Fukushima Japan Earthquake: nuclear plants Fukushima part 2

[Hiddencamper]

Very interesting detail and illuminating as to Japanese policy. Sort of an all or nothing approach.
Against that, Jim Hardy's reference ( http://www.nrc.gov/docs/ML1012/ML101270372.pdf ) concludes with this punch line:

'if the operators would be able to open both pressurizer relief valves after the core heatup
starts, this would have positive effect on further progression of the severe accident.'

To me, that suggests the Japanese policy may possibly have helped make matters worse than they could have been.Separately, I can only say 'Amen' to hiddencampers sharp reminder that manuals and procedures exist for good reason. They exist to guide operators in real life.

Fortunately we here at PF have the opportunity to speculate more freely. Also, in the case of Fukushima, the results were so poor that we must reexamine whether there was any possibility of some action that would have lesser consequences.

So the issue facing the operators, paraphrased, is how to set the reactors to melt down as gracefully as possible, knowing there is only a few hours of battery power to monitor the process.

Jim's link is for pwr plants. It involves not depressurization during station blackout which can result in not having sufficient inventory for longer coping times, vs depressirizing and wasting inventory early but having the accumulators available.

Bwrs have a very different response. One major difference is that bwrs have an absolutely massive amount of steam relief capacity compared to PWRs, allowing a rapid emergency depressurization which also provides steam cooling to the core.

For a BWR, the safest place to be is with a depressurized flooded core. The challenge is even if you performed an emergency blowdown the moment the earthquake was over, on the loss of DC power the relief valves would have shut and the core would have repressurized. Additionally blowing down the core early removes IC capability, so there was no way to really say those actions could have been taken.

In order for things to be "graceful" the operators needed to A: recognize they had no valid level indication and make a transition to the core flooding EOP, B: quickly got batteries from cars to open up relief valves using car batteries, and C: lined up a portable or fire pump for injection. Even if the fire pump could not be lined up, just having the core repressurized will minimize the potential for containment failure.

The other issue is the diagnosis for entering the core flooding contingency is that you need to observe reference leg boiling. The operators should not simply enter because of a momentary loss of indication. However in this case the reference legs boiled by the time operators got indications back, so they never saw the transition.

This is ultimately one of the things that made daiichi and Daini different. With no DC power at daiichi the operators couldn't even make decisions in the EOPs, and could not take the proper or best steps to protect the core until it was too late.

Station blackout analysis for bwrs assumes you stay hot and pressurized, unlike a pwr.

 

[etudiant]

Thank you very much, hiddencamper, for this more detailed explanation. It certainly clarifies the gap between the BWR and PWR emergency procedures.
Two points that leave me still confused.
First, you note the operators need to go to core flooding EOP, but later say ...just having the core repressurized will minimize the potential for containment failure' .
Is that a typo and should be 'depressurized'?
Second is the discussion about the 'reference leg boiling'. Afaik, this is a topic that has not been explored in this thread. It seems a crucial indication, yet your comment suggests it must be observed, because there is no telltale to inform the operators of a crucial change in the reactor status. Is that correct?
Also, is there an overview discussion somewhere that would help inform the forum members such as myself to whom this aspect of reactor management is an unknown?

The final note, that 'BWR station blackout analysis assumes the reactor stays hot and pressurized' seems diametrically opposite to the core flooding EOP.
How do the operators decide which procedure is the proper one to follow?

 

[Hiddencamper]

Thank you very much, hiddencamper, for this more detailed explanation. It certainly clarifies the gap between the BWR and PWR emergency procedures.
Two points that leave me still confused.
First, you note the operators need to go to core flooding EOP, but later say ...just having the core repressurized will minimize the potential for containment failure' .
Is that a typo and should be 'depressurized'?
Second is the discussion about the 'reference leg boiling'. Afaik, this is a topic that has not been explored in this thread. It seems a crucial indication, yet your comment suggests it must be observed, because there is no telltale to inform the operators of a crucial change in the reactor status. Is that correct?
Also, is there an overview discussion somewhere that would help inform the forum members such as myself to whom this aspect of reactor management is an unknown?

