Japan Earthquake: nuclear plants Fukushima part 2

[a.ua.]

http://iss-atom.ru/book-4/glav-1-8.htm
FEATURES decontamination CHERNOBYL
Anyone is interested, can be read using Google translator

 

[nikkkom]

The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990

http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=NP-6931

 

[nikkkom]

For those who claim that only relatively few people died in nuclear accidents.

The example of how Soviets did not account for exposure of army conscripts in Chernobyl.

http://iss-atom.ru/book-4/glav-1-4.htm

Eyewitness acoount of a worker who was involved in drilling boreholes for nitrogen injection. It happened around 15 May 1986.

"In a pit 100 meters away from Unit 4 we were drilling horizontal boreholes up to 140 meter long for liquid nitrogen cooling of the burning reactor. The pit was 4 meter deep, with Japanese drill ТОР-LS made by "Tone Boring"...

Inside the pit radiation was 1.5 - 2.5 roenthgen/hour. But chunks of reactor graphite blocks were lying around the pit and radiation levels were from 40 to 400, in one spot even 800 R/h. Since our workers had to go up there too for some drilling instruments, it was raising radiation exposure. Maximum allowed dose was 25 R, anyone who took more was evacuated. We asked commander of chemical defence army unit to help with it, to reduce the number of people who had to leave because of taking the limit. He fulfilled our request very simply: soldiers came on a truck, loaded graphite chunks with bare hands, and drove away. You can imagine what dose they got doing that..."Chernobyl reactor's graphite was giving about 2000 R/h on contact. That is, 0.5 R/second.

 

[HowlerMonkey]

I've noticed that among the people who died very quickly that many died from a "piece of fuel lodged on a transformer" which was near a turbogenerator 7.

I only see reference to something like "10,000" or more as a reading but nothing on how this source was recovered.

 

[zapperzero]

I believe there is no engineering reason to shut the plant then. There had been concern about possible cumulative damage to the reactor pressure vessel, but these have been alleviated.

The uncertainties associated with annealing aside, many things which are very improbable on a 40-year timescale and were left out of the design basis, are by definition certain to happen at least once, on an infinite timescale.

 

[Hiddencamper]

The uncertainties associated with annealing aside, many things which are very improbable on a 40-year timescale and were left out of the design basis, are by definition certain to happen at least once, on an infinite timescale.

Annealling doesn't restore the full strength of the vessel. There's a good graph of it out there (that I can't find at this moment), that basically shows how you gain some life back, but its never going to be as good as the first dozen EFPY (Effective full power years) on the vessel. You can probably get most pressure vessels to 80 if you anneal, and you have been very kind on your vessel.

Assuming the plant was well designed and has no other major concerns like ASR (Seabrook station), and they haven't done an excessive amount of freeze seals on their ASME class 1 pipe, the most limiting component for life of the plant is the beltline of the reactor pressure vessel. This is the region around the active fuel. When a reactor vessel is built, it is rated with a certain number of EFPY (Effective full power years). For plants in the US, it was assumed that you would have 32 EFPY out of the vessel. This is the equivalent of an 80% capacity factor for the life of the plant, or running 32 years at 100% capacity factor. EFPY looks at the effects of radiation/neutron fluence to the vessel itself.

All vessel EFPY estimates include a certain number of normal, abnormal, emergency, and faulted cycles. Normal cycles are things like boltup, normal heatup, normal cooldown, hydrostatic pressure tests, daily power reduction, and a certain number of scram cycles. Abnormal looks at things like loss of some feedwater heating, turbine trips with bypass, rapid heatup and cooldown, cold water shocks. Emergency conditions look at things like turbine trip w/out bypass, total loss of feedwater + ECCS injection, MSIV (main steam valve) fast closure, maximum cold water injection shock. Faulted looks at DB LOCA and things with rapid depressurization, rapid pressurization, and things which could severely fault the vessel. There are engineers that keep track of every heatup/cooldown of the plant, and they have to tally these cycles against the number assumed in the original design of the reactor.

