequences from earthquakes and tsunamis, the Fukushima Daiichi did not specify a well defined design base for tsunamis Instead, TEPCO referred to historical experiences for tsunamis which indicated OP + 5.7 as the maximum tsunamis, although they raised the level to 6.1 m on 2006 The ground level of the site was set at 10 m above average sea level (OP Onahama Pail) However, the OP + 13.1 m tsunami ran up the site and flooded the D/G and electrical panels, resulted in unprecedented prolonged SBO This indicates the difficulties in predicting the natural phenomena and protecting the site against such phenomena The author published a new approach for incorporation of risk for earthquakes and tsunamis (Saji, 2014) • Water samples acquired from the T/H Detailed characteristic data • • • 1.3 Forensic engineering studies on severe accidents Recently forensic engineering practices are being compiled especially in the field of civil engineering (Terwel et al., 2012) Formally forensic engineering also involves testimony on the findings of these investigations before a court of law or other judicial forum, when required A similar approach for the evaluation of the 1979 Three Mile Island Unit accident was performed by James O Henrie without legal implication (Henrie, 1989) Unlike a computer simulation of an event, forensic engineering is the evaluation of the recorded data and damages as well as examining the surviving components after an event to • • obtained by TEPCO and in which chemical and radiological analyses were performed by JAEA soon after the accident Independent radiation monitoring data obtained around the periphery regions of the Daiichi site by the Fukushima Municipal Government Remote aerial 131I measurement data taken by USDOE and recovered by JAEA in the soil contamination data Chronological data compiled by TEPCO from March 11–16 The data contains detailed operator actions, local indications of instrumentation and observations of the actual state of components TEPCO’s robotic inspection data soon after the accident as well as the recent robotic inspection data taken inside the PCVs and 1F1 torus room Identification of “core debris” inside the Unit PCV by introducing a submarine type robot in July 2017 Replacement of its reactor core shroud which is a stainless steel cylinder surrounding a nuclear reactor core It helps by directing this cool water towards the reactor core, providing stability to the nuclear reactions Even after yeas there remain as many as 128,000 evacuees in Fukushima 317 Nuclear Engineering and Design 324 (2017) 315–336 G Saji • Muon imaging data taken and analyzed jointly by several research institutes • Hamaoka Unit accident analysis reports • The author’s study on “radiation-induced electrolysis” (Saji, 2016, RPVs, the injected water should leak down to the PCV Since the flow rate of the injected water is still as much as m3/h (TEPCO, 2017b), the PCVs should have been filled with the injected water However, such a difficulty has not been experienced In addition, the temperature of the accumulated water in the PCV that is at an ambient temperature This observation is not consistent with the melt-through scenario, since the decay heat, estimated by TEPCO is 70/90/90 kW for Units 1–3, respectively (TEPCO, 2017b) Their water temperatures are around 17 °C whereas the temperature at the bottom of the RPV is 21.5/27.3/24.8 for Units 1–3 respectively More recently, starting July 19, 2017, TEPCO released a series of results of investigation of Unit PCV by employing a newly developed submarine type (ROV) robot (TEPCO, 2017c) The most remarkable finding is that they finally identified “core debris” drooping down from the bottom head of RPV like a stalactite and then piled up inside of the accumulated water at the bottom of PCV In their photos it is remarkable to observe that there is no steaming, although the estimated total decay heat is as large as 127 kW in 1F3 after 2000 days since the reactor trip If the debris contains some fraction of the core materials, steam bubble should be visible This observation suggests that the currently observed debris should consist of solidified molten metals More detailed discussion is included in Section 5.3 Under this situation, what is needed is an unfolding method to identify the root cause of the accident based on the footprints left by the accident This is the basic motivation for writing this paper 2017) Unique features of the Fukushima Daiichi accident 2.1 Fukushima disaster: un-experienced type of severe accident The Fukushima accident is an un-experienced type of severe accident with the loss of the ultimate heat sink (LUHS) and prolonged SBO which was further aggravated by the loss of I & C power With these unexperienced accident initiators, the author believes that attention should not be focused mainly on the current core status based on the TMI experience Rather it is essential to investigate whether the various safety provisions, notably containment vessels and ECCS, which are designed based on DBA-LOCA, have some residual safety functions even under LUHS and prolonged SBO This accident is very different from both TMI (small LOCA aggravated by inappropriate operator actions) and Chernobyl (reactivity accident due to intrinsically unsafe graphite moderated light water cooled reactor configuration with a positive void coefficient) The extrapolation of the knowledge on the degraded core status obtained from these earlier severe accidents may not be applicable Currently a “melt-down” scenario has been hypothesized and is practically the consensus in Japan Although it has not been internationally defined as such by IAEA, this appears to mean that the reactor fuels have molten together with in-core structural materials forming a volume of “corium.” Due to high decay heat during the active phase of the severe accident (as long as days in Daiichi), the corium is considered to have melted through the lower head of the RPVs and relocated at the bottom of PCVs Based on this hypothesis, most of TEPCO’s decommission activities have been directed towards identifying the location and properties of corium, which should be a lava-like mixture of nuclear fuel and other structural materials first observed after the Chernobyl accident In spite of their effort, TEPCO has not been able to confirm its existence at the bottom of the PCVs, where the corium should have melted through the lower heads of the RPVs The “melt-down” scenario was developed through mechanistic analyses by running severe accident analysis codes (e.g., MELCORE, MAAP and SAMPSON) Unfortunately no data from the SPDS (safety parameter display system) was available to verify the code predictions during the active phase of the FDA The SPDS was developed based on the TMI experience and installed in most NPPs around the world It was also installed at the Fukushima Daiichi, however it was not available for accident management and little data was recorded due to the loss of the I & C’s power and failure of its local detectors (Toshiba, 2012) Nevertheless, since these codes predicted a series of melt through of the lower reactor vessel heads, current efforts are directed towards robotic inspection of the bottom of the heads with remote cameras as well as dose and temperature measurements of the environment inside of the PCVs In spite of their effort, the observed data does not support the existence of the large radiation heat sources attributable to corium In addition recent cosmic ray muon imaging results from 1F2 concluded that most of the core debris should still be retained in the lower head as well as on the Reactor Support Plate These results were released on July 26, 2016 by a team from TEPCO, IRID, KEK, Tsukuba University and Tokyo Metropolitan University (TEPCO, 2016) The results imply that the mechanistic analyses applying the severe accident analysis codes may not be reliable without plant data when performing the reverse (unfolding) analysis However, the “meltdown” hypothesis should be reconsidered in view of the recent robotic inspection results inside of the PCVs (TEPCO, 2017a) If the core debris has molten through the bottom head of the 2.2 Unit-specific accident progression During the Fukushima Daiichi accident, which occurred at 15:37 on March 11, 2011, each unit revealed its own peculiar accident progression as summarized below • 1F1: • • • Very early leakage of radioactive species into the reactor building, as early as 21:51 on March 11, followed by the initial hydrogen explosion, which occurred the next day, March 12 This motivated many analysts to assume the very early onset of a zirconium-hydrogen reaction 1F2: the largest leakage of highly contaminated water into the basement of the Turbine Hall (T/H) This unit is likely responsible for the severe soil contamination with radioactive cesium deposited in a northwest direction from the plant It also contaminated the near field with radioactive iodine to the south of the plant 1F3: the most severe hydrogen explosion among these units: however, there is no trace of radioactive contamination attributable to this unit Nevertheless un-heating solidified debris has been identified recently 1F4: hydrogen explosion occurred even though its reactor core had been evacuated to the spent fuel pit No serious radioactive releases Even after years since the accident, the exact accident scenarios that explain these morphologies remain unknown 2.3 GE’s assessment on performance of Mark I containments at Fukushima Daiichi Fukushima Daiichi Units 1–4 are BWRs were constructed by importingthe technical bases of Mark I containments (GE Report, 03/19/ 2011) The following statements are quoted from GE’s summary page, although the author is not necessarily in agreement with their entire statement “Early reports regarding Units 1–3 stated plant operators used safety relief valves to relieve pressure in the reactor pressure vessel In addition, when the fuel rods became uncovered, hydrogen formed in the core (due to zirconium/water reaction) and was also transported into the wet well when the reactor vessel safety relief valves opened The combination of steam and hydrogen flowing into the wet well 318 Nuclear Engineering and Design 324 (2017) 315–336 G Saji increased the temperature and pressure Since there was no on- or off-site power available, there was no means of cooling the wet well water Over time, the pressure in the primary containment rose over the design pressure To avoid a containment breach, venting became necessary Upon venting it is believed that vented hydrogen gas caused the explosions at these units.” Table Results of nuclide analysis by JAEA (Bq/ml) Sample Sampling date: mm.dd Start of Counting: mm.dd Duration of Counting (s) I-131(Bq/ 8.04d ml) Cs-134 2.06a Cs-137 30.0d Ba-140 12.7d La-140 40.3 h Sr-89 50.5d Sr-90 29.12a GE noted the following points: • Concurrent long-term loss of both on-site and off-site power for an • • extended period of time is a beyond-design-basis event for the primary containment on any operating nuclear power plant The Mark I containment vessels appeared to have held pressure to well above the design pressure The response of the reactor pressure vessel and reactor in general agree with the severe accident management studies performed in the 1980s and early 1990s 1F1 T/B BFL Accum Water 2011.3.24.9.40 2011.4.13.12.50 1F2 T/B BFL At S Stairway 2011.3.27.20.40 2011.4.13.18.34 1F3 T/B BFL Leaked in Water 2011.3.24.21.00 2011.4.14.21.00 2000 2000 2000 3.00E+04 2.00E+06 1.60E+05 1.20+05 1.60+05 560 300 5.70E+01 2.10E+01 2.60E+06 2.80E+06 2.40E+05 2.20E+05 7.00E+05 1.40E+05 1.40E+05 1.60E+05 1.50E+04 1.70E+04 8.60E+04 1.50E+04 containing radioactive species, accumulated in the basements of the R/ B as well as the T/H The injected water carried the radioactive species and leaked out of the RPV into the PCV, until finally accumulating into the basements of T/Hs The continued water injection gradually raised the water level, on top of the tsunami water, both of which flooded the basements of these buildings When the accumulated water purification systems were constructed and began operating around June 17, a portion of the water volume as well as radioactive species were removed There is a further complexity to the volume of accumulated water, due to an unidentified amount from an underground water source which seeped into the original amount In spite of these uncertainties, it should be possible to estimate the leakage of radioactive species from the PCV by knowing the concentration of radioactive species and total volume of the accumulated water on the same sampling date The amounts of release not accounted for are: (1) direct release from the R/B to the environment, such as through the hydrogen explosions, venting events through the operator’s actions, as well as several “spontaneous venting” events; (2) an amount of radioactivity deposited inside of the R/B but not in the accumulated water in the basement of the T/H; (3) sludge of the insoluble species (e.g., refractory species, such as Zr, Mo, Ce, Np, Pu, Cu, U and intermediate