The final note, that 'BWR station blackout analysis assumes the reactor stays hot and pressurized' seems diametrically opposite to the core flooding EOP.
How do the operators decide which procedure is the proper one to follow?

It was a typo. I did mean depressurized.

As for reference leg boiling, I read at one point that at unit 1, they got water level indication back at one point and it was high (top of the narrorw range indication), and believed it was real, as they also thought the IC was operating. This was incorrect, as reference leg boiling was in progress at this time causing an erroneously high level indication.

So, in the Reactor Pressure Vessel control EOP, the RC/L (Reactor Control/Level) branch, there is an override that says "If Level Indication is suspect, then exit this leg and enter the core flooding contingency". It also has a reference to the RPV saturation temperature table, which plots RPV pressure, drywell temperature, and saturation point on a plot. Generally, the EOP basis say that you should not say your level indications are suspect simply because you exceeded saturation temperature, however you do need to monitor for boiling. It's typically not appropriate to transition to core flooding simply because you have a momentary loss of level indication for other reasons, especially if you qualitatively knew what level was or what it was doing prior to losing the indication. So with no indications to go off of, and very few "data points", its hard to say the operators SHOULD have transferred to core flooding or not.

As for the core flooding contingency, the whole purpose of it, is that you lose the ability to determine if level is above top of fuel using your level indications, so instead you depressurize the core and flood it until you either achieve the minimum steam cooling pressure (for scram failure cases), or until water is overflowing out the steam lines (typically your suppression pool level will steady out). Then you can assure that you have adequate cooling because you have level at or above the steamlines, which is above the top of the fuel. You are also completely depressurized, so RPV temps will be <= 300 degF and lowering.

For BWR station blackout, that's a short term event (4-8 hours) with a specific progression, which assumes you have one of IC/RCIC/HPCI/HPCS in service. Reference leg boiling is not supposed to happen in the short duration of an SBO, as you are going to get some form of AC power back to at least 1 decay heat removal train within the coping time.

 

[etudiant]

Thank you, hiddencamper, very helpful and informative.
Is it not surprising that something as critical as the reference leg boiling should be so relatively poorly indicated?
Presumably this reflects the difficulty of determining the water level and state inside the reactor, (iirc the Three Mile Island accident also arose because the operators had no clear measure of the reactor water level). Is there any hope of a better sensor to resolve this uncertainty?

 

[Hiddencamper]

Thank you, hiddencamper, very helpful and informative.
Is it not surprising that something as critical as the reference leg boiling should be so relatively poorly indicated?
Presumably this reflects the difficulty of determining the water level and state inside the reactor, (iirc the Three Mile Island accident also arose because the operators had no clear measure of the reactor water level). Is there any hope of a better sensor to resolve this uncertainty?

It's kind of similar to TMI, but not the same. The issue PWRs had, was that there was no direct in vessel water level measurement, only pressurizer level, which can be inaccurate during a loss of subcooling margin with an open relief valve. For BWRs, all your level instruments are directly in vessel so you always know what core water level is (provided there's no reference leg boiling). In both designs, these indications are all vulnerable to reference leg boiling. You can confirm there's no boiling using a combination of qualitative factors and measurements of drywell/containment temperature and RPV pressure.

Another thing you can do, is install a gamma monitoring system which would detect if radiation levels rise due to a loss of water level, however that would be reliant on some form of electrical power as well. So it wouldn't have really helped in the Fukushima Daiichi case. Their in-drywell rad monitors would have been able to identify something wrong if they had power, without the need for a special monitor specifically for level.

The water level, RPV pressure, and drywell/containment temperature indications are all backed up by uninterruptable power supplies off of the station vital DC batteries. On a loss of all AC and DC power (something that wasn't analyzed), you lose all indications, so it doesn't really help. ALL DC power is never assumed to be lost under design basis events.