Each one of the events has a tabulated number of cycles. When reactor vessels are designed, they assume a poorly operating plant for 40 years, and that's how they come up with how many cycles they will assume. They then add these cycles to the total fatigue curves for the vessel end of life, (before they even happened). It's kind of like writing a check that you can cash later. As long as the number cycles I put my reactor through is less than what we assumed originally, the vessel is still good for use. If I go over the number of assumed cycles, I have to evaluate it. So for example, my plant assumes 10 total loss of feedwater heating events, where we go from full to no feedwater heating. As long as I don't use all 10 of those cycles, my vessel's EFPY curves are still conservative/bounding and my vessel is acceptable for use. If a plant uses up all of 1 type of cycle, let's say I used all my turbine trips or core spray injection cycles, I am allowed to do some limited substitution of a "Worse" event. So for example, high pressure core spray plants typically assume 10 HPCS starts over the life of the plant, and if I used most of those and I feel I need more margin, I can take some turbine trip without bypass cycles and convert them into HPCS injection cycles. The "exchange rate" (so to speak) is never favorable, as the goal is to ensure that your EFPY fatigue curves for your vessel are always bounding. I should make a note, that for the faulted conditions, like reactor emergency blowdown, only 1 cycle is assumed for the reactor, due to the extreme stresses it puts on the vessel. For people who work in GE plants, there's a set of prints that show these assumed cycles. For everyone else, if you look at a US plant's FSAR (Final Safety Analysis Report), generally somewhere in chapter 3. An example of this is in the following link from LaSalle station's FSAR (US BWR plant, BWR/5 Mark II)

http://pbadupws.nrc.gov/docs/ML0813/ML081330054.pdf

If you go to section 3.9, they talk about thermal-mechanical transients. If you go to table 3.9-24 they talk about each type of transient, the temperature changes it causes, how many cycles are assumed. You guys may like looking through all of chapter 3, as it discusses seismic criteria, wind/tornado flooding, etc.

As ASME code evolved, as we've removed those test capsules from the reactor to check for neutron embrittlement, as more refined computer analysis have been developed, we've learned a lot about radiation/neutron fatigue and its effects on the vessel. One thing we learned, is we overestimated neutron damage a significant amount. Another thing we learned, is changing core design allows us to reduce neutron leakage from the core, which further reduces neutron damage to the vessel beltline. This is the reason why many plants were able to perform EPU (extended power uprate), increase their power by 20%, and their original 32 EFPY fatigue curves were STILL bounding.

As plants age and have to go through license extension, these fatigue cycles, the EFPY fatigue curves, are all looked at. To get a license extension in the US, the licensee has to demonstrate that after another 20 years of life that the vessel will either A: still be bounded by existing analysis, B: that an updated analysis using new methods/codes and data shows that the plant is still bounding, or C: repair the vessel (annealing) such that it can withstand the extension period. A plant is allowed to not demonstrate this at the time of license extension as long as their is a commitment to do one of the previous options prior to reaching the original calculated end of life on the vessel. Palisades nuclear plant (CE PWR in the US) has chosen to go this route. There's also option D, shut down the plant, which is an economic decision if the plant does not want to anneal and option B isn't going to give them enough margin to extend the license.

Many plants are bound by A if they were "Good" to their vessels during the first 25-30 years of life. Some plants have to use B, like Palisades. I do not know of any US plant planning to anneal. Oyster creek is an example of a plant that had to use a new analysis, and now their 32 EFPY curves were extended to 38 EFPY. In order to extend these curves, they had to take penalties to the vessel minimum temperature for criticality, the number of allowable cycles they have, and they had to use new computer codes.