species such as Sr, Ru, Ba) likely deposited at the bottom of the basement On May 22, 2011, TEPCO released “Results of Analysis of Accumulated Water in the Turbine Building (JAEA)” (TEPCO, 2011a) as summarized in Table The sampling was made under a very high radiation field, with contact dose rates of nearly Sv/h Although it may not be representative with just one set of samplings, the data are very precious In principle, by knowing the total volume of accumulated water, the total amount of radioactive species leaked in this pathway can be estimated Unfortunately, the levels of the accumulated water only became available months after sampling For this reason, the total amount of accumulated water was assumed to be close to the estimated total amount of injected water up to the sampling dates as shown in Table By multiplying the values in Tables and 2, and adjusting the results to the shutdown activity on March 11, the total radioactive inventory in Bq are summarized in Table This table is converted to a fraction of radioactive species with respect to the shutdown inventory in Table The results indicate that the release fraction (i.e., the ratio of released/shutdown inventory of radioactive species) into the However, a similar statement of “a beyond-design-basis event” made by TEPCO was thoroughly criticized by the Japanese public and has not exonerated them from blame for the accident The author intends to clarify why the radiation exposure to the public could not have been evaded, in spite of TEPCO’s commendable accident management activities under very restrictive situations 2.4 Bypass leak in Unit through drywell flange leakage GE’s early assessment as quoted above indicates that the Mark I containment vessels appeared to have held pressure to well above the design pressure However, the environmental monitoring data which will be discussed in Chapter indicates that Unit 2’s dry well flange leaked several times on March 15–16, behaving like a mechanical safety valve resulting in a series of puff releases (i.e., several short duration releases due to overpressure in the PCV) which were not filtered nor confined for removal of aerosol through precipitation on the walls Such a release during the severe accident was preventable as stated in the IAEA’s Safety Guide No NS-G-1.10 (IAEA, 2004) The following basic requirement is quoted from Section 6.4 of IAEA’s Safety Guide “For existing plants, the phenomena relating to possible severe accidents and their consequences should be carefully analyzed to identify design margins and measures for accident management that can be carried out to prevent or mitigate the consequences of severe accidents.” Additionally, the blowout panel of the R/B was inadvertently opened which was said to be due to the hydrogen explosion in nearby Unit 1, which occurred on March 12 With the large hole in the R/B, it was unable to serve as a secondary confinement function resulting in a bypass leakage directly from the reactor vessel Note that the SGTS, which was supposed to filter the effluent leakage before discharging from the stack, was not available due to SBO With an unfiltered bypass leakage, the main cause of the environmental contamination to the NWN direction from the plant induced three orders of magnitude more severe land contaminations brought on by the events at Units and Apparently the dry well flange did not have the necessary safety margins for the FDA Amount of radioactive releases 3.1 Radioactivity leaked into the basement of T/H The forward analysis to predict the amount released to the environment is extremely difficult, however an indirect estimation can be made from the sampling data from the accumulated water as explained in this section The pure-water/seawater injection was initiated during the early phase of the accident and continued during the removal of decay heat Since the injected amount exceeded the necessary amount for decay heat removal with evaporation, a large portion of the water, Table Estimated volume of accumulated water Location Total injection (m ) 319 1F1 T/H BFL 1F2 T/H BFL 1F3 T/H BFL 2633 9824 4225 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Table Total amount of radioactive species in accumulated water Table Cs-137 deposition and areas Species Half Life 1F1 T/H BFL 1F2 T/H BFL 1F3 T/H BFL Zone Cs-137 deposition Area (km2) Cs-137(PBq) I-131 Cs-134 Cs-137 Ba-140 La-140 Sr-89 Sr-90 8.04d 2.06a 30.0d 12.7d 40.3 h 50.5d 29.12a 1.36E+15 3.16E+14 4.21E+14 8.93E+12 6.52E+17 2.36E+11 5.53E+10 3.38E+17 2.55E+16 2.75E+16 1.43E+16 1.78E+21 1.08E+16 3.69E+14 1.16E+16 5.92E+14 6.76E+14 3.84E+14 5.93E+19 5.72E+14 3.95E+13 VIII VII VI V IV III II I Total > 3E+06 (Bq/m2) 3E+06 – 1E+06 1E+06 – 6E+05 6E+05 – 3E+05 3E+05 – 1E+05 1E+05 – 6E+04 6E+04 – 3E+04 3E+04 – 1E+04 < 1E+04 1E+04 – 6E+06 76 304 208 292 2048 769 2101 6862 NA 12,660 0.34 0.61 0.17 1.31 0.41 0.06 0.10 0.14 NA 3.13 Table Ratio of released/shutdown inventories Species Half Life 1F1 T/H BFL 1F2 T/H BFL 1F3 T/H BFL I-131 Cs-134 Cs-137 Ba-140 La-140 Sr-89 Sr-90 8.04d 2.06a 30.0d 12.7d 40.3 h 50.5d 29.12a 1.02E−03 3.57E−04 3.76E−03 3.41E−06 NA 1.38E−07 5.57E−07 1.47E−01 1.67E−02 1.42E−01 3.16E−03 NA 3.67E−03 2.15E−03 5.07E−03 3.88E−04 3.49E−03 8.50E−05 NA 1.94E−04 2.31E−04 Daiichi to the Emergency Response Center to assist in performing quick dispersion calculations by SPEEDI for accident management However, due to the SBO’s creating further adverse conditions which were induced by the “loss of I & C power”, neither the environmental radiation monitoring nor plant data from process computers were available Due to this lack of info related to the event, one of the most precious data was obtained via remote monitoring by the US DOE/NNEA, where the first survey data was disclosed on March 22 (US DOE/NNEA, 2011) More detailed data were collected by fixed wing and helicopter survey flights at altitudes ranging from 150 to 700 meters The cesium deposition was determined from aerial and ground-based measurements Using US technology, the Ministry of Culture, Education and Sports (MEXT) produced wider land contamination maps, which were verified with detailed in-situ soil measurements Table has been updated by using the new set of data for the integration of the total amount of 137Cs deposited on the land The results show approximately a factor of 0.4 less than the author’s previous results (Saji, 2013) The difference was traced down to the dose range of Zone VIII (previously Zone V), where the upper bound was reduced by an order of magnitude Due to the uncertainties in the remote monitoring soon after the accident with the strong gamma spectra from radioactive iodine isotopes, it is likely that the DOE assumed a large safety margin in the most highly contaminated zone The in-situ soil sampling data now revealed that the highest dose rate in the most contaminated region was 3000 kBq/m2 In updating Table 5, the recently obtained wider (> 100 km from the Fukushima Daiichi) land contamination map3 was used to include above 10 kBq/m2 zones, although some decrease in dose rate due to weathering (approximately 40%/y) is anticipated Because of this concern, the earlier August 30, 2012 data4 were used for the region within 100 km With detailed land contamination density data now available, it was a straightforward process to estimate the total amount of the cesium simply by calculating the areas of each zone The estimated gross land contamination of approximately 3PBq in 137Cs indicates that the environmental release is much less than 75–86 PB (1996 estimation) reported for the Chernobyl accident (UNSCEAR, 2000) The estimated environmental release fraction of 137Cs, which is the ratio of the environmental releases to the shutdown core inventory, amounts to only 1.5% of the1F1 core inventory or 1.2% for 1F2 and 1F3 The best estimate for the environmental release is 1–2% of the shutdown core inventory of 137Cs, considering uncertainties in the land contamination density maps This value should not be confused with a fuel failure rate since the containment system had a large decontamination factor The actual fuel failure can be an order or so larger in the reactor vessel These values indicate the following land contamination characteristics: accumulated water ranges from 14% for 1F2 to a fraction of a percent for “volatile” species (i.e., I and Cs) in 1F1 and 1F3 Those for “intermediate” species (i.e., Ba, La, and Sr) are even smaller in magnitude This suggest that the major pathway of radioactive release from the fuel cladding should be from a puncture in the fuel cladding due to high temperature ballooning failure The ballooning failure occurs through the softening of fuel cladding at a high temperature combined with overpressure induced with gaseous and volatile species The large release fraction for 1F2 indicates that the release passage of 1F2 may be different from both 1F1 and 1F3 The difference suggests an involvement of the S/C where the steam, containing the radioactive species, is discharged from the RPV into the S/C pools The radioactive species are decontaminated in the pool water by a factor of two orders of magnitude of reduction In addition, the large explosive sound occurred near the S/C between 6:00∼6:10 on March 15 may indicate another “internal hydrogen explosion” event in the S/C, which resulted in the leaking of the highly contaminated suppression pool water into the R/B through degradation of the flange of PCV Further discussion will be made in Section 5.2 Also, the behavior of the release fraction of Cs-134 and Cs-137 are strange since the former depends strongly on the burn-up through a fuel management sequence whereas the latter is relatively independent In Fukushima Daiichi, the reactor cores contain fuels with a variety of fuel management histories, some of which date back to more than 10 years There is a high possibility that radioactive releases are more from the aged fuel assemblies, many of which may have punctured due to their higher internal fission gas pressure It is necessary to calculate the activation data of each assembly, instead of a core average, when the status of an individual fuel assembly would be confirmed years from now The release fractions indicate that a large portion of volatile species (as well as rare gases) have been released from the core into the reactor water at least in 1F2 Although the estimation of radioactive species into the accumulated water may become a source term for the marine environment release, it does not provide any source term information for the atmospheric releases that produced widespread land contamination around the Fukushima Daiichi (1) The accident resulted in a severe land contamination (zone VII and 3.2 Aerial releases of radioactive cesium The Japanese regulatory body was expecting that the environmental radiation monitoring data could have been transmitted from Fukushima 320 http://radioactivity.mext.go.jp/ja/contents/7000/6289/24/203_0928.pdf http://radioactivity.mext.go.jp/ja/contents/6000/5043/24/11555_0830.pdf Nuclear Engineering and Design 324 (2017) 315–336 G Saji Fig Wind direction observed at the Fukushima Daiichi site during the active phase (March 11–16, 2011) VIII) covering an area of 380km2, where rehabilitation will be prohibitive without proving a substantial reduction in the dose rate of greater than a factor of 10 (2) The total area of high concentration (> 100 kBq/m2) covers an area of up to 2900 km2, although most of the contaminated area is located in the Abukuma Mountain Chains the reliable source of information, however their data were not available for accident management purposes due to SBO As a quick measure, the on-site data was collected by TEPCO’s monitoring car5, which showed the changing wind directions every few hours during the first week of the accident as shown in Fig (TEPCO, 2011b) This plot was made by extracting the wind direction data from TEPCO’s archived monitoring data, recorded from March 11–31, 2011 Due to the wake (i.e., a track of atmospheric turbulence) of on-site buildings, this wind direction may not be the true direction of the elevated release through the hydrogen explosions The wind direction is shown in a clockwise azimuth with (or 12) pointing to the North and o’clock pointing to the East The original data was obtained by TEPCO’s two monitoring cars, one mostly parked near the Front Gate located to the WSW of the Daiichi and the other near MP-4 in a WWN direction The wind direction indicated by each dot is an average value from the previous dot, since the original data were recorded every ∼ ∼ 10 ∼ 30 ∼ 60 depending on the change in dose rates When the wind direction crosses the 12 o’clock direction, additional data points were included before and after This smoothing of the trend graph was performed to indicate a global trend for the hourly wind direction Even with this processing of the data, the wind direction changed too frequently to identify preferential wind directions that were stable for a few hours When the wind direction value is larger than o’clock, the wind is blowing