After 9/11 the US BWRs got portable "measurement devices" as part of the b5b program. You hook alligator clips up to the leads coming from your transmitters and it uses a built in battery to get a voltage measurement, which you can use a table to correlate to actual water level/ temperature/ etc. This would allow you to have indications even if you lost ALL AC and DC power. Japan didn't implement the b5b program or anything similar to it, so they didn't have this equipment available, which is why they had to go to great lengths to scavenge car batteries just to see what was going on.

 

[jim hardy]

Is there any hope of a better sensor to resolve this uncertainty?

i don't know that this principle is better but it's diverse...

there was research in 1980's for a level instrument based on a long tube with thermocouples, some heated and some unheated,
inside the vessel alongside the core. When level drops the heated ones get hotter than the unheated ones because steam doesn't conduct away the heat so well.
So by the differences in temperature one can infer level.

I don't recall whether a practical version ever made it to production.

 

[Hiddencamper]

For reference, this is the RPV control EOP for a generic BWR. This is the "less pretty" version of EOPs. Many plants use KLR services to generate their EOPs and they are much better structured and easier to look at. I'm honestly surprised at how hard it is to find legible copies of BWR EOPs on google.

Anyways, you can see the override down the RCL/1 RPV Level branch that says, IF RPV Level CANNOT be determined Enter EOP 206 (in this case, the RPV flooding EOP).

I know a lot of people are asking what could have been done. This is an example of what BWR operators have to work with. Sadly I cannot find the containment EOPs or some of the contingencies. 2 contingencies are built into this EOP (steam cooling and alternate level control)

http://www.nrc.gov/docs/ML1100/ML110060122.pdf

Edit: Looks like its hope creek. Here are some other EOPs:
RPV Flooding Contingency http://www.nrc.gov/docs/ML1100/ML110060135.pdf
Emergency Blowdown Contingency http://www.nrc.gov/docs/ML1100/ML110060125.pdf
RPV Flooding during ATWS (scram failure) contingency http://www.nrc.gov/docs/ML1100/ML110060126.pdf
Primary Containment EOP http://www.nrc.gov/docs/ML1100/ML110060123.pdf

Still can't find secondary containment or ATWS EOPs

 

[etudiant]

Thank you very much, hiddencamper, this post has made understanding the problem facing the Fukushima operators much easier.

A first reaction is: 'Oh my, these look like system engineering flowcharts, where a lot of the action implications can only be understood by people very well versed in the system.'
Navigating these procedure charts with partial or non functioning instrumentation seems unlikely to produce a good outcome.

It is not obvious either that increased computer systems support would be beneficial. Airbus pioneered the civil use of fly by wire, where all flight controls are operated by the computer based on pilot input, modulated by system overrides when these inputs are beyond the airplane operational limits. In practice, this sometimes creates conflict of the 'what's it doing now' variety, bad in an airplane, worse in a nuclear plant.
In both cases, a reset switch would be nice to have, but seems beyond our current capabilities.

 

[gmax137]

... I don't recall whether a practical version ever made it to production.

Yes, Jim! The Combustion Engineering supplied HJTC (heated junction thermocouple) system is used to monitor liquid level in the reactor vessel upper head. It does not, however, extend down into the vessel downcomer.

 

[jim hardy]

It does not, however, extend down into the vessel downcomer.

That'd be quite a mechanical feat , to put something long and slender(translate flexible) in the downcomer.
I can only imagine how turbulent is the flow there.

The whole core barrel flexes ever so slightly, and that can be detected in excore neutron detector signals by DSP ( which never ceased to amaze me) .

Thanks gmax !

old jim

 

[jim hardy]

Has anyone ever seen a better photograph of this little square ?http://cryptome.org/eyeball/daiichi-npp3/daiichi-photos3.htm 9th one down

at 10X

just curious if i missed something.

old jim

 

[Tom.G]

That's about all I could get out of it. I don't know how much detail will be left when this gets posted, so here is a brief description:

The bright spot seems the center of a larger structure about six times its diameter. To the upper left of the bright spot, there are two or three strong radial elements radiating thru the dark area, and the outer part of the structure has much faint radial detail showing as fainter dark lines

.