This brings up one more point. The whole basis behind the beltline fatigue, is the fact that the reactor needs to survive 1 emergency blowdown with a coldwater shock without breaking. The vessel is assumed to break if its Reference nil-ductility temperature after the accident is at or above 200 degrees F (this means the vessel will not be ductile while steaming is in progress). For those who are not mechanically based, Nil-Ductility Temperature is the temperature where something goes from brittle to ductile. When you are below NDT, an object will shatter, while an object above the NDT will bend and flex. In the US, there is a safety limit applied to the NDT (I think you take the vessel NDT and add 70 deg F to it for BWRs. I'm not sure how this works in PWRs). The vessel cannot go critical if it is below this safety limit NDT. As the vessel is fatigued, one option to extend the life of the vessel, is to raise the temperature allowed for criticality and pressurization. A typical BWR is required to be above around 120 deg F before going critical. A plant can opt to raise this temperature (no higher than 200 deg F) if necessary to maintain adequate safety margin to the NDT and extend the life of the reactor vessel. PWRs can do the same thing, raise the minimum temp for criticality, however I do not believe PWRs have the same flexibility with these limits as PWRs do. PWRs are likely to suffer much worse damage due to cold water shock events, and need to have larger safety limits.

This whole thing I'm talking about is why reactor vessels have a normal limit of heatup/cooldown of 100 deg F per hour, and why Fukushima unit 1 was cycling its Isolation Condenser on/off. Obviously if the operators knew they were about to lose the ability to cycle the IC, they would have left it on and chosen to violate their cooldown limit, rather than cycle it where they lost the IC.

I know this was a bit wordy, but with the number of comments on vessel life and fatigue, and previous discussions on the isolation condenser, I thought this might be a good thing to put up.

 

[etudiant]

Thank you, hiddencamper, for another very informative reply.
It does indicate that there is a solid engineering rationale for limiting the operating life of nuclear plants.
It would be very helpful if some of our international contributors could outline how these issues are handled in other jurisdictions, as all this is important additional information, which I've never seen previously.

Separately, conventional fuel boilers presumably have the same thermal burdens but enjoy essentially indefinite lives because the neutron damage is absent in their case.

 

[Hiddencamper]

Thank you, hiddencamper, for another very informative reply.
It does indicate that there is a solid engineering rationale for limiting the operating life of nuclear plants.
It would be very helpful if some of our international contributors could outline how these issues are handled in other jurisdictions, as all this is important additional information, which I've never seen previously.

Separately, conventional fuel boilers presumably have the same thermal burdens but enjoy essentially indefinite lives because the neutron damage is absent in their case.

Not having any neutron fluence is a part of it. Another big thing to remember is that fossil plants don't have to keep pumping feedwater after a scram. The feedwater heaters are powered by turbine drain steam, which means after a scram/turbine trip your feedwater temperature is going to have a substantial temperature drop. ECCS injections are even worse, as that water can be as low as 70deg F going into a 545 deg F vessel. Fossil plants get to ignore all that and can just let feedwater shut off so they can keep their boiler in hot standby ready to fire up again when they fix the problem.

 

[rmattila]

As regards the operating license of the reactor pressure vessel, I suppose the practice is about the same in all countries. Pressure vessels in VVER-440 PWR reactors are especially challenging with respect to neutron fluence due to very small diameter of the RPV and a welding seam in the core region.

As regards the extension of the plant operating licence, the Finnish practice takes advantage of the flexibility enabled by having a very limited diversity in the reactor fleet. The key idea is constant evolution of the design bases. Whenever new regulatory guides are published, they apply directly to new plants, but also the old plants must present plans on how to follow the design basis changes specified in the guides. In order to keep their operation licenses, the plants must submit a comprehensive periodic safety review every 10 years, including the list of exceptions where the plants don't fulfil the design criteria for new plants. Some things you can't do much about - such as resilience against a large airplane crash or the earthquake resistance of the buildings - but improvements are usually needed for the plants to keep or renew their operation licenses.