from land to the Pacific Ocean At the time of the hydrogen explosion in 1F1 (15:36 on March 12), the wind direction was from the ocean to land in a SSE direction This wind direction is consistent with the TV news video where a semi-spherical cloud is moving in a northerly direction Another “internal hydrogen explosion” occurred in 1F2 with the suppression chamber pressure “down-scaled” (off-scaled to zero, at 06:14 on March 15) The wind direction was also from the ocean to land Prior to this event, at 11:01 on March 14, the hydrogen explosion occurred in the Reactor Building of 1F3 At the time of this event, the recorded wind direction was from land to sea However, this information is not necessarily consistent with video coverage at the time of the explosion, as it shows that a tall mushroom cloud was traveling in a southern direction In addition to TEPCO’s monitoring posts, the Fukushima Prefectural Government had 25 monitoring posts scattered around the Fukushima site Their data was also not available for accident management purposes due to the tsunami or failure of their telemetry system, in spite of their battery and engine power backups Fortunately, most continued to Environmental monitoring of radioactive species 4.1 Analysis of environmental monitoring data One of the unique features of the Fukushima Daiichi accident is in the simultaneous evolution of its accident sequences, which resulted in the environmental radiation releases from one unit after another The resultant releases left significant land contamination especially to the northwest of the damaged plants In order to appreciate the scientific implication of the areas marked with the footprints of land contamination, it is necessary to clarify whether they were left with one dominating release event or the superposition of multiple releases from different units at different times The land contamination maps indicate that the radioactive plumes passed over the area thereby contaminating the soil Therefore people who resided in the area at the time of the plume passage may have inhaled the radioactive effluent, especially in the form of 131I and 133I Although an atmospheric dispersion assessment has been used in the investigation of the environmental release and consequence assessment studies (Chino et al., 2011), this approach is not applicable in the case of the FDA The reason why an atmospheric dispersion assessment would not be dependable in the case of the FDA is due to the following limitations: (1) No actual source term data was available to begin with (2) No release height data was known (3) No local wind velocity and on-site wind direction data could be transmitted for assessment due to SBO (4) The terrain of the nearby Abukuma Mountain range is not incorporated in the dispersion model with a flat landscape assumption For example, Iidate-mura is located 500 meters above sea level and is one of the most highly contaminated highland stretching in a NW direction from the Daiichi The village looks like a gorge through which the plumes released at different times passing through this flat With such a terrain, it is necessary to consider a change in the atmospheric pressure due to the height which may have induced the radioactive rainfall Data from the monitoring car substituted the eight failed stationary monitoring stations by periodically measuring doses near the location of each station In addition, a portable survey meter was placed near the front gate, located to the west of the Daiichi There was also meteorological equipment in the car TEPCO’s monitoring stations installed along a semicircular boundary of the site (MP-1 ∼ MP-8) and on the stack should have been 321 Nuclear Engineering and Design 324 (2017) 315–336 G Saji release mechanism Further explanation will be provided in Section 4.3 work silently and automatically stored the data measured during the active phase of the accident on their own hard drives On September 24, 2012, the Fukushima Prefectural Government recovered the radiation monitoring data and posted it on their homepage only in Japanese (Fukushima Prefectural Government, 2012) These data were plotted for each monitoring station by the author and displayed in a form of a “mandala” (pictorial disambiguation)6 as shown in Fig The figure was highlighted with the following screening criteria to select eight representative points from the original 23 data sets released from the Fukushima Prefectural Government home page, namely; 4.2 Chronology of the Fukushima accident The chronology as summarized in Table is extracted from reference (TEPCO's Investigation Report, 2012) to highlight those events directly related with the environmental releases (1) March 11: Dose rate increased in 1F1 R/B only h after the arrival of the tsunami (2) March 12: Two large peaks in dose rate curve recorded by the monitoring station located at Yamada station (Fig 4), occurred from that morning until noon They were also recorded at other monitoring stations located to the NW (#19 Kamihatori, 5.6 km from Daiichi) and NWN (#22 Namie, 8.6 km from Daiichi) as shown in Fig These releases induced severe land contamination Unit is very likely responsible for these events These peaks occurred before the venting operation, indicating leakage from PCV’s flange without the scrubbing effects of the suppression pool water (3) March 13: Two large releases (Fig 4) were from 1F3, although they did not leave significant land contamination, since the step-wise increase in ground shine is only 1.0 nSv/h They were the result of the venting from S/C in which scrubbing of the effluent should have been effective (4) March 14: Hydrogen explosion in Unit It left no remarkable contamination (Figs and 4), since they are not correlated with either the venting operation or the explosive phenomena It means that the aerosol was deposited onto the inner wall of the reactor building by the time of the hydrogen explosion occurred TEPCO’s on-site meteorological data was indicating that the wind was towards the in-land direction; therefore the plume should have been detected, if it reached one of the monitoring posts (5) March 15: Three large peaks in the dose rate curve (Fig 5) are recorded in the monitoring station located in the WNW (#16 Yamada, 4.1 km from Daiichi), leaving behind severe land contamination Unit is very likely responsible for these events In particular the spontaneous depressurization of D/W, which occurred around noon, appears to have been the result of effluent releases without scrubbing by the suppression pool water The leakages occurred three times since step-wise increases in the ground shine after the plume passages are repeated three times These phenomena are likely due to leakages from the D/W flange of the PCV (6) March 16: Following the large releases that occurred on March 15 from 1F2 (Fig 5), two additional large leakages continued through the next day as shown in Fig The amount of these releases are among the largest, and likely from 1F2 Unfortunately TEPCO’s timeline does not cover March 16 and no data is available for further study (7) The exact timing of the hydrogen explosion in 1F4 is not known although the large explosive sound and quaking was felt at 06:14 on March 14 at the Unit side ceiling of Units 3/4's common control room It is likely due to this event, although, damage to the 5th floor of the R/B was visually confirmed at 06:55 on March 15 • One representative location from each of the sectors, covering the western half of the Daiichi • Including hourly data at least until March 15, 2011 These representative data were plotted to cover the most active phase of the accident over six days from March 12 to 18 A zoom-in view of Fig is shown in Fig 4, by focusing on the Yamada Station, Futaba-machi (#16; 4.1 km WNW from the Daiichi), since it contained the most representative overall dose information recorded during the plume passage as well as the ground shine after land contamination in the WNW to NW direction Since Fig fails to show that another large release event occurred at midnight to the early morning of March 15 in a SWS to S direction, Fig is also selected to supplement the missing information In the following explanation, the terminologies (e.g., plume passage dose, ground shine and land contamination) are illustrated in Fig In these figures, a sharp increase in the dose rate curve represents a plume passage dose which decays quickly after its passage After reaching the peak, the dose rate decreases with an asymptotic decay curve which levels off at a larger background than before the plume passage This increase represents a ground shine dose due to land contamination This step-wise increase in the ground shine represents a contamination of soil due to radioactive materials The initial two large releases from 1F1 occurred on March 12 left an order of magnitude larger ground shine than prior to the arrival of the plume The first release is likely due to the overpressure leakage through the dry well (D/W) flange of the PCV’s top head, whereas the second release is due to the venting performed at 9:15 on March 12 The issue of the dry well flange leakage is discussed in Section 5.5 There is also a possibility that the first peak is due to the “internal hydrogen explosion” at the wet well vent pipe induced a small crack since a leakage of water has been identified in the suppression pool room as explained in Section 6.2 The effect of venting is indicated only by a plume passage dose without an increase in ground shine The heavily contaminated corridor stretching to the northwest is likely the superposition of the large releases from 1F1 on March 12 (shown in Fig 4) and from 1F2 during March 15–16 (shown in Figs and 5) The latter induced an additional increase in contamination by orders of magnitude on top of the ground shine left from Unit These correlations are identified by collating with the chronology of the accident as summarized in Table in comparison with the environmental contamination maps The large release occurred on March 15 at the time when the explosive sound occurred in Unit 2’s suppression pool as shown in Fig This graph is connected with the three large release events recorded in Fig on March 16 The wind direction changed from SWS on March 15 to WNW on March 16 Unit appears to have released effluent repeatedly for more than a day, thereby heavily contaminating the near field of the Daiichi The radioactive iodine was the main constituent of the release This is a unique feature of the releases indicating a different 4.3 Reconstruction of land contamination densities due to iodine In general, retrospective reconstruction of 131I map is very difficult due to its short half-life However on June 27, 2013, JAEA (Japan Atomic Energy Agency) gave a press release on their successful reconstruction of the 131I land deposition maps based on the spectral data taken by the US DOE through their aerial remote monitoring operation performed from April 2–4, 2011 (Torii, 2013) In their analysis, Dr T Tori performed a reverse (unfolding) Monte Carlo Simulation by incorporating the attenuation of the gamma ray from the source to the detector as well as the detector characteristics The term “mandala” is a religious picture often used in Buddhism to illustrate the spiritual world as in the case of a Russian icon in Christianity In the picture, Buddha is usually surrounded with the images of saints 322 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Fig Survey map of environmental radiation monitoring data recovered by the Fukushima Prefectural Government It has often been assumed that the radioactive iodine was released in the form of CsI The author has confirmed the chemical stoichiometry by investigating the numerous reported soil contamination samples taken in the highly contaminated region stretching in the NW direction from the Daiichi However the 131I map created from the newly developed method indicates that iodine is not necessarily released as CsI Dr Torii’s 131I map is more reasonable since the estimated shutdown inventories of 131I/134Cs at the time of the accident are 2292/170 PBq, respectively, for both Units and The elemental iodine is scarcely soluble in water, although it is a strong oxidizing agent Therefore, the since the peak energy (365keV) was barely detected Fig is copied from their press release The figure shows a comparison of the land contamination density (Bq/m2) of 131I (left) and 134Cs (right) as of June 16, 2011 In comparison with the 134Cs map the 131I map indicates: • Large near field deposition of I occurred in the southern direction as far as 20 km away from the Daiichi; • Cs induces highly contaminated corridor stretching from ap131 134 proximately 20–30 km NW from the Daiichi more heavily than 131I 323 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Table Simplified chronology directly related to environmental releases Day Unit Time and Events 3/11 All All 14:46 Earthquake 14:48 Reactor trip 15:37 Tsunami induced SBO 21:51 Dose rate increased in 1F1 R/B 3/12 1 1 02:30 D/W pressure reached 840 kPa (abs) 04:23 0.59 µSv/h at front gate D/W head flange leakage started(?) 