Peer Review: Email about roadmap and accident details at Unit 1

From: Genn Saji
Subject: Earthquake (170, Dec 16-23)
Date: Sun, 25 Dec 2011 13:03:34 +0900

Dear Colleagues:

280th – 287th day!

I. “Mid- to Longterm Roadmap for 1F1 to 1F4 towards Decommissioning”
On December 21, the Committee on Countermeasures for Mid- to Longterm Issues, which was jointly organized within the Government and TEPCO, approved the “Mid- to Longterm Roadmap for 1F1 to 1F4 towards Decommissioning”.  The new roadmap was jointly drafted by TEPCO, Agency for Natural Resources and Energy and NISA.  It is released only in Japanese:
Main text: http://www.meti.go.jp/earthquake/nuclear/pdf/111221_01b.pdf
R&D: http://www.meti.go.jp/earthquake/nuclear/pdf/111221_01f.pdf
With the declaration of completion of Step 2 of Roadmap on December 16, a new project framework became necessary after achieving the “cold-shutdown” status.  In the new Mid- to Long-term Roadmap, a completion of decommissioning is scheduled as long as 30 to 40 years, where the following three phases are projected.
Phase 1 (<2 years): Until initiation of transfer of spent fuels from the SFPs.
Phase 2 (<10 years): Until initiation of retrieval of fuel debris
Phase 3 (30-40 years): Until completion of decommissioning
In view of uncertainties of the actual core status, HP’s “(Project)  judgement points” are newly identified before going into next major tasks.  The followings are some of the HPs.
HP 1-1:  Successfully closed leakage paths from the reactor building to the turbine building.  Completion of closing leakage paths by repairing the PCVs.
HP2-1:    Decision on methods for reprocessing and storing of damaged fuels.
HP3-1:    Decision of ways of closing leakage paths from the reactor building to the turbine building.  Decision of approaches for repairing the PCVs.
HP 3-2: Successfully filling water in the PCV, achieving “water sarcophagus” status of the lower portion of PCV.  Project decision for viewing methods inside PCV.
HP 3-3: Determination of ways of repairing the upper portion of PCV.
HP 3-4: Completion of filling PCV with water.  Decision of in-vessel viewing.
HP 3-5: Establishment of methodology of retrieval of fuel debris
HP 3-6: Decision of processing and disposal of fuel debris
As can be seen from these major HP’s, the whole project planning is very complex and confusing, since  an activity is switched to one of the plural next tasks, depending on whether the initial approach will be successful or not.  Honestly speaking, I do not understand their approach, since the follow-on tasks will depend much on actual cores debris status in the RPV and PCV.
From this point of view, it is strange to find that there is no HP specified at the end on Phase 1, during which “Partial Observation inside PCV” is projected.  The overall project planning will undoubtedly be influenced greatly, depending on whether the claimed melt-through into the PCV has really occurred or not.  It is disappointing to find no promising approach found for better characterization of status inside of PCVs.  I believe, the current “Gas Treatment System” installed in 1F1 and 1F2 are showing precious data of contamination levels inside the damaged PCVs through their sampling lines.  The observed low contamination levels in the PCV atmosphere and the moderately high radiation monitoring data of CAMS (Containment Atmosphere Monitoring Station)  readings do not seem to be consistent, not supporting TEPCO’s gross melt-through scenario. It is very disappointing to find that there is no planning to restore radiation monitoring inside PCVs.  Currently, in 1F1, both CAMS D/W(A) and CAMS D/W(B) are designated as out of order, showing 1.00E-02 Sv/h and 1.20E+01 Sv/h respectively.  The CAMS data of Fukushima Diichi are more strongly indicating that the contamination level inside PCV is due to condensation of the suppression pool water, which contained cesium but not due to core debris buildup.  A few Sv/h level of dose rate can easily occur at the hot spots with steam leakage inside of the Reactor Building as shown in the recent dose rate maps released (only in Japanese) from TEPCO.
http://www.tepco.co.jp/nu/fukushima-np/images/handouts_111203_02-j.pdf
It is also disappointing to find that it will take until 2020 to develop a rotational mirror scope for inspection inside of the PCV. Obviously, one of the largest objectives of the Phase 1 should be characterization of the the extent of core damage for better follow-on project planning, I believe.
Apart from these technical issues, the Mid- and Longterm Plan include Chapter 7, titled Collaboration with International Communities.  In this chapter, a typical bureaucrats’ composition is included  in just 10 lines, with just one long sentence for the future planning.  “We will try to enhance collaboration with various research institutions of the foreign governments and private corporations for practical use of knowledges and experiences pertaining to countermeasures against accidents, in view of necessities of efficient and effective implementation of research and development towards decommissioning which is expected to take a longterm and in a large scale.”
After nine months, transparency for the international communities have decreased substantially, in my observation. Mr. Edano, Minister of METI, and Mr. Hosono, Minister in Charge of Fukushima Disaster, instructed that “an investigation for establishment of a new research institution for future dissemination of technologies obtained through implementation of the Mid- and Longterm Roadmap by integrating national and international wisdom. However, this defect is deeply rooted in the Japanese society and no immediate solution is in sight, I am afraid.  This is one of the reasons that I feel I have to continue this update best I can.