As an example, both BWR reactors are currently being backfitted with RCICS to fulfil the Post-Fukushima requirement of a complete loss of the plant's internal electricity network. The operation licenses are due in about 5 years from now.

 

[gmax137]

Annealling ...

Hiddencamper has done an admirable job explaining some aspects of reactor vessel aging, but I would add that to 40-year license term is not based on the vessel life -- it is actually the other way around. The 40 year license term is specified in the Atomic Energy Act, and was, as far as I can tell, based on similar license terms for large hydro electric dams and (possibly) radio station licenses. The NRC website hints at this:

The Atomic Energy Act and NRC regulations limit commercial power reactor licenses to an initial 40 years but also permit such licenses to be renewed. This original 40-year term for reactor licenses was based on economic and antitrust considerations -- not on limitations of nuclear technology. Due to this selected period, however, some structures and components may have been engineered on the basis of an expected 40-year service life.

from http://www.nrc.gov/reactors/operating/licensing/renewal/overview.html

 

[jmelson]

Thank you, hiddencamper, for another very informative reply.
It does indicate that there is a solid engineering rationale for limiting the operating life of nuclear plants.
It would be very helpful if some of our international contributors could outline how these issues are handled in other jurisdictions, as all this is important additional information, which I've never seen previously.

Separately, conventional fuel boilers presumably have the same thermal burdens but enjoy essentially indefinite lives because the neutron damage is absent in their case.

Coal, etc. plants don't have a HUGE pot of hot water, they have many small tubes exposed to the
hot flue gas. A defective tube can be plugged if it develops leaks, until the next major shutdown.
The tubes corrode away and have to be periodically replaced, but that is a tractable overhaul
job. A major failure of a boiler tube leads to an unplanned shutdown, but it doesn't destroy
the plant or cause a radioactive release. So, the danger related to these is on a different
scale. But, due to their smaller size, they can withstand much quicker thermal transients
without major failure.

The nuclear RPV can't be replaced without tearing the entire plant apart, so that just isn't
done. And, the consequences of a failure of the RPV or immediately adjacent parts
could be totally catastrophic.

Jon

 

[LabratSR]

From Ex-Skf

Fukushima I Nuke Plant: TEPCO Estimates 25 Sieverts/Hour Radiation at the Bottom of Exhaust Stack for Reactors 1 and 2

http://ex-skf.blogspot.com/2013/12/fukushima-i-nuke-plant-tepco-estimates.htmlWhat could cause readings this high in the vent pipe?

 

[etudiant]

From Ex-Skf

Fukushima I Nuke Plant: TEPCO Estimates 25 Sieverts/Hour Radiation at the Bottom of Exhaust Stack for Reactors 1 and 2

http://ex-skf.blogspot.com/2013/12/fukushima-i-nuke-plant-tepco-estimates.html


What could cause readings this high in the vent pipe?

Vaporized/pulverized fuel material?
In a core melt situation, especially with some water still below the core, it seems plausible that molten bits of core hitting the water might get entrained by steam.
I've not seen any reference to such in the various core melt scenarios, but assume it has been considered. Or maybe we will learn something more about reactor failure effects in the real world.

 

[a.ua.]

From Ex-Skf

Fukushima I Nuke Plant: TEPCO Estimates 25 Sieverts/Hour Radiation at the Bottom of Exhaust Stack for Reactors 1 and 2

http://ex-skf.blogspot.com/2013/12/fukushima-i-nuke-plant-tepco-estimates.html


What could cause readings this high in the vent pipe?

Maybe it's because the pair were removed from the 2 and 1 (in part) from the reactor containment, without the participation of the torus?.
Note that the geometry of the pipe.

* Maximum radiation in the near knee from the main stack.

 

[nikkkom]

From Ex-Skf

Fukushima I Nuke Plant: TEPCO Estimates 25 Sieverts/Hour Radiation at the Bottom of Exhaust Stack for Reactors 1 and 2

http://ex-skf.blogspot.com/2013/12/fukushima-i-nuke-plant-tepco-estimates.html


What could cause readings this high in the vent pipe?