09:15 Vent valve (MO) was manually opened 14:30 D/W confirmed depressurized 15:29 1015 μSv/h at MP4 15:36 Hydrogen explosion in 1F1 R/B 3/13 3 3 3 08:41 08:56 09:08 09:20 14:15 14:31 21:10 3/14 06:14 Large explosive sound and shaking of the ceiling at the Unit side of Unit 3/4 common control room 07:20 D/W pressure stabilized at 0.5 MPa (abs) 10:30 High dose rate in 1F4 prevented worker entry to the R/B 11:01 Hydrogen explosion in 1F3 R/B 11:15 RPV pressures 0.195/0.203 MPa (abs) at Chanel (A/B) D/W Pressure 0.330 MPa (abs) and S/C pressure 0.390 MPa (abs) Both RPV and PCV were judged sound 23:46 D/W pressure 750 kPa (abs) Fig Aerial doses at Yamada, 4.1 km WNW of Fukushima Daiichi 3 3/15 Fig Aerial doses at Matsudate, 14.2 km SWS of Fukushima Daiichi 2 2 vent line configured to the rupture disk 882 μSv/h at MP4 (8:41 venting started) RPV depressurized and water injection through DDFP DW is judged likely vented 905 μSv/h at MP4 Dose rate inside R/B 100 ∼ 300 mSv/h Confirmed pressure decrease in D/W 00:05 D/W Pressure 740 kPa (abs) Unable to vent (D/W head flange leakage started?) 06:14 Large explosive sound and floor quaking D/W pressure 130 kpa(abs), S/C pressure kPa Evacuation of staff 06:55 Damage on 5th floor of R/B confirmed 09:30 11,980 μSv/h at Front Gate 11:20 D/W pressure 730 kPa (abs) 11:25 D/W pressure 155 kPs S/C pressure 16:05 ∼ S/C venting interrupted but recovered This activity repeated several times through early April plumes traveled in accordance with the thermo-hydraulic effects (i.e., temperature of the initial released steam containing radioactive species and release height as well as wind direction and velocity) of atmosphere and not a simple dispersion mechanism applying Gauss’s atmospheric turbulence model Fig Mechanism of changing dose rate curves during the plume passage at each monitoring station 4.4 Summary of environmental releases in correlation with chronology The review of the environmental dose rate correlated with the chronology of each unit clarified the following points: radioactive release as CsI may indicate that the release is from a water environment, such as from the suppression pool water, in which CsI has been dissolved The release of mostly iodine appears to indicate that it is directly from the D/W of the PCV without going through the suppression pool water By correlating the 131I land deposition map (Fig 5) with the chronology (Table 6), the internal hydrogen explosion occurred in Unit S/P at 06:14 on March 15 which induced a large release of 131I to the SWS direction of the Fukushima Daiichi site The potential flow of the plume with the high concentration of 131I in a southern direction from the Daiichi is a new finding and important to public safety Also the new 131I land contamination map is puzzling from the point of view of its release mechanism The highly contaminated corridor stretching in a NW direction from the Daiichi is likely the result of numerous plumes released from both Unit and and occurring at various times This is implied by the near field (< km) 134Cs land contamination map as shown in Fig In the figure, the routes of the numerous plume passages were more clearly reproduced These footprints indicate that the (1) A series of venting operations after scrubbing with the suppression pool water (W/W venting) left minor soil contamination in 1F1 and 1F3 (2) A series of “spontaneous venting” events appear to have occurred in Unit after the hydrogen explosion involving S/C The explosion deteriorated the dry well flange seals leading to the effluent leakage When the overpressure in the PCV is reduced through the spontaneous venting, it leads to the reduced pressure boiling of the suppression pool water releasing a large amount of dissolved radioactive species contained in the S/P (3) Although the devastation of the secondary containment system (i.e R/B) should have been prevented, it did not result in a large release upon the explosion in 1F3 It is likely due to the precipitation of particles in the aerosol before the explosion As a matter of fact the rubble of the building was reported to be highly contaminated 324 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Fig Comparison of land contamination density (Bq/m2) of 131 I (Left) and 134 Cs (Right) as of June 16, 2011 including soil sampling data taken at each spot Hydrogen generation and explosions closed configuration, this portion of the pipe is left cold, allowing condensation of steam from the RPV during normal operation The condensation separated the dissolved hydrogen and oxygen gases generated due to the radiological decomposition of the reactor water Upon commencement of the routine ECCS test, the mixed gas exploded and shattered the OD 165 mm pipe with an 11 mm wall thickness Since the isolation valves were closed automatically 30 s after the breach, the accident was terminated without developing into a small LOCA In spite of the extensive follow-up experiments performed with closed pipe containing mixed gas and steam, only cases exploded out of 104 tests performed with various noble metal concentrations at the surface The author suspects that radiation may become an ignition source, since the high energy charged particles produce “spurs” which is an agglomeration of the secondary charged particles before being thermalized Inside the “spurs” the temperature should be extremely high at the molecular scale The Hamaoka hydrogen explosion issue was left unresolved at the time of the FDA The regulatory body (NISA at that time) instructed the owners of the 14 BWRs with similar ECCS designs to remove the accumulated gas and water before conducting the monthly ECCS tests For the removal of gas it was necessary to cool down the piping by closing the isolation valves This preparatory operation usually took several days but the ECCS operation without the removal of the accumulated hydrogen was feared to be risky It is a strange directive since accidents in need of the activation of ECCS cannot be scheduled At the time of the FDA, the circulation of reactor water was terminated, which should have induced a more favorable situation for the accumulation of hydrogen and oxygen gas such as in the RCIC and HPCI steam turbine piping Unfortunately, direct evidence of this mode of pipe rupture has yet to be confirmed in 1F1 Nevertheless a steam jet ejection was observed on June 4, 2011 from the pipe sleeve penetrating the floor near the wall 5.1 Early pipe break and hydrogen release (1F1) As listed in Table 6, an increase in dose rate inside of the 1F1 R/B prevented the entry of workers as early as 21:51 on March 11, approximately h after the arrival of the tsunami Entry to the R/B was restricted at 23:05 in consideration of the high dose rate This resulted in a serious restriction for accident management purposes such as venting by manually opening the vent valve Nevertheless TEPCO’s staff entered the R/B for venting, resulting in a dose > 100 mSv even when wearing an “air set” (respirator) This phenomenon motivated some Japanese scientists to suspect an earthquake-induced pipe break or failure of the Isolation Condenser (I/C) leading to a very early core melt down event However this author believes that it is due to an “internal hydrogen explosion” which is preceded by the earlier event at the Hamaoka Unit in Japan followed with Brunsbüttel BWR in Germany, both of them occurred in 2001 This mode of pipe break could have been prevented if more in-depth studies had been completed as explained below A brief summary of this accident is appended in the Annex C The Hamaoka accident is important due to the fact that a hydrogen explosion inside of the primary steam-water coolant may induce an “internal hydrogen explosion” when dissolved hydrogen is separated from the primary coolant This separation mechanism should have also existed in the W/W vent pipe during the course of the accident in which the suppression pool water which reached temperatures as high as 160 °C The Hamaoka accident occurred in a part of the ECCS (RHRS) at the top portion of the steam piping from the RPV During operation, this portion of the stand pipe is open to the steam of the primary circulation water at its inlet side, however, valves are kept closed at the exit to the RHRS Heat Exchangers which is at an ambient temperature Due to this 325 Nuclear Engineering and Design 324 (2017) 315–336 G Saji 130 kpa(abs), S/C pressure kPa The author believes that this event should be due to an “internal hydrogen explosion” in the suppression pool (S/P) At the exact same moment, another large explosive sound and quaking of the floor was observed in the common control room of Units and Since no down scaling of the S/P pressure was reported in Unit 3, the author judges that the event occurred in 1F2 but not 1F3 5.2.1 Hydrogen generation through radiation-induced electrolysis Since the scientific cause for a series of hydrogen explosions during the FDA has not been established, the author investigated his basic theory named “radiation-induced electrolysis (RIE)” by applying the estimation of the amounts of H2 generation during the active phase of the FDA (Saji, 2016) The author’s theory was originally developed by including Faraday’s Law of electrolysis into the basic time-dependent material balance equation of radiation-chemical species for his study on accelerated corrosion phenomena (Saji, 2017) As such this theory applies to the early phase of the accident before the loss of water levels in the reactor cores, although the following simulations were performed from the time of the seismic reactor trip to the hydrogen explosions, since the water levels were recovered by water injection Through this mechanism as much as 58,000 m3-STP of hydrogen gas is estimated (Fig 9) to have accumulated inside the 1F2 PCV prior to the “internal hydrogen explosion” which occurred at 06:14 on March 15 With this large volume of hydrogen gas the explosion was a viable possibility Fig Near field land contamination maps of 134Cs (Note) The 134Cs map was shown here instead of 137Cs since the three distinct highly contaminated streaks are more clearly distinguished, due to the grouping of the contamination density data of the TIP (Travelling In-vessel Probe) room in 1F1 The contamination level on the floor was 4.7 Sv/h, one to two orders of magnitude larger than from nearby areas These data were acquired during the robotic survey inside the R/B On November 14, 2012 the floating robotic inspection inside the S/C Room (the room housing the S/C) identified leakages from one of Sand Cushion Drain Pipes and from the upper portion of the S/C Vent Pipe This indicates that there would have been a pipe break somewhere in the upper part of the RPV (TEPCO Handout, 2013a) The leakage rate was as large as tap water coming out from a garden hose (TEPCO Handout, 2013b) All of the leaked water initiated from the upper part of the RPV was collected in these pipes However potential damage of piping in the upper part of PCV has not been performed yet, since there is not enough space to insert a robotic camera Fig Accumulated H2 in Unit (m3-STP) The calculation is made to estimate the total hydrogen inventory within the primary water pressure boundary, which includes its recirculation piping and the RPV The top half of the RPV was assumed initially filled with steam The hydrogen generated in the irradiated reactor water through the “radiation-induced electrolysis” mechanism is released mostly in the top half of the RPV The pressure increase in the reactor vessel is released to the suppression pool through safety valves (spring action) and release valves (nitrogen gas-driven) Therefore most of the hydrogen gas should have been accumulated in the S/C The damage to the 1F2’s blowout panel, which is said was induced by the hydrogen explosion7 of 1F1 at 15:36 on March 12, facilitated the release of hydrogen gas which likely prevented a severe explosion in the R/B of 1F2 However, the “internal hydrogen explosion” of the Suppression Pool resulted in the leakage of a large amount of radioactive species through this damaged blowout panel thereby severely 5.2 Hydrogen explosion in 1F2 suppression pool The series of hydrogen explosions, devastating the reactor buildings one after another, are the most dramatic events during the Fukushima disaster; however the 1F2 R/B evaded this form of hydrogen explosion At this unit an outstandingly large amount of radioactive leakage to the basement of the T/H has been detected from March 24–27, 2011 suggesting a leakage of the suppression pool water Also, as presented in Table 6, a large explosive sound and quaking of the floor were observed in the suppression pool area of 1F2 on March 15, 06:14 The control room staff observed the D/W pressure was at The exact timing and cause for activation of the blowout panel is not known It is also said to have been activated through the hydrogen explosion in 1F3 which occurred at 11:01 on March 14 In spite of these uncertainties, the implication of the following discussion