 

[Rive]

just curious if i missed something.

old jim

Brings back memories about the big pixel huntings in the first days...

Cryptome with those compressed and resized pictures never was a good site to start with. I'll look up something , but at first sight that spot looks like a simple beam crossing from the upper structure.

Ps.: As it seems I can only link images... Here it is: http://keptarhely.eu/view.php?file=20160817v00k1x3i5.png
http://keptarhely.eu/view.php?file=20160817v00k1x3i5.png

 

[jim hardy]

Brings back memories about the big pixel huntings in the first days...

i was glued to the computer for months.

Your link has somewhat better resolution.

remember how a long lens compresses distance..

as i said just idle curiosity now, one of those loose ends for me. Headbolt tensioner is above and slightly right as best i recall.

old jim

 

[Sotan]

A new topic appeared these days in the reports posted daily by Tepco: the drain sump pit located at the bases of Units 1/2 smoke stack.

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160825_02-j.pdf
(in Japanese)
In this report of Aug 26 they explain that the area around the basis of the smoke stack is still very radioactive - as confirmed by measurements taken from a distance at the end of 2015. The drain sump pit at the basis of the smoke stack needs some investigation to learn more about the water level and radioactivity. They have been preparing for this investigation using mockups, and now they are planning to go for it. (It is not a lot of water, if I understand correctly the sump pit is only about a cubic metre in volume - but probably it is very radioactive and poses a risk if it starts leaking.)
Basically, with a cutting tool manipulated with a crane arm from behind shielding panels, they will cut an opening in the concrete wall/roof of the drain sump pit, check the water inside and start operations for storing that water in a safer place.

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160826_03-j.pdf
(in Japanese)
Next-day report on the start of the operations.
The two photos are available in unusually good resolution here:
http://photo.tepco.co.jp/date/2016/201608-j/160826-01j.html

 

[mheslep]

still very radioactive

I wonder what qualifies as very radioactive in the context of units 1 or 2 and Tesco analysis.

 

[Bandit127]

I don't know what quantifies as "very radioactive" but it surveyed at 10 Sv/h after the accident so it probably still qualifies.

 

[Rive]

Also, those mobile rad-shields on the photos are surely not for just show...

 

[Sotan]

Hm... I checked again and to be honest I must say I inserted the word "very" without it being needed. So easy to get carried away. My apologies, the text doesn't really say "very". A more careful translation would be "the air radiation dose in the area is still high and hinders investigations". As for the radioactivity of the water - if there is any water in the pit - they don't really know anything, but they seem to be making careful preparations for its transfer if they find it there.
Page numbered 3 in the first link gives a few readings of the atmospheric radiation dose in that area at the basis of the smoke stack, most recently in October 2015; highest is 12 mSv/h, measured at 1.2m height.
As Bandit127 mentioned, in other spots, especially in the immediate vicinity of some pipes designated as "SGTS", they measured more than 10 Sv/h in August 2011 and still 2 Sv/h in October 2015. (http://www.tepco.co.jp/nu/fukushima-np/handouts/2015/images/handouts_151026_04-j.pdf - page 16)
This time, the workers operating the remote controlled machines will be using a room located at the basis of the stack, in which the radioactivity is much lower (0.04 mSv/h).

 

[Sotan]

Fukushima nuclear plant prepares for typhoon (NHK news)
http://www3.nhk.or.jp/nhkworld/en/news/20160829_22/

 

[Sotan]

Remember the X-6 penetration to which access is needed for an internal investigation of Unit 2 PCV, but proved difficult to decontaminate to an acceptable level until now?
On Aug 26 this report was published (in Japanese) as part of a new "mid- and long-term roadmap" update:
http://www.tepco.co.jp/nu/fukushima-np/roadmap/images1/images2/d160825_08-j.pdf

Just a few points I picked from it:

-Decontamination work at the entrance of X-6 started in October 2015. They washed and vacuumed and scraped the floor and walls in the area, but the measurements on the floor still indicated values as high as 8 Sv/h (in January). Even if they were to use some form of shielding, it was assumed that doses in the air in the work area would not fall below 100 mSv/h.