The exhaust stack is a pretty large, and cold, metal pipe. When steam went through it, some of it condensed on the walls and subsequently drained down. With a lot of dissolved Cs, I guess...

 

[Gary7]

Story hitting the papers today is that according Tepco, much of the water injected via fire-engine pumps into units 1-3 in the early days of the accident never reached the cores. The piping leading to the cores from the external inlet splits off at several points, and Tepco is speculating that a lot of the fire-engine pumped water went into one of these diversions. In unit 1 there are ten different locations where the pipe branches off. In units 2 and 3 there are four different branchings. Tepco calculated that in order to avoid a meltdown they needed to pump over 10 tons of water every hour into the cores. They pumped in 75 tons. At the end of March 2011 they verified the presence of water in the unit 2 tank, where none was expected. (There is no clarification of what tank they are referring to).

http://www.tokyo-np.co.jp/article/national/news/CK2013121402000120.html?ref=rank

 

[Gary7]

Another news source has slightly more clarification. JPN 47 News is saying that there were vents in the pipes that ensure the water goes to where it is needed, but at the time of the accident the radioactivity was so high it became difficult to operate the vents, and that the water flowed into pipes where no one had anticipated it would flow.

http://www.47news.jp/CN/201312/CN2013121301002447.html

 

[mscharisma]

Story hitting the papers today is that according Tepco, much of the water injected via fire-engine pumps into units 1-3 in the early days of the accident never reached the cores. The piping leading to the cores from the external inlet splits off at several points, and Tepco is speculating that a lot of the fire-engine pumped water went into one of these diversions. In unit 1 there are ten different locations where the pipe branches off. In units 2 and 3 there are four different branchings. Tepco calculated that in order to avoid a meltdown they needed to pump over 10 tons of water every hour into the cores. They pumped in 75 tons. At the end of March 2011 they verified the presence of water in the unit 2 tank, where none was expected. (There is no clarification of what tank they are referring to).

http://www.tokyo-np.co.jp/article/national/news/CK2013121402000120.html?ref=rank

I am puzzled by this. I could have sworn this was discussed in this thread quite a while back, but I sure as heck can't find it now. Of course, I could have read it elsewhere but one way or another this was not news to me. Anyone else surprised by this being news now or am I just losing my mind? (The second part of this question is rhetorical, so please don't answer that.)

 

[Astronuc]

I am puzzled by this. I could have sworn this was discussed in this thread quite a while back, but I sure as heck can't find it now. Of course, I could have read it elsewhere but one way or another this was not news to me. Anyone else surprised by this being news now or am I just losing my mind? (The second part of this question is rhetorical, so please don't answer that.)

It is probably discussed early on in the original thread. It was not clear that the water level indicators were reading correctly, and after the explosions it was surmised that little, if any water, got to the cores, which would have been sitting in dry steam, or effectively in adiabatic conditions.

Rapid oxidation of the cladding (and production of hydrogen) implies high temperatures, and not necessarily melting temperatures, since chemical reactions begin at lower temperatures, e.g., eutectic temperatures. The dissolution of Fe, Cr and Ni (in steels and nickel alloys) in Zr starts around ~850°C, well below the melting temperature of Zr alloys. Rapid oxidation occurs as well.

Normally in a BWR, the cladding temperature is <300°C on the outer surface, and the coolant temperature is at saturated conditions ~285-288°C (depending on operating pressure).

 

[Rive]

As I recall, it was an external link, maybe from Ex-SKF.

Ps:after some googling:
http://ex-skf.blogspot.hu/2013/03/nhk-55-of-water-injected-into-reactor-3.html
http://ex-skf.blogspot.hu/2013/04/now-they-tell-us-only-10-of-water-was.html

As it seems, it's really an old story.

 

[Gary7]

I think the additional details and the confirmation of the Managing Director made this a page-one story yesterday.