is unaffected 326 Nuclear Engineering and Design 324 (2017) 315–336 G Saji suppression chamber the vent pipes exhaust into a toroidal vent header which extends circumferentially all the way around the inside of the suppression chamber Extending downward from the vent header are ninety-six down-comer pipes which terminate about three feet below the suppression pool minimum water level” The down-comer pipes opened under the pool water should have prevented the nitrogen gas covering the suppression water surface In this basic configuration, the atmosphere of the “free air volume of the Suppression Pool” was likely air sealed in from the time of the last pool water exchange for maintenance and inspection The periodic repainting maintenance is indispensable since the suppression chamber is constructed of carbon steel in which a large volume of pool water is stored Under this situation, as much as 6900 N-m3 of oxygen could have existed in the “free air volume” of the S/C This amount is not far from the chemical stoichiometry compared with 20,000 N-m3 of hydrogen injected through the SRVs into the S/C through the pressurized “feed and bleed” operation In addition, since the suppression pool water was not degassed, there would be as much as 19.4 N-m3 of dissolved oxygen when the water is saturated with atmospheric oxygen at room temperature, although not comparable with the oxygen in the cover gas Therefore, a more than sufficient volume of oxygen is available for the “internal hydrogen explosion” This explosion would have induced a distortion of the flange and failure of the seal due to the over pressure from the explosion as illustrated in Fig 11 The mixed gas of hydrogen and oxygen should have existed equally in Units 1–3, however the hydrogen explosion in S/C appears to have occurred only in Unit When the “internal hydrogen explosion” occurred in the “free air volume” of the S/C in Unit 2, the resultant high overpressure induced down pressure on the surface of the suppression pool water This down pressure forced the suppression pool water to flow back through the “eight large vent pipes (81″ in diameter)” to the bottom head of the PCV The resultant huge high speed water jets collided at the bottom head and should have induced violent splashing of water The splashing dislodged the gratings embedded in the platform installed for exchange of control rods as illustrated in Fig 11 This phenomenon appears to have actually occurred at the time of FDA as observed during the recent robotic inspection performed down to the pedestal region of the PCV through the CRD exchange rail (TEPCO, 2017a) contaminating the NW direction of Fukushima Daiichi Most of the releases were from the wet well where the decontamination factor of the suppression pool water is as large as two orders of magnitude However, a series of the dry well head flange leakages occurred without this decontamination process Thus a severe environmental contamination occurred in the northwest direction of the Fukushima Daiichi 5.2.2 Failure of RCIC and “feed and bleed” operation The decay heat removal in 1F2 was established by the RCIC which was manually started at 15:39 of March 11, soon after the tsunami’s arrival It was configured by injecting water from the Condensate Storage Tank (CST) which contained pure water at the ambient temperature In order to save the water inventory of the CST, the water source was switched to S/C at 2:55 on March 12 followed with seawater injection initiated at 19:54 on March 14 in 1F2 In spite of these water injection efforts, the decay heat was removed through bleeding of steam vapor since no heat sink was available due to LUHS thus “feed and bleed” cooling should have been effective Due to the RPV pressure increase, Safety and Release Valve (SRV) started to bleed the steam into the S/C, resulting in a gradual decrease of the water level as shown in Fig 10 The safety valve function of SRV is spring action and works without high pressure I/C nitrogen gas Upon failure of RCIC, the water injection operation was performed after depressurization of the RPV The “feed” operation was performed with fire trucks Therefore, “feed and bleed” cooling was the main mechanism for decay heat removal due to the LUHS during the active phase of the FDA It should also be noted that even when the water level decreased to the bottom of the active fuel, there was a large volume of water left in the lower plenum of the RPV This water should have contributed to the cooling of the core to a certain extent, however its effect in preventing core melt is still controversial This question may not be answered until the actual status of the fuel and core debris are confirmed years from now 5.2.3 Hydrogen release into suppression chamber At the time of the RCIC failure, approximately 5.8 × 104 N-m3 of hydrogen accumulated in the total primary water (water in the RPV and in recirculation lines) of approximately 900 m3 From this volume, approximately × 104 N-m3 of hydrogen gas is estimated to be released into the S/P with a water volume of 3800 m3 in Mark containment Although the D/W atmosphere is nitrogen, there is no specification of the suppression pool atmosphere This implies that the air atmosphere could have existed in the S/P’s atmosphere (GE Technology Systems Manual) However, the GE Report explains “during normal operation, the drywell atmosphere and the wet well atmosphere is inert (filled with nitrogen), and the wet well water is at the ambient temperature” (GE Report, 03/19/2011) This report describes the Mark I containment design currently in use at the 23 U.S reactors and its ability to fulfill its safety function in containing fission product releases under design basis conditions This practice may not have been followed at the Fukushima Daiichi As a matter of fact, there is no nitrogen gas charge line connecting directly to the S/C, although the charge lines are found connected to the D/W in TEPCO’s illustration (TEPCO’s Press release, 2015) which is also posted by the Nuclear Regulatory Agency The 1F2 S/P atmosphere was changed to nitrogen gas two years after the accident by way of the oxygen analysis rack line The S/C is designed to release the steam-water mixture to the suppression pool by postulating a DBA-LOCA The following explanation is given in GE’s Technology Systems Manual “The interconnecting vent network is provided between the drywell and suppression chamber to channel the steam and water mixture from a LOCA, to the suppression pool and allow non-condensing gases to be vented back to the drywell Eight large vent pipes (81″ in diameter) extend radially outward and downward from the drywell into the suppression chamber Inside the Fig 10 RPV water level decrease with “feed and bleed” operation 327 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Fig 11 “H2 Explosion” in S/P → Flange Leakage trip If the debris contained some fraction of the core materials, steam bubbles should be visible However the observed debris appear to be cold in temperature This observation suggests that the debris should contain solidified molten metals They should consist of low melting temperature steel structures such as control rods, control rod housings and stub pipes, all of them are made of stainless steal It is likely that the fuel rods were standing vertical or broken into pieces and then theses were gradually melted in-core low melting point structural metals A part of the fuel debris should have been piled up at the bottom of RPV since its density is higher than the molten steel The heated molten steal structure eventually broke the reactor pressure boundary and drooped down into the water, either solidified or deposited at the bottom of PCV During this process, the initially hot molten stainless steel droplets induced a gigantic turbulence of the accumulated water, which dislodged the steel grating placed on the platform of the PCV This scenario is consistent with the current ambient temperature (23.4∼25.6 °C) of the accumulated water at the bottom of PCV 5.3 Core melt and the most serious hydrogen explosion in 1F3 The most severe hydrogen explosion during the FDA occurred in 1F3 at 11:01 on March 14 However this event left no footprint in any of the radiation monitoring records as shown in Figs and 5.3.1 Identification of non-heating core debris in 1F3 Starting July 19, 2017, TEPCO released a series of results of investigation of Unit PCV by employing a newly developed submarine type (ROV) robot The most remarkable finding is that they finally identified “core debris” drooping down like a stalactite from the bottom head of RPV and then piled up inside of the accumulated water at the bottom of PCV (TEPCO, 2017c) Some of their photos are copied as shown in Fig 12 Since these photos were released as a TEPCO’s press handout, detailed explanations were not included In spite of lack of details, these pictures are very important in contemplating what occurred in 1F3 during the active phase of the FDA In the photos it is remarkable to observe that there are no steam bubbles, although the estimated total decay heat is as large as 127 kW in 1F3 after 2000 days since the reactor Fig 12 Photo images of core debris found at the bottom of 1F3 PCV (TEPCO, 2017c) 328 Nuclear Engineering and Design 324 (2017) 315–336 G Saji boundary but solidified to plug the crack Therefore initiation of the core melting should have been triggered by the loss of RPV pressure occurred after March 12 at 13:05, just one day after the reactor trip This early loss of reactor pressure indicates an occurrence of a “small break LOCA” by accelerating penetration of preexisting stress corrosion cracking in some in-vessel structures, such as the CRD housings and stub tubes Unfortunately this type of hidden defects were just started to come to the attention of the nuclear industries such as by EPRI (EPRI, 2000) Also in Japan a guideline for inservice inspection of RPV was compiled in (JANTIVIP06 2008, only in Japanese) It is likely such inspection has never been performed before the FDA, since there were no TEPCO’s press release dealing with the ISI for the in-vessel components although the industries’ concern was focused on replacement of reactor core shrouds just before the FDA Recently the author published the basic mechanism of accelerated corrosion of in-vessel components (Saji, 2017) With this mechanism, the reactor core behaves as a gigantic cathode whereas the out of core region becomes anodic, where SCC is accelerated due to de-passivation Although the hydrogen explosion in Unit was triggered due to the gigantic tsunami, it was a coincidence that the small leak from the small break LOCA occurred simultaneously while operating RCIC The author sincerely request nuclear industries to perform ISI of in-vessel components of BWR with the similar design 5.3.2 Early termination of water injection with RCIC Despite SBO, 1F3’s battery power survived unlike in the case of both 1F1 and 1F2 By saving consumption of the battery power, RCIC and HPCI were utilized for keeping its water level above the top of the active fuel zone The RCIC was manually started at 15:05 on March 11, however it automatically tripped at 15:25 due to the high water level in the RPV In order to save the battery power, the operators adjusted the flow rate settings and manually adjusted the flow rate to avoid tripping and restarting cycles With this manipulation, RCIC continued to inject water until March 12 at 11:36 The RCIC became inoperable due to low RPV steam pressure, which drives the steam turbine of RCIC The RCIC system operates on high-pressure steam from the reactor itself, and thus is operable with no electric power other than battery power to operate the control valves TEPCO operators tried to maintain the reactor water levels within ± 75% of top and bottom of the active fuel level At 12:35 of March 12, the HPCI automatic started when detecting the low water level set point in RPV Since the HPCI had higher capacity than the RCIC it consumed more steam for driving the HPCI turbine Since the HPCI was used at pressure lower than the minimum operation pressure (0.69 MPa), its operation was continue to be unstable On March 13 at 9:25 the RPV was depressurized and HPCI was switched to DDFP to splay cool the suppression pool water The RPV pressure decreased to such a low level that inhibited further operation of both RCIC and HPCI at 11:36, with only 20 h of total operation time In contrast with the Unit 3, the Unit 2’s RCIC appears to have worked at least until its depressurization of RPV at 18:02 on March 14, with the total operation time of 75 h This comparison indicates that there should have been a leak from the “Small-Break LOCA” in the 1F3 RPV In order to confirm this hypothesis, changes in the 1F3’s RPV pressure and water level are plotted in Fig 13 Note that RPV pressure was lost on March 12 at 13:05 but temporally recovered at March 13 on 5:00 It is very likely its water level was around the bottom of active fuel level (BAF), although the reactor water level gauges were not available until March 13 at 4:00 (The water level gauge failed on March12 at 20:36) Therefore, the active part of the reactor fuel was exposed to steam while its lower portion was mostly kept under water for days With the fuel uncovered after 3/12 13:05, it is likely that the low melting point core metals should have been molten down to the bottom head of RPV and penetrated the reactor pressure 5.3.3 Environmental releases from 1F3 The vent line was configure at 8:41 on March 13 up to the rupture disk which is installed between the final vent line and exhaust stack Soon after that, at 8:56 on March 13, a dose rate of 882mSv/h was detected by a monitoring car which stayed near MP This indicates that the venting was successful although the status of the rupture disk is still unknown The rupture disk is designed to burst at 427 kPa (gauge), which appears to protect the PCV whose design pressure is approximately 0.38MPa (gauge) During the venting operation there was no record that the D/W pressure reached that level The containment leak rate test is performed at 0.26 MPa (gauge) in which the allowable leak rate is 0.348%/d for 1F1 The design philosophy for installing the rupture disk on top of the final vent valve is not appropriate As a matter of fact TEPCO falsified this test and was examined by organizing an independent outsider investigation committee (Investigation Committee Report, 2002) In the report, there are several Fig 13 1F3 RPV pressure (MPa) and water level (m) 329 Nuclear Engineering and Design 324 (2017) 315–336 G Saji statements on installing “diaphragms” at the flanges among the measures to avoid leakage from valves Although no further statement is reported, it is very likely the “diaphragm” means that the rupture disk was installed to comply with the regulatory leak rate in the other Fukushima Daiichi units Because of this particular design configuration, the venting operations appeared to have performed below the burst pressure of the rupture disk, resulting in a release of hydrogen rich steam and radioactive effluent throughout the R/B by way of the SGTS ventilation duct network As a matter of fact, the decrease of the D/W pressure was confirmed at around 9:20–9:35 on March 13 At 14:45 on the same day, the dose rate inside of the R/B surged to 100–300 mSv/h The workers judged that the hydrogen explosion of the R/B was imminent and started to evacuate the field crew but the crew resumed their activity before 17:00 Similar tense situations continued for the crew continued until the hydrogen explosion occurred at 11:01 on March 14 Nevertheless, there remained no footprint from the radioactive releases recorded by monitoring stations even at the time of the hydrogen explosion as indicated in Figs and It is likely that most of the radioactive effluent released inside the R/B deposited on the walls and ventilation filters resulting in no serious environmental contamination In fact, when estimated in a similar method as done in Section 5.2 approximately 13,000 N-m3 of hydrogen gas is released into the R/B The temperature of the suppression pool water is assumed to have been 150 °C with a water volume of 2980 m3 The total air volume of the R/B is assumed to have been at 180.000 m3 The estimated hydrogen concentration is 7%, which is well above the lower hydrogen explosion limit in air With these results it is not surprising to have experienced the severest hydrogen explosion in 1F3 but not sufficient in hydrogen volume enough to induce the explosions in both 1F3 and 1F4 This issue is discussed in the next section still left unresolved whether the hydrogen explosion occurred at 06:14 on March 15 was in 1F2 or 1F4 However the largest release of radioactive inventory into the basement of the 1F2 T/H supports the failure of the S/C in the 1F2 unit In addition, the explosion left no remarkable land contamination around 1F4 The amount of hydrogen generation through “the radiation-induced electrolysis” has been reported in (Saji, 2016) Perhaps a part of the hydrogen is due to an unintended flow from 1F3 R/B and a part is due to the radiation-induced electrolysis from 1F4’s Spent Fuel Pool 5.5 Performance of drywell flanges during the accident Mohamed M Talaat and his colleagues recently conducted a stateof-the-art thermo-mechanical finite element analysis (Talaat, 2014) A previous overpressure evaluation conducted in 1990 identified leakage at the D/W flange connection as the controlling containment failure mode The previous evaluation was based on hand calculations assuming an approximate structural model for analysis by formula It did not include the effects of temperature variation through the wall thickness and the rotation of the D/W head flange due to the sporadic bolt clamping force Their conclusions include the following results At temperature levels of 150 °C and 205 °C, the silicone rubber O-rings are still effective in resisting leakage, and the median overpressure capacities are 0.86 and 0.72 MPa (125 and 104 psi), respectively In spite of these theoretical estimations there is a little known of the leakage test results performed at the Brunswick Mark I containment in the 1970s During the test workers found that the containment pressure of 70 psi (482 kPa) (gauge) pushing upward against the inner dome of the drywell head lifted it off the drywell flange enough to provide a pathway for air to leak from the containment At Fukushima NPPs the design pressure of the PCV is 0.43 MPa (gauge) for 1F1 and 0.38 MPa (gauge) for 1F2 ∼ 1F4 It is mandatory to perform the leakage rate test every year In 1F1 the regulations required that the leak rate be 0.348%/d at 0.26 MPa (gauge) Due to the large volume of PCV, the leak rate test with a nitrogen atmosphere at the design pressure was judged to be impractical However, such a leak rate test at nearly half of the design pressure does not guarantee the safety function of the dry well head flange during an accident A potential leakage of the double O-ring seal is checked by a sampling pipe in 1F1 The head flange leakage had likely occurred in 1F1 as indicated in the first large peak observed during the later part of the morning on March 12 (Fig 4) At 2:30 on March 12 D/W pressure reached 840kPa (abs) whereas the design pressure is 530kPa (abs) TEPCO’s staff manually opened the Large Vent Valve (motor driven) at 9:15 on March 12, however no evidence could confirm its success until installing a portable air compressor to open the Vent Valve (AO) at 14:30, by then they found that the D/W pressure had already reduced This indicates that the “spontaneous venting” should have occurred by way of the head flange leakage This is also the case in 1F2 until 06:14, March 15 when a large explosive sound and floor quakes were experienced The D/W pressure was 130 kpa(abs) and S/C pressure went down to 0kPa This “internal hydrogen explosion” should have resulted in a distortion of the D/W flange inducing a series of puff releases which continued through March 16 as shown in Figs and Although the Mark I PCVs appears to perform well during the FDA, the hydrogen explosion inside the S/C destroyed the containment function in 1F2 Lastly no footprint of the flange leakage was identified in 1F3 via the monitoring data It is due to the fact that the PCV pressure never reached its design pressure due to repeated venting This venting activity repeated several times throughout early April, while the seawater injection continued during this period 5.4 Hydrogen explosion in 1F4 The discussions in Section 5.3 on the spread of the vented hydrogen gas into the 1F3 R/B explains a potential root cause for a portion of the hydrogen gas also released into the 1F4 R/B In 1F4 the entire core loading of fuel assemblies were evacuated to the SFP, therefore no reactor configuration has existed at the time of the FDA TEPCO has been explaining the root cause of the hydrogen explosion in 1F4 is due to the venting of 1F3 through their common vent stack This route is unlikely due to the rupture disk installed in the vent line prior to the common stack However, the vented hydrogen gas spread by way of the duct network and should have exhausted through the common R/B exhaust stack, which measures 120m in height The stack is also shared with 1F4 for exhaust of ventilation for the R/B Before discharging the potentially contaminated room air, these exhausts are filtered with SGTS filter banks which are installed both in 1F3 and 1F4 ventilation exhausts TEPCO found in their post-accident survey that the dose rate of the filter bank is 6.7 mSv/h at the outlet side and 0.1 mSv/h at the inlet side, indicating that the radioactive effluent should have flowed from the 1F3 exhaust to 1F4 R/B Although the deposition of radioactivity on the outlet side of 1F4 SGTS filter may be due to the hydrogen explosion in 1F3 but may not support continued inflow of hydrogen gas from 1F3 The hydrogen explosion in 1F3 occurred at 11:01 on March 14, whereas the hydrogen explosion of 1F4 is said to have occurred at 06:14 on March 15 almost one day later The leakage of hydrogen gas from 1F3 to 1F4 should have been terminated upon the hydrogen explosion in 1F3 The exact timing of the hydrogen explosion in 1F4 has not been well established The staff staying in Units and 4’s common control room heard a large explosive sound and quaking of the ceiling on the 1F4 side at around 06:14 on March 15 Although TEPCO’s seismic wave analysis suggests that this quaking should have occurred at 1F4, there is no exact evidence that was due to the hydrogen explosion at this unit It is Discussions and lessons learned Isolation of radiation from the public by providing containment 330 Nuclear Engineering and Design 324 (2017) 315–336 G Saji systems is the fundamental requirement of nuclear safety This fundamental safety provision was seriously degraded during the FDA The containment system consists of primary and secondary containment where the latter was devastated in all Units Unit 1’s primary containment vessel appears to be damaged due to the overpressure leakage of the D/W flange in spite of death-defying manual venting operation in the high radiation environment as early as h after the arrival of the tsunami TEPCO’s staff were exposed to over 100mSv irradiation in very short work time inside the R/B In Unit 2, the hydrogen explosion in the suppression chamber appears to have induced the series of D/W head flange leakage events The most serious damage experienced to safety provisions, which are provided to impede the accident progression, can be summarized as below: valves should have been arranged in parallel in consideration of the single failure criteria Such a parallel isolation valve configuration is actually used in the BWR/2 Core Spray System, however it was changed to a serial arrangement with one valve left closed in each train of valves in the BWR/2 ECCS design (GE Technology Systems Manual – NRC) This is also logical but having a configuration where two isolation valves are normally open is not comprehensible Unfortunately the detailed drawings are not disclosed, the author is unable to clarify where this issue is rooted The author was told during a symposium that TEPCO‘s position is that these two accidents are not correlated Nevertheless, the risk of an “internal hydrogen explosion” should have existed in Unit (1) “Internal hydrogen explosion” likely occurred in 1F2’s Suppression Pool which resulted in the failure of PCV Together with the loss of the Secondary Containment function, the radioactive effluent was directly released into the environment and induced the most serious environmental contamination (2) Devastation of the reactor buildings, which comprise a part of the secondary containment boundary to halt the environmental releases of the radioactive effluent The loss of the secondary containment system was experienced in 1F1, 1F3 and 1F4 This mode of failure was evaded in 1F2 through an inadvertent actuation of its blowout panel but resulted in the loss of its containment function Fukushima Daiichi NPPs were constructed by applying the concept of Mark I Containment which utilizes a toroidal shaped suppression pool These primary containments are enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield and auxiliary and enclosure buildings Most of the auxiliary systems which deal with the primary water (e.g., ECCS) are located inside of the secondary containment boundary Although the safety role of the secondary containment was of concern as early as 1988 by S.R Green (Greene, 1988), their purpose was assumed to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents Green was already concerned that the deflagration of this hydrogen within the secondary containment would result in pressure loadings which might threaten the structural integrity of the secondary containment The FDA was not contained within the DBA-LOCA It was an un-experienced beyond the design basis accident due to LUHS combined with long-term SBO The mitigation strategies contemplated at that time depended heavily on the SGTS, which was not available due to long-term SBO However, the FDA revealed that