- They would like to attain a value of 20 mSv in the air, behind shield - a value considered low enough to allow a worker to spend 5 minutes in there.

- Page 4: initially they were thinking of installing a simple shield based mainly on a 90 mm thick, simple square lead plate. This is called Case 1. As an alternative, they now came up with a more complicated shield design, with multiple lead plates arranged as a container and aiming to provide better blocking of radiation. This is Case 2. These two designs have been analyzed to see what results they can provide.

- Page 5 shows the results of the simulation for Case 1 shielding; the conclusion is "not good enough".

- Page 6 shows that Case 2 shielding could give the desired effect; the higher remaining values (17.3 and 13.2 mSv/h) are due to the poorer shielding in areas that will be occupied by various devices to be used effectively in the later investigation, devices which have less shielding power.
So they are now trying to design and construct that complicated shielding structure, which poses a weight problem (they can only move a 2t weight at a time in the area), so most likely the "shielding container" will have to be made of two separate parts.The advantage is that it can be installed by remote controlled machinery and it can achieve the desired degree of shielding without requiring further decontamination work, which has been a big pain and unsuccessful until now.

- From this report I (think I) learned somethign about the various radioactivity measurement results that appear in Tepco reports. If you look at page 3, you see the highest value given for the floor is 278 mSv/h. On the other hand, Page 10 shows some radioactivity values, measured in June this year, which are rather in the Sv range (exceeding the device maximum scale of 10 Sv in one spot); these are actually measures in June this year. Page 13 explain the mystery: the "lower" values reported on page 3 are measured using a device equiped with a lead "colimator" which has a reduction factor of 1/500... I did then try to read a little about radiation detection and measurement, but it's such a complex subject. Two conclusions I draw, 1) numbers are not everything, their meaning & the method behind them must be well understood, and 2) as you see it is very likely that I make mistakes every so often in my posts, so... I kind of rely on you to correct them when needed.

 

[Sotan]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160905_05-j.pdf
One-page follow up regarding the sump pit located at the bottom of the smoke stack of Unit 1-2.
They made an opening in the concrete walls of the pit.
This is a photo/screenshot of the inside, but it doesn't say much... to me at least.
Just the photo and two arrows, one pointing at a pump, one at some "absorbing sheet".

 

[Sotan]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160909_04-j.pdf
(in Japanese)
They measured the water level in the drain sump pit at the foot of the smoke stack of Units 1/2.
Water level is at ~60 cm.
As mentioned before, the internal dimensions of the sump pit are about 1m x 1m x 1m.

 

[Sotan]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160913_01-j.pdf
(in Japanese)
They removed the first piece of the side walls of the cover of Unit 1 Reactor Building. A few small photos are included.

Based on other reports in the media, the cover walls dismantling will continue until late November. Removing the debris from the operating floor will follow, and if all goes well, they will start to remove the spent fuel from Unit 1 SFP sometime in 2020.

 

[Sotan]

Update of the last post above (too late to edit, it seems):
- Nice little video of the panel removal operation available here:
http://www.tepco.co.jp/tepconews/library/archive-j.html?video_uuid=hz09ww85&catid=61699

 

[Sotan]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_161007_04-j.pdf
(in Japanese)

One page report about Tepco finishing the removal of the top panels of the walls of the cover of Unit 2 building.
It's just the announcement and 4 photos taken from a distance.
The operation ended without any dust-related problems.