II. Effects of aging in the 1F1
There exists a growing concern that some aging phenomena may have induced the Fukushima disaster, especially the earthquake which preceded the tsunami, especially in 1F1 which was commissioned in March 1971.  The plant has just started to run into a life extension beyond the initial design life of 40 years.  In order to respond to this concern, TEPCO released a handout on December 15, their assessment of effects of aging in the Fukushima accident.  The 40 page handout in Japanese is a short overview of the aging effects, consisting of (1) Preservation and maintenance activities towards aging management, (2) Relevance of aging phenomena during the early phase of the disaster.
According to the overview, TEPCO’s aging management is integrated in their overall maintenance strategies, in which special view points are incorporated for providing countermeasures against aging phenomena.  Their activities are implemented through the PSR (Periodical Safety Review) conducted in every ten years and PLM (Plant Life Management) of every 10 years.  Before going into the operation beyond 40 years, TEPCO replaced the following components around the RPV (Reactor Pressure Vessel) in line with the PLM. Reactor Water Recirculation Piping, In-vessel components, tube bundles of Isolation Condensers, Intermediate Cooler of Steam Ejector, Piping of CRDM Drive Unit, Steam Regeneration Heater, Core Spray Pump Motors and Casing Cover of Reactor Water Re-circulation Pump.  Due to the widespread stress corrosion phenomena among BWRs, most of the in-vessel components have been replaced, such as Core Shroud, Upper Reactor Grid Plate, Core Support Plate, Jet Pumps, Feed Water Sparger, Core Spray Sparger and In-Vessel pipings.  However, the BWR lower plenum structure, including Control Rod Guide Tubes, Stub Tubes, Guide Tubes and Instrument Tubes have not been replaced at that time.
As pointed out by Dr. S.A. Hodges, BWRs incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of PWR accident sequences. The effect of the BWR procedural and structural
differences upon the progression of a severe accident sequence during the period preceding movement of core debris into the reactor vessel lower
plenum has been discussed such as in:
http://www.osti.gov/bridge/servlets/purl/6343189-Nf9EJg/6343189.pdf
http://www.osti.gov/bridge/servlets/purl/6124656-R8y05j/6124656.pdf
http://www.ornl.gov/info/reports/1989/3445606042920.pdf
How a potential failure of the BWR lower plenum structure affected the accident sequence of Fukushima disaster is not known at all.  I suspect these components have never been volumetrically (with ultrasonic or X-ray) inspected since commissioning.  I suspect it is not known to what extent these lower vessel components have degraded by aging.
Going back to TEPCO’s report on aging, their aging management technology integrate seismic safety assessment with component degradation assessment.  The latter assessment is essentially adaptation of US NRC’s GALL, where the degradation mechanisms are identified from generic mechanisms (e.g., SCC, FAC/wall-thinning, general corrosion, corrosion fatigues, wear, high-cycle fatigue, thermal aging and degradation) as well as operational experiences. They concluded that their aging management technology is well matured and can be continue operated with adequate maintenance strategies.
Especially on the effects of aging on seismic motion of the Northern Japan Great Earthquake (2011), they concluded that there were no direct effects, although further detailed study is necessary to quantify a change in seismic responses for low cycle fatigue, irradiation-induced SCC and irradiation embrittlement.  However, even for these cases, no appreciable change has been identified after their detailed study on identified critical components.
I think it is not likely that the seismic oscillation induced failures of components, however, it is not known whether the aging induced additional difficulties for mitigation and amelioration during the progress of the severe accident.  It is because a required strength during the severe accident is completely different from what has been contemplated within design bases.  For example, it is likely some kind of failure should have occurred in 1F1 around the midnight of March 12, when high radiation dose rates were observed both at PCV as well as in the turbine hall.  However it is not likely the failure is induced by seismic force.  Also there were ten large releases detected by temporary radiation dose monitors.  However, several of them are not well co-related with neither venting nor hydrogen explosion of 1F1 to 1F4.  I am suspecting the failure due to the “internal hydrogen explosion.”
The Japanese approach towards aging is adapted from US NRC’s NUREG-1801“Generic Aging Lessons Learned Report “, dubbed as GALL report. http://ocw.mit.edu/courses/nuclear-engineering/22-39-integration-of-reactor-design-operations-and-safety-fall-2006/readings/nureg_1801_v1.pdf
The GALL approach was originally developed to cope with licensing renewal needs, due to the significant slowdown of construction of new plants induced by both TMI and Chernobyl accidents.  It is necessary to satisfy 10 CFR Part 54“Requirements for Renewal of Operating Licenses for Nuclear Power Plants”in the US.  The Japanese approach generally follows IAEA I-GALL(InternationalGenericAgingLessonsLearned), promoted by NISA.         Although degradation phenomena in LWR components have started observed more than 50 years ago, the exact mechanism of degradation has not been well identified.  In view of this, the GALL approaches are based on degradation phenomena, collected through operational experiences.  However, in the corrosion degradation, there exist “incubation period” in which degradation phenomena can be suddenly actualized after a long latent period.  A typical example is degradation of the reactor core shrouds as well as the recirculation piping, which have been replaced with “corrosion proof” 316L stainless steel, occurring in many BWRs.
This issue dates back to the US Navel Reactor Program headed by Admiral Rickover, in my view.  In the Corrosion and Wear Handbook (1957),  he left the following preface: “Even with the large backlog of experience with water systems, there is still a great deal that must be learned.  I hope that some readers of this handbook will, starting from the data herein, work toward a firmer understanding of these problems.”  (Preface of Corrosion and Wear Handbook for Water-Cooled Reactors, Naval Reactors Handbooks, D.J. DePaul Ed., March 1957, in the historical report at the time of Atoms for Peace. (This address was given by Dwight D. Eisenhower before the General Assembly of the United Nations on Peaceful Uses of Atomic Energy, New York City, December 8, 1953.  Some of the previous military secrets were disclosed to world scientific community as a gift of United States of America for peaceful use of atomic energy.)  The corrosion degradation issues have been considered resolved by development of new materials and optimization of water chemistry, however, the issue has been far from solved issue.  I have been investigating the fundamental corrosion mechanism, by going back to the time of Adm. Rickover’s preface, for last ten years, and concluded that the “macro-cell” (I am calling “Long-cell”, in honor of R. Pope) corrosion mechanism is responsible for the diverse corrosion phenomena being experienced in LWRs.  Let me attach the White Paper I wrote on this issues.