the reactor building is a kind of prefabricated structure without safety enhancements The design basis leak rate is 50–100% per day at 30 mm Aq, which is a de facto industry standard for the heating and ventilation system design for ordinary concrete buildings without windows such as in the case of warehouses Therefore how hydrogen was generated and leaked into the secondary containment is one of the key points of the FDA As to the hydrogen generation during the severe accident, the author has theoretically demonstrated that a large volume of hydrogen can be generated through “radiation-induced electrolysis” even without a core melt The theory integrates Faraday’s law of electrolysis into the radiation chemical material balance equation (Saji, 2016) As to the second point, how hydrogen leaked into the reactor building, it is partly through the leakage of the vented hydrogen into the general ventilation duct network of the reactor building rooms According to reference (TEPCO's Investigation Report, 2012), there are two vent lines, one for D/W venting and the other for W/W venting These two vent lines joins together into a single line towards their common vent stacks (shared between Units and Units and share another common vent stack) However, before leaching to the stack, it is also connected to the general ventilation duct network of the reactor building in which SGTS is installed It is likely that the designer expected mitigation of effluent through the decontamination factor of the SGTS which has both a re-combiner and aerosol filters However, due to SBO, SGTS was not working and moreover, the isolation valves (dampers) failed to close due to SBO Rather the duct work helped in spreading hydrogen gas throughout the entire reactor building Separation and independence of the “hardened vent line” (USNRC, 20122015) was not integrated in the Fukushima Daiichi In addition to this mechanism, there is a high probability of leakage from the flanges of the PCV due to over pressure while waiting for permission to vent at Fukushima As a matter of fact as early as 23:50 on March 11, TEPCO’s 1F1 staff confirmed a pressure of 600 kPa(abs) 6.2 Devastation of the reactor buildings The apparent cause of this phenomenon is undoubtedly the hydrogen explosion As reported in reference (Saji, 2016, 2017), generation of hydrogen during severe accidents may not be limited to a high temperature zirconium-steam reaction Rather the “radiation-induced electrolysis” mechanism should have induced a large amount of hydrogen generation in the BWR water chemistry Unlike in the case of PWRs, hydrogen is not dosed to suppress the radiological hydrogen generation above the “critical hydrogen concentration” (Elliot and Bartels, 2009) 6.1 Pipe break through “internal hydrogen explosion” As summarized in the Annex C, the hydrogen explosion occurred during a routine ECCS test at Hamaoka Unit in 2001 The pipe ruptured as a part of ECCS (at the top portion of the steam piping from the RPV) in which the steam was condensed separating the hydrogen and oxygen A similar pipe rupture incident occurred one month later at the Brunsbüttel NPP in Germany The cause of the ignition was not clearly identified but it is suspected to have been induced by minuscule particles of noble metals (in the case when using GE’s “noble metal water chemistry”) deposited on the surface of the piping However, only out of 104 tests ignited in a mixture of hydrogen and oxygen contained in a closed cylinder into which steam was injected from the side Due to uncertainties, the regulatory body (NISA) instructed the power companies to remove the accumulated gas and water before conducting the monthly ECCS test with a similar design in the 14 units Power companies followed this instruction by isolating the ECCS several days before conducting such a test to cool the system since the steam temperature and pressure is as high as 270 °C, MPa Unfortunately no fundamental solution was developed before the FDA The author however recently revisited this unsolved issue and found that the configuration of isolation valves for the RHRS system in Hamaoka (BWR/4, Mark-1) as shown in Figure A2, as well as Fukushima Unit (BWR/3, Mark-1), is strange The two isolation valves at the PCV boundary are arranged in a series, indicating that these valves should normally be in the open configuration when considering the single failure criteria With the normally open valve, the steam should condense in the RHRS piping which is cold during normal operation, separating the dissolved hydrogen and oxygen These isolation 331 Nuclear Engineering and Design 324 (2017) 315–336 G Saji and have filters for Dry Well venting There should be no rupture disk in the vent line whereas the D/W design pressure is 430kPa(gauge) They could manually start venting at 9:15, only on March 12, since they had to wait for the nearby population to evacuate before being able to begin A series of “spontaneous venting” occurred both in 1F1 and 1F2 This behavior indicates that the flange joint was behaving like a safety valve for a pressure vessel When the effluent leaks from the flanges of the PCV, it means there is a functional failure Recall that an intrinsic safety factor of is incorporated into Section for “design by analysis” of the ASME Boiler and Pressure Vessel Code for structural integrity The PCV should likewise be designed with an intrinsic safety factor of for functional integrity against leakage from the flange joint Without these vulnerabilities, the environmental release could have been at least three orders of magnitude less, combined with TEPCO’s excellent accident management mitigation taken during the active phase of the FDA Such a large reduction in ground shine should have been feasible if the internal hydrogen explosion was prevented which likely occurred in the air volume of suppression pool of 1F2 However, the author has some reservation as to the core melt scenario introduced in Subsection 5.3.1, since the recent interim report of the muonic imaging report states the following conclusion: “The evaluation at present shows the possibility that some fuel debris remain in the core and at the lower area of RPV, but massive and high density material has not been found “ Conclusions and lessons learned Almost all of the investigation reports published to date considered that the gigantic earthquake and tsunami induced an unprecedented beyond design basis accident and therefore the resultant consequences were unavoidable The author does not share this view, since design fortification in consideration of the prevention and protection against design basis events alone is insufficient He believes that the environmental contamination and public exposure could have been substantially mitigated had the vulnerabilities as identified in this forensic engineering study been removed Acknowledgement Much of the contents of this report are extracted from the author’s periodical e-mail updates on the FDA, so far distributed with the following records: • Daily updates, starting March 12 with Earthquake (1) ∼ July 25 with Earthquake (135) • Twice a week updates, starting July 26 with Earthquake (136) ∼ October 14 with Earthquake (160, Oct 11–14) • Once a week updates, starting October 22 with Earthquake (161, Oct 14-21) ∼ July 27, 2012 with Earthquake (203, July 20–27, 2012) • Biweekly, starting August 10 with Earthquake (204, July 27 - Aug 10, 2012) to March 15 with Earthquake (248, Mar 21 - Apr 4, 2014) • Once a month, starting April 1, 2014 with Earthquake (249, Apr 30, 2014) ∼ Earthquake (288, July 1–31, 2017) The Earthquake series is intended to provide scientific minutes of the Fukushima Daiichi nuclear disaster The minutes have been distributed widely in English to the author’s international colleagues Some of them kindly provided comments during the reviewing process of this paper Appreciation is directed to Ms Dana Pandolfi for her effort of technical editing, although the entire context is the author’s responsibility (1) The threat of hydrogen generation through “radiation-induced electrolysis” especially in BWRs (2) Potential threat of an “internal hydrogen explosion” in the suppression pools The cover gas of the suppression pool water should have been nitrogen (3) The potential threat of an “internal hydrogen explosion” in pipes where the steam condensation and accumulation of hydrogen and oxygen gases might occur as in the case of the Hamaoka Unit accident This might be limited to those plants with similar PCV design (4) Leak rate of the PCV should have been testable at its design basis pressure The intrinsic safety factor of the containment flanges against overpressure effluent leakage should have been for the functional integrity of the PCV (5) Spread of the hydrogen gas from the vent lines through duct works connected to SGTS The hardened vent line should be independent Annex A: Current decay heat and “melt-down” scenario This Annex is developed to clarify whether the corium (the lava-like mixture of fissile material created during the nuclear accident) has relocated down to the bottom of the Primary Containment Vessel (PCV) by melting through the bottom head of the reactor pressure vessel (RPV) Recent representative temperature data inside both the RPV and PCV are available from TEPCO’s Web Site (TEPCO, 2011b) By referring to the temperatures measured on 2017/06/26, Table A-1 indicates that both the RPV and PCV are stabilized thanks to adequate water flow rates of water injection into the RPVs If the “melt-through” of the RPV bottom heads had occurred, most of the injected water should flow down to the PCV, which would have eventually resulted in overflowing; however such events have not been reported by TEPCO In addition, the relocation of the corium should indicate that the large decay heat of the corium should exist at the bottom of PCVs In the above mentioned handout, the following decay heat estimated by TEPCO were included The author verified TEPCO’s results by referring to previous fuel management records to incorporate the effects of long-lived nuclear species and operation cycle durations accumulated in aged fuel The results indicate that TEPCO substantially under estimated the results as summarized in Table A-2 This comparison indicates that TEPCO’s decay heat was likely estimated by using some reference decay curve heat (e.g ANSI/ANS 5.1) multiplied by the nominal thermal power of each unit The estimated decay power is not consistent with the “melt down” scenario Instead it more likely indicates that “in-vessel retention of core debris” was likely achieved, thanks to the commendable severe accident management efforts during the active phase of the Fukushima Daiichi accident Table A-1 Current representative water temperature of the PCV and RPV Units 1F1 1F2 1F3 RPV (°C) PCV (°C) 22.1 ∼ 22.3 22.1 ∼ 29.5 28.1 ∼ 28.5 28.8 ∼ 28.7 23.7 ∼ 25.7 23.4 ∼ 25.6 332 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Table A-2 Verification of TEPCO’s decay heat Units 1F1 1F2 1F3 Decay Power (TEPCO-2016) Saji (2000 days) 70 kW 153 kW 90 kW 229 kW 90 kW 127 kW Annex B: Overview of hydrogen generation mechanisms B.1 Brief overview of high temperature zirconium-steam reaction The root-cause for a series of hydrogen explosions is one of the least known although almost all of the investigation reports explain the accident scenario as a tsunami-induced SBO (station blackout, loss of all AC power) which resulted in the core “melt-down” accidents Through the core melt a high temperature Zr-steam reaction generated a large amount of hydrogen gas, which leaked into the Reactor Building and then exploded However, TEPCO has not been able to identify any evidence demonstrating the existence of large heat generating corium (molten fuel debris) despite their robotic investigation of the reactor pressure vessel inside the containment primary vessel even six years after the accident It has been widely explained that the hydrogen gas is initially generated by the rapid oxidation of the zirconium alloy tubes (“fuel cladding”) that surround the sintered uranium-dioxide fuel pellets in commercial reactors Some scientists insist that the zirconium cladding in a water or steam environment would undergo a rapid and self-sustaining heat generating (exothermic) oxidation reaction Although it is an exothermic reaction, such a self-sustaining reaction can be excluded by referring to the experiment performed as early as in 1954 by W A Bostrom (Bostrom, 1954) During the course of his investigation several samples were heated above the melting point The sample consisted of a flat disk 3/4″ in diameter and 1/2″ thick A carbon sample holder was used and the entire assembly was placed about 6″ under water This sample was held above the melting point for about 10 s The reaction proceeded quite rapidly but not with great violence, indicating that the oxidation of Zircaloy-2 with water is not self-sustaining for the specimen dimensions employed even at temperatures somewhat above the melting point Several other samples were melted under water inadvertently, and