 

[mscharisma]

Thank you all. You and your posts continue to be of great help to a lay(wo)man like me to get and hold on to a grasp of at least the basics of what's going on. Much appreciated!

 

[etudiant]

Seems that the cost of a nuclear exit is too steep for Japan.
http://www.asahi.com/english/articles/TKY201312140119.html

The recommendation to 'embrace nuclear power' may be well founded, but the marketing looks to be a challenge. Of course, the benefit may be that the Japanese government takes more direct responsibility for the industry, rather than having TEPCO serve as a spear catcher.

 

[Hiddencamper]

I am puzzled by this. I could have sworn this was discussed in this thread quite a while back, but I sure as heck can't find it now. Of course, I could have read it elsewhere but one way or another this was not news to me. Anyone else surprised by this being news now or am I just losing my mind? (The second part of this question is rhetorical, so please don't answer that.)

Typically fire suppression piping is non-seismic, and utilizes a "Ring Header", meaning that there is a main pipeline that feeds the entire fire system for the reactor building. If you had a leak or break in any point, you may not have gotten any water onto the reactor.

Really, if you were unable to confirm water was getting onto the core, you should have just abandoned trying to save the cores, and switched to flooding the heck out of containment to protect it from breaching when the hot debris ejection occurs. This is what the US EOPs (emergency operating procedures) have you do, if you cannot flood the core then you exit all EOPs and enter all SAMGs (Severe accident management guidelines) which direct you to flood containment.

 

[Rive]

Really, if you were unable to confirm water was getting onto the core, you should have just abandoned trying to save the cores, and switched to flooding the heck out of containment to protect it from breaching when the hot debris ejection occurs.

They did try it at the middle of the first week as I recall (or at least the news were filled with 'flooding the drywell'). I don't know at which point had they gave up.

Ps.: sorry, I missed. It was around the middle of April at least for U1.

 

[zapperzero]

exit all EOPs and enter all SAMGs (Severe accident management guidelines)

I seem to recall a discussion early on about how there was no SAMG document where/when needed? And the plant management basically played it by ear?

 

[Hiddencamper]

I seem to recall a discussion early on about how there was no SAMG document where/when needed? And the plant management basically played it by ear?

This is true.

 

[gmax137]

Typically fire suppression piping is non-seismic...

While that is true in US plants, I think the fire piping in Japan is seismic. Maybe someone closer can comment on that.

Not that it really changes your argument, but..

 

[Hiddencamper]

While that is true in US plants, I think the fire piping in Japan is seismic. Maybe someone closer can comment on that.

Not that it really changes your argument, but..

I think you're right. I think Japan's building codes would require seismically capable piping.

 

[SteveElbows]

I think the additional details and the confirmation of the Managing Director made this a page-one story yesterday.

The acknowledgment of these things also has implications for TEPCOs own analysis of things like core melt. It allows them to update such stuff to be somewhat more credible. I certainly spent a while complaining about their analysis in the past, and one of my complaints was the rather optimistic way they appear to have made calculations combining decay heat estimates with how long each reactor supposedly went without sufficient cooling.

I haven't seen an english version of the report they issued on the 13th, and I know they looked at other issues too, but I see for example on page 37 of the following document a diagram indicating the implications. i.e. we now see a large blob of melted core in the pedestal area rather than the very small blob with only partially melted fuel rods as seen in their 'optimistic' reports of the past. If I recall correctly they mostly applied the optimistic scenario to reactors 2 & 3 in the past. They couldn't manage such optimism with reactor 1 because even with the faulty assumption that pumped water all reached its target, there was still too much decay heat & too long a time elapsed to get 'PR happy' results out of the models for reactor 1.

http://www.tepco.co.jp/cc/press/betu13_j/images/131213j0101.pdf

 

[Most Curious]

What is the downside to flooding containment? Could it have been done simultaneous with trying to refill the RPV?