The much larger versions of the photos and a 68 Mb video can be found here:
http://photo.tepco.co.jp/date/2016/201610-j/161007-01j.html

 

[etudiant]

Thank you, Sotan, for these updates.
They show that serious work is continuing in Japan, even though it no longer generates headlines in the foreign press.
Removing spent fuel from these reactors SFPs will be challenging, as these structures have experienced serious explosions and contamination. So the 2020 date probably reflects the need to fabricate some support facilities that neutralize those risks. Japan is certainly breaking new ground here and presumably after a few more years the fuel pools at reactors 2 and 3 will also be emptied. Going beyond that to recover the corium looks to be much more challenging.

 

[turi]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_161007_04-j.pdf
(in Japanese)

One page report about Tepco finishing the removal of the top panels of the walls of the cover of Unit 2 building.
[...]

Thank you. But is this really about Unit 2? My almost inexistent Japanese only decripts references to Unit 1.

 

[Sotan]

Of course you're right, turi.
No idea what made me type Unit "2" there
My apologies and thanks for pointing it out.

 

[Sotan]

Tepco is cautiously moving towards reducing the amount of water poured into the damaged reactors:
http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_161019_07-j.pdf
(in Japanese)

Page numbered 2: while at present they are pouring 4.5 m3/h in each reactor, calculations based on the decay heat made around August this year show that 1.4 to 1.8 m3/h should be sufficient.

Page 3: for starters they will gradually cut at most 1.5 m3/h for each reactor (interestingly, such cuts aren't that easy to make and it's not easy to go straight to 2m3/h, for example, as there are various limitations regarding to equipment such as valves and alarm signals etc; this is why they feel they still need to pour 3 m3/h, for now.).

Page 4: they will keep a close eye on the plant parameters, to see the effect of the reduced water injection. They want to keep the temperature in the lower side of RPV and in the PCV under 65 Celsius; they want to check every hour the amount of water actually going in; they will even pay attention to the dust monitors.

Page 5 shows a flowchart they will use in the transition phase. The water flow will be cut in 0.5 m3/h increments and will reverse to the previous value in case that negative changes are observed. The whole transition period, until things settle, might take about a week.

Page 6 shows the expected changes brought about by reducing the water input to 3 m3/h, in view of the time allowance they have in case water flow is stopped.
Reactor 1 RPV bottom is now at 28 degrees Celsius and would take 8.4 hours to reach the limit value of 80 degrees Celsius if water cooling stopped completely. Reactor 2: now at 33 degrees Celsius, would take 8.3h to reach 80 degrees Celsius. Reactor 3: now at 31 degrees Celsius, would take 8.9h to reach 80 degrees Celsius. These are the values at present; if the water flow was cut to 3 m3/h it is presumed that the RPV bottom temperatures would rise by 7-8 degrees Celsius (theoretically; practically it might be a bit less), and if water flow stopped they would have 7.2, 6.9 and 7.3 hours until the RPV would reach 80 degrees Celsius.

Page 8: past data showing how Reactor 1 responded to the change in water flow. A rise in temperature is observed only about 10 hours after the water flow reduction, and it took about 5 days from the manoeuver till the temperatures got stable again (new value, higher by about 8 degrees Celsius).

Page 9: in case of Reactor 2, the effect of reducing the water flow was almost immediate, and stability was reached again after about 4 days, about 10 degrees Celsius higher.

Page 10: in case of Reactor 3, it took about 12 hours to notice a rise in temperature after reducing the water flow, and new temperature stability (about 4 degrees Celsius higher) was reached after about 7 days.

 

[Sotan]

One-page report on Units 1-2 smoke stack inspection using drones
http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_161020_03-j.pdf
(in Japanese)
Today they used a drone to inspect the inside of the stack, only to find a beam or some other kind of construction material stuck inside at 10-20m from the top of the stack. So they gave up the inspection of the interior and will use data - images and radiation measurements - obtained during the exterior inspection (finished earlier) to assess the state of the stack.