III. Update of TEPCO’s Internal investigation committee report on Fukushima Disaster
I introduced this subject in Earthquake (168, Dec 2 – 9).  On December 22, TEPCO released an update of the investigation committee report in 122 pages in Japanese athttp://www.tepco.co.jp/cc/press/betu11_j/images/111222p.pdf.  I intend to review this update in detail in my next mail.  Let me introduce some of the new findings that may clarify core status of 1F1.  It appears that IC (Isolation Condenser) appears to have worked much better than what has been explained by TEPCO.  There is an operator comment pertaining to IC operation performed at late March 11, saying that he confirmed an operation manual description; “IC is capable to operate without feed water into the shell side (secondary side)  as long as 10 hours” in an event of loss of ultimate heat sink.  This operational mode is equivalent to the secondary side “feed and bleed” operation in PWRs.  By combining with the primary side “feed and bleed” operation, which mechanically releases the steam from the PCV into the SC (Suppression Chamber) by sacrificing the primary water inventory, it is more likely that PCV has been cooled as long as a day after arrival of the tsunami.
At 21:30 of March 11:30, he re-opened both MO-2A and MO-3A valves to continue operation of IC-A, judging that there is a water inventory in the secondary side of IC-A. After opening both MO-2A and 3A valves, he heard a sound of steam generated from the IC-A.  He worried that IC may never be restarted once it went down by closing the valves.  He also confirmed that IC-A is working by observing a steam release from the 1F1 Reactor Building.  A member of Electric Generation Team also confirmed the steam from outside of the Essential Aseismic Building.  Later, two operator team tried to confirm the water levels of shell side of IC as well as RPV by walking into the Reactor Building, however, the operator went inside of the building had to retreat immediately by the high dose rate detected by Alarmed Pocket Dosimeter.
I think this operator’s action may likely prevented the gross melt-through in 1F1, by making the essential judgement.  These descriptions summarized above two paragraphs are newly added in the TEPCO’s report.
Prior to this, at 18:18 on March 11, it was found that both MO-2A and MO-3A valves were indicating in the closed position shown in the casually recovered Control Panel.  He suspected that the valves were automatically closed by an isolation signal provided to protect for a pipe break accident in the IC-A line.  However, the steam release from IC decreased soon and at 18:25, he closed the MO-3A valve, worrying about a possible loss of the shell side water in IC-A, as well as no alternative water injection line had been configured yet.
At 16:44 of March 11, the water level inside of RPV was found kept at TAF+250 mm, and again at 21:19 of March 11, at TAF+200 mm, leading the headquarter to judge that IC-A had been working.  These readings indicate that the rector core was still submerged, although the level indication became unavailable at 17:25 on March 11.  (The previous report was indicating TAF+550 mm at 22:00 on March 11. )
The first injection of pure water started at 03:00 on March 12, by injecting only 1.3 m3.  By 14:53, approximately 80 tons of pure water has been injected.  The sea water injection was not became ready until after the hydrogen explosion, which occurred at 15:36 on March 12.  In my previous “back of an envelope calculation”, I estimated as long as 8 hours to reach TAF after the reactor trip. However, if the IC-A continued working after 21:30 of March 11 with “feed and bleed” cooling capability, there is a possibility that the gross melt-through in 1F1 could have been avoided, although a gross leakage of fission gas from the fuel pins may have occured.  As a matter of fact, the water level was found remained at 65% in IC-A. This indicates that it was working even after arrival of the tsunami. since water level should have been kept at 85% after the reactor trip with AC electric power supplied by DGs.
Further investigation is requested, since the effect of “feed and bleed” cooling is essential in clarifying the core status of 1F1.

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