Would it have even been possible? Now water is pumped into the RPV (they think) leaks into the drywell and then into the basement. Clearly there are some major leaks, both out of the RPV and the dry / wet wells. Guess the $64,000 question is how much of the leakage is result of RPV penetration during meltdown and how much came from earthquake damage.

 

[etudiant]

To fully flood the facility might require 200,000 cubic meters of water.
The fire engines on the site had maybe 4000 liter/min pump capacity, so 50,000 fire truck minutes of pumping.
There is about 10,000 minutes/week, so assuming they had 5 fire trucks, they could have flooded the site in a week.
Of course, there was a lot of water in the plant to start with and maybe they had 10 fire trucks, but at best it would have taken several days from the time they started. Seems the missing SAMG was really missed!

 

[Hiddencamper]

The acknowledgment of these things also has implications for TEPCOs own analysis of things like core melt. It allows them to update such stuff to be somewhat more credible. I certainly spent a while complaining about their analysis in the past, and one of my complaints was the rather optimistic way they appear to have made calculations combining decay heat estimates with how long each reactor supposedly went without sufficient cooling.

I haven't seen an english version of the report they issued on the 13th, and I know they looked at other issues too, but I see for example on page 37 of the following document a diagram indicating the implications. i.e. we now see a large blob of melted core in the pedestal area rather than the very small blob with only partially melted fuel rods as seen in their 'optimistic' reports of the past. If I recall correctly they mostly applied the optimistic scenario to reactors 2 & 3 in the past. They couldn't manage such optimism with reactor 1 because even with the faulty assumption that pumped water all reached its target, there was still too much decay heat & too long a time elapsed to get 'PR happy' results out of the models for reactor 1.

http://www.tepco.co.jp/cc/press/betu13_j/images/131213j0101.pdf

On page 37 I think, it appears they initiated ADS. Someone needs to translate that. If they did activate ADS, it will be the first time the system has been activated in a functioning reactor.

 

[turi]

On page 37 I think, it appears they initiated ADS. Someone needs to translate that. If they did activate ADS, it will be the first time the system has been activated in a functioning reactor.

Ah, ADS as in Automatic Depressurization System and not as in Accelerator Driven System. That confused me a tad bit.

 

[Sotan]

Hello everybody.
This is my first post on these forums.
But I have been following this thread for a while as I am interested in the discussion of Fukushima plant developments.
I am in no way a specialist in this field or in any physics field therefore I doubt I could contribute to these discussions. But I do happen to know Japanese at a fair level, though, and I thought… perhaps I could help with the Japanese translations, sometimes. Unless there's somebody better at this too, that is.

This is a test posting; please tell me if I am useful or I am in the way; no offense will be taken if it's the latter.

------------------------
This is an attempt to translate & summarize some of the things related to ADS from the TEPCO report of 20131213, link given in the posts above.

- ADS is mentioned first on page 32, in a chapter dedicated to the issue "The cause of the sudden/fast decrease of pressure in reactor #3 (and the possibility of it being due to some hole that appeared in the main installations of the reactor)".

The situation or level of knowledge before this study:
It was believed that the sudden decrease in reactor 3 pressure that occurred on March 13 around 19:00 hours was the result of operator action - that is, the opening of SRV (Safety Reliev Valves?).

The results of this study:
It was established that the decrease of reactor pressure occurred while the operators were making preparations to manually decrease the pressure. There is the possibility that the pressure decreased as specific conditions were met for the activation of ADS.

(jumping to page 33 - graph of reactor pressure vs time; on the time axis time increases from right to left)

(moving to page 34)

Investigation of the conditions needed for the activation of ADS

The sudden, quick decrease in pressure could be explained by the activation of ADS, but we used to believe that the conditions for the activation of ADS had not been met on reactor 3.

*one of the conditions for the activation of ADS is making sure that the low pressure water system is ready for operation.

The diagram on page 34 indicates that while 3 factors