 

[Sotan]

Tepco finished removing the wall panels of the cover building of Unit 1.
Several photos and a movie can be seen here:
http://photo.tepco.co.jp/date/2016/201611-j/161110-01j.html

 

[SteveElbows]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_160909_04-j.pdf
(in Japanese)
They measured the water level in the drain sump pit at the foot of the smoke stack of Units 1/2.
Water level is at ~60 cm.
As mentioned before, the internal dimensions of the sump pit are about 1m x 1m x 1m.

The following english pdf has radiation measurements of the water. Since years have gone by since I learned something about how to interpret the scale of such readings, and I've gone really rusty on this front in the meantime, would anybody be so kind as to say something about these numbers and their implication? Many thanks.

http://www.tepco.co.jp/en/nu/fukushima-np/handouts/2016/images/handouts_160913_02-e.pdf

Gross beta: 6.0x10⁷Bq/L
Cs-134: 8.3x10⁶Bq/L
Cs-137: 5.2x10⁷Bq/L

 

[Rive]

The following english pdf has radiation measurements of the water. Since years have gone by since I learned something about how to interpret the scale of such readings, and I've gone really rusty on this front in the meantime, would anybody be so kind as to say something about these numbers and their implication? Many thanks.

Way higher than it was for the water in the turbine building basements right after the accident.
http://www.tepco.co.jp/en/press/corp-com/release/betu11_e/images/110327e15.pdf
Quite serious stuff, even if there is just a small amount of it.
However, I think this concentration implies that there was no leak.

 

[nikkkom]

What are they waiting for with this water? Why not pump it into a metal box, add some cement powder and make it a solid?

 

[etudiant]

Surprised there is so much Cs-134.
I'd thought that given its 2 year half life, it would be a fraction of 1% of the Cs-137 concentration. So what is generating the extra Cs-134?

 

[Rive]

The report linked above states, there was ~ equal amount of Cs-134 and 137 around in the water at the time of the accident.
68 months: 2.8 times the half time of CS-134 passed, so ~ 1/7 part of the original amount remained.
That ~ fits with the numbers I think.
I hope I did the math correctly...

 

[etudiant]

The report linked above states, there was ~ equal amount of Cs-134 and 137 around in the water at the time of the accident.
68 months: 2.8 times the half time of CS-134 passed, so ~ 1/7 part of the original amount remained.
That ~ fits with the numbers I think.
I hope I did the math correctly...

Thank you, that explains things nicely.

 

[Sotan]

The most recent "monthly progress report" translated in English has been posted on the METI site on October 27.
I am a little late but maybe some of you haven't seen it.
http://www.meti.go.jp/english/earthquake/nuclear/decommissioning/pdf/20161027_e.pdf

Also, the IRID site has several relatively new reports on topics such as "laser gouging technology for fuel debris", "concrete injection technology for repairing water leaks in the PCV", "full scale mock-up facility to simulate the lower part of the PCV".
http://irid.or.jp/en/topics/

 

[LabratSR]

Earthquake reported off Fukushima

http://geofon.gfz-potsdam.de/eqinfo/event.php?from=rss&id=gfz2016wxkb

Tsunami Warning Issued

http://www.jma.go.jp/en/tsunami/

 

[elektrownik]

Cooling stopped in Fukushima Daini reactor 3 after earthquake.

 

[LabratSR]

Tsunami warning issued after quake off Fukushima in Japan

http://hosted.ap.org/dynamic/stories/J/JAPAN_EARTHQUAKE?SITE=AP&SECTION=HOME&TEMPLATE=DEFAULT&CTIME=2016-11-21-16-20-38

 

[elektrownik]

SFP in unit 3 also lost cooling but they say that there is enough water to cool SFP.
(This is from NHK English Live stream)

 

[etudiant]

Surely the fuel in SFP 3 is decayed enough by now that the water does not need much cooling.
If power is restored before year end, I'd think there would not be a serious problem, but it would be better to hear from someone expert.

 

[Sotan]

From what I just heard from FNN (Japanese TV station), the cooling stopped at Unit 3 SFP in Fukushima Daini plant (not in the damaged Daiichi).
A safety device signalled that the water level is too low, therefore the cooling system stopped. But they presume it was only the sloshing caused by the earthquake. The temperature of the water rose by about 0.8 degrees Celsius (to a maximum of almost 30 degrees) before cooling resumed, so no danger there.

A tsunami of about 1m height has reached the shores in the area of the Fukushima nuclear plants, but didn't cause any additional issues.

Edit: more detailed report here:
http://kyodonews.net/news/2016/11/22/89417

 

[etudiant]

From what I just heard from FNN (Japanese TV station), the cooling stopped at Unit 3 SFP in Fukushima Daini plant (not in the damaged Daiichi).
A safety device signalled that the water level is too low, therefore the cooling system stopped. But they presume it was only the sloshing caused by the earthquake. The temperature of the water rose by about 0.8 degrees Celsius (to a maximum of almost 30 degrees) before cooling resumed, so no danger there.

A tsunami of about 1m height has reached the shores in the area of the Fukushima nuclear plants, but didn't cause any additional issues.

Edit: more detailed report here:
http://kyodonews.net/news/2016/11/22/89417

Thank you, Sotan, for an informative update. No US news service realized that it was Daini that was involved, so this is material new information for us here.

 

[Hiddencamper]

If it is the spent fuel pool cooling system, there is a low NPSH (Net positive suction head) trip on the spent fuel pool surge tanks. This also doubles as a low water level trip.

Due to vortexing and flow effects, any kind of sloshing can cause this instrument to have a false trip. Several US BWRs have removed this trip or added a time delay relay because it is problematic during refuelling activities.

 

[Sotan]

http://www.tepco.co.jp/nu/fukushima-np/handouts/2016/images2/handouts_161121_04-j.pdf
(in Japanese)

Short report on how Tepco actually dug a 1.2m deep hole in the ground, in one point right next to the "ice wall", to confirm that the soil is frozen.
There's a photo of a thermometer measuring minus 13 degrees Celsius.

As usually, larger images and a video (which I haven't been able to see) are available at:
http://photo.tepco.co.jp/date/2016/201611-j/161121-01j.html

 

[Sotan]

https://www.nsr.go.jp/data/000170874.pdf
Document (in Japanese) submitted by Tepco to NRA, dated November 18, containing a variety of topics.
What I found most interesting is the section from page 37 to 45 (page numbers as shown by Adobe Reader), with results of the video investigation of the Operating Floor of Unit 1.
Page 38: Until now they were able to confirm the position of several of the concrete blocks that make up the upper 2 layers of the 3-layered well plug. Namely, they can see a little of the upper layer's North and South block (center block not visible), as well as Center and West blocks of the middle layer. The other blocks can't be seen yet, they hope to see more as they remove debris from the surroundings. Middle layer's Center block has moved up and in the process has raised up the west tips of the upper layer's North and South blocks.
Page 39 gives a description of the well plug. A total of 9 huge concrete blocks, placed in 3 layers.
Page 40 shows the general location on the operating floor of the DSP slot plug, well plug and ceiling crane, as well as photos of the debris that allow a peek at the well plug blocks as described above. (I confess I don't understand very well what these photos are saying.)
Reactor well plug concrete blocks appear displaced in various directions... DSP slot plug (?), same... Unfortunately I lack the knowledge to interpret those photos.
Page 42 gives radiation measurements on the Operating Floor. Values over 50 mSv/h in the area of the reactor well plug. (Together with the well plug info, they seem very suggestive. Did the massive well plug concrete get jolted up during the accident? The hydrogen explosion wasn't likely to cause the movement of the concrete plug, I think...?)
Page 43-45 - damage sustained by the ceiling crane's wheel... runway girder... trolley.

 

[Rive]

...

Thanks for this document.

I might be wrong but for me the relative low radiation around the plug (and ~ the same levels for the whole floor) suggests that for this unit the containment cap did not failed.
I don't know how much void space might be under those concrete blocks, but the displacement might be due vacuum after the explosion and not due steam release.