it was observed that although the reaction proceeded more rapidly, it did not become violent or autocatalytic in nature It has also been observed during other experiments and also during arc-melting that molten zirconium dropping into water does not react violently Since the time of these early studies, a number of investigations have been performed as reviewed by L Baker, Jr (Baker, Conf-830816) The most significant point of the succeeding studies is that the kinetics of the high temperature steam reaction of Zircaloy alloys, uranium and austenitic stainless steels have been found to be consistent with parabolic law behavior often observed in such experiments This phenomenon is induced by quickly forming the protective oxide films on the surface of metals, retarding the corrosion phenomena The rate of hydrogen generation from zirconium-steam reaction is not so violent as widely contemplated for the Fukushima accident The conceptual chemical reaction channel assumed in these hypotheses is through Zr + H2O → ZrO2 + 2H2 However, it is questionable to anticipate the water or dissolved hydrogen molecules which diffuse through the protective oxide film covering the zirconium surface in case of the Fukushima accident since the fuel rods were almost always covered with the primary water, except for a few hours when the water injection was interrupted B.2 “Radiation-induced electrolysis” (RIE) Since the scientific cause for a series of hydrogen explosions has not been established, the author investigated his basic theory named “radiation- Fig B-1 Configuration of Fukushima Daiichi Unit 1–3 333 Nuclear Engineering and Design 324 (2017) 315–336 G Saji induced electrolysis (RIE)” by applying the estimation of the amounts of H2 generation during the active phase of the Fukushima accident The author’s theory was originally developed by including Faraday’s Law of electrolysis into the basic time-dependent material balance equation of radiation-chemical species for his study on accelerated corrosion phenomena As such this theory applies to the early phase of the accident before the loss of water levels in the reactor cores, although the simulations were performed from the time of the seismic reactor trip to the hydrogen explosions The RIE configuration assumed in estimating the hydrogen generation in the primary water is illustrated in Fig B-1 In such an open configuration of RPV, the generation H2 is not suppressed as pointed out in the Spinks and Wood’s textbook (Spinks and Woods, 1990) Thus accumulated hydrogen and air mixture in the free air volume of the suppression pool resulted in the explosion, since there were no direct charge line for nitrogen gas from its nitrogen gas generator to the suppression pool Through this mechanism as much as 29,400 m3-STP of hydrogen gas is estimated to be accumulated inside the PCV just prior to the hydrogen explosion which occurred one day after the reactor trip in 1F1 With this large volume of hydrogen gas and air the explosion was a viable possibility upon the “venting” operation In view of this observation, hydrogen generation from the spent fuel pools was also estimated by applying the RIE mechanism With a mix of different levels of radioactivity of spent fuel, a variance in the absorbed dose rate of water through γ-decay heat should have existed This configuration induced an electrochemical potential difference between the highly radioactive region where there was spent fuel stored by evacuating the core and less radioactive fuels stored for several years The spent fuel was stored in racks placed at the bottom of the pool where the wall was covered with a stainless steel lining The metallic contacts enabled electric conduction between the highly radioactive fuel assemblies and the cooled spent fuel The contemplated RIE mechanism is illustrated in Fig B-2 The author searched for a potential radiation chemical mechanism for the hydrogen explosion in Unit of the Fukushima Daiichi during the accident by changing the pool water temperature and flow velocity in the spent fuel During the trial calculations SBO was found to have induced a rapid initiation of electrolysis when the pool water temperature reached approximately 40 °C.(Figs C-1 and C-2) The present estimation of the hydrogen generation rate is still large enough to have induced the explosion in 1F4 SFP since as large as 1000 Nm3/ d is estimated when the pool water temperature exceeds approximately 40 °C It also revealed that the behavior of the radiation chemical process is much more complicated than simply the dependence on temperature It depends on the management of spent fuels in the SFP, absorbed dose rate and volume of irradiated- and mixing volume of water as well as its flow velocity (i.e residence time of the water staying in the highly active region of spent fuel) In order to abide to these complexities the author proposes the simple solution of inserting a ceramic insulator to prevent direct metallic contact of the spent fuel racks to the SFP liner thereby disconnecting the flow of electrons from the anodic cooled fuel assemblies However the author has reservations regarding the current results as our knowledge is extremely limited as to the chemical characteristics of the cooling water especially in the core region In particular the application of the author’s basic approach for his RIE corrosion study to the severe accident situation has been established essentially without experimental data to verify However the observed phenomenology of a series of hydrogen explosions during the Fukushima accident is not contradictory to the author’s prediction Obviously understanding the phenomena occurring through radiation-induced electrolysis in the Fukushima accident is not complete, and the theoretical framework of radiation chemistry applicable to severe accidents in LWRs has to be more firmly established Fig B-2 Radiation-induced electrolysis in 1F4 SFP 334 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Appendix C Annex C: A fact sheet of Hamaoka Unit accident This appendix summarizes the hydrogen-explosion/pipe-rupture accident that occurred at the Hamaoka Unit as shown in Fig C-1 The pipe rupture occurred at a section of the ECCS (at the top portion of the steam piping from the RPV) in which the steam was condensed separating the hydrogen and oxygen as illustrated in Fig C-2 Outline of the Hamaoka Unit accident • Rated power: 540 MWe, BWR • Occurred: 17:02 of November 7, 2001 • Plant State: Steady state operation, during a scheduled ECCS test for manual startup of HPCI pump • Water Chemistry: Hydrogen water chemistry, supplemented with the noble metal™ water chemistry Detection of the accident • Operators’ observation: Explosive noise observed by the operators of the central control room as well as at local panels • Trip sequence: HPCI trip followed with a containment isolation resulting in the automatic closure of the isolation valves in approximately 30 s • Environmental monitoring: 30–40 nGy/h (no change) • Confirmation: A maintenance crew accidentally went inside of the reactor building, at 17:20 and identified that the 1F and 2F floors were wet with water It was concluded that the fire alarms were triggered due to steam leakage Fig C-1 Pipe Rupture at the Steam Condensate System of RHRS (Pipe size: OD = 165 m/m, WT = 11 m/m) Fig C-2 Schematic diagram of RHRS Line B The pipe rupture occurred at a section of the ECCS (at the top portion of the steam piping from the RPV) in which the steam was condensed separating the hydrogen and oxygen 335 Nuclear Engineering and Design 324 (2017) 315–336 G Saji Post-accident inspection • Estimated leakage: approximately tons of steam The estimated radioactive leakage was 8ì10 Bq ã Causes of the explosion: Hydrogen burn induced from hydrogen gas accumulated at the top portion of the Steam Condensate piping system The hydrogen was transported with the steam which condensed in the cold pipe at the top thereby separating the hydrogen gas • Hydrogen concentration: 0.6% volume in the similar riser pipe location of RHRS A and 19% of O gas Similarly in Unit RHRS A, H = 46%, O = 23%: RHRS B, H = 27%, O = 23% • Estimated hydrogen accumulation after months of operation: 6–8 meter from the condensed water surface, with H = 66% and O = 33%, 2 2 2 N2 = 1% Follow-up test results • Cause of ignition: It is suspected to have been induced by minuscule particles of noble metals deposited on the surface of the piping However, only out of 104 tests ignited in a mixture of hydrogen and oxygen contained in a cylinder into which steam was injected from the side • Ignition tests: Self-ignition at 340–370 °C, 5–8 MP with dry mixture of hydrogen and oxygen No self-ignition with steam Some cases of selfignition with a noble-metal catalysis • Combustion to Detonation transition: At 1–2 m from the ignition point • Maximum plastic deformation of pipes: greater than 23% • Brunsbüttel Accident: A similar pipe rupture incident that occurred on December 14, 2001 at the Brunsbüttel l NPP in Germany In this plant, the noble metal chemistry is not used Instruction by the regulator (NISA) • Remove accumulated gas and water before conducting the monthly ECCS test with a similar design in the 14 units • Power companies followed this instruction by isolating the ECCS several days before conducting such a test to cool the system since the steam temperature is as high as 270 °C, MPa No drastic solution developed before the Fukushima disaster Saji, G., 2016 Root cause study on hydrogen generation and explosion through radiationinduced electrolysis in the Fukushima Daiichi accident Nucl Eng Des doi http://dx doi.org/10.1016/j.nucengdes.2016.01.039 Saji, G., 2017 Radiation-induced electrolytic phenomena with differential radiation cell in water-cooled nuclear reactors Nucl Eng Des Doi information: http://dx.doi.org/ 10.1016/j.nucengdes.2017.03.022 Spinks, J.W.T., Woods, R.J., 1990 An Introduction to Radiation Chemistry-third ed John Wiley & Sons Talaat, M.M., 2014 Overpressure fragility evaluation of a Mark I drywell using thermalmechanical finite element analysis In: Proceedings of the ASME 2014 Pressure Vessels & Piping Conference PVP2014 July 20–24, 2014, Anaheim, California, USA TEPCO, 2011 Results of Analysis of Accumulated Water in the Turbine Building http:// www.tepco.co.jp/nu/fukushima-np/images/handouts_110522_04-j.pdf (in Japanese) TEPCO, 2011 Archive (monitoring data update for 2011/2/11-2011/3/21 in Japanese) http://www.tepco.co.jp/nu/fukushima-np/f1/images/f1-20110311_0321.zip TEPCO, 2012 Investigation Report of the Fukushima Nuclear Accidents, 2012 http:// www.tepco.co.jp/en/press/corp-com/release/2012/1205638_1870.html TEPCO Handout, 2013 Results of Investigation around Lower Parts of Unit Vent Pipes at Fukushima Daiichi NPS (Second Day) http://www.tepco.co.jp/en/nu/fukushimanp/handouts/2013/images/handouts_131114_05-e.pdf TEPCO Handout, 2013 (in Japanese) http://www.tepco.co.jp/en/nu/fukushima-np/ handouts/2013/images/handouts_131113_15-j.pdf http://www.pbadupws.nrc.gov/ docs/ML0230/ML023020246.pdf TEPCO, 2015 https://www4.tepco.co.jp/cc/press/betu15_j/images/150119j0201.pdf TEPCO, 2016 Locating fuel debris inside the unit reactor using a muon measurement technology at Fukushima Daiichi Nuclear Power Station http://www.tepco.co.jp/ en/nu/fukushima-np/handout/2016/images/handouts_160728_01-e.pdf TEPCO, 2017 Unit Primary Containment Vessel Investigation at Fukushima Daiichi Nuclear Power Station (Investigation results by the self-propelled investigation device) http://www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/ handouts_170216_01-e.pdf TEPCO, 2017 Reduction of the flow rate of the water injection in Unit 1∼3 http://www tepco.co.jp/nu/fukushima-np/handouts/2017/images1/handouts_170522_05-j.pdf TEPCO, 2017 Progress of Unit PCV internal investigation (Preliminary report of July 21 investigation) (PDF 351KB) http://www.tepco.co.jp/en/nu/fukushima-np/ handouts/2017/images/handouts_170722_01-e.pdf Terwel, K., et al., 2012 An initial survey of forensic engineering practices in some European countries and the USA In: ASCE 6th Congress of Forensic Engineering, November 1-3, 2012, San Francisco, California, USA Toshiba, 2012 Issues in I & C system in Unit and during the Fukushima Daiichi Nuclear Power Station http://www.aesj.or.jp/~safety/H240810seminorsiryou2.pdf Torii, T., 2013 Enhanced analysis methods to derive the spatial distribution of 131I deposition on the 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(DID)” is the basic approach for prevention of the occurrence of DBEs, as well as for protection from the further evolution of the events and mitigation of their consequences The concept of the DID... conditions including prevention of accident progression and mitigation of the consequences of a severe accident Level 5: Mitigation of the radiological consequences of significant external releases of. .. a large release of 131I to the SWS direction of the Fukushima Daiichi site The potential flow of the plume with the high concentration of 131I in a southern direction from the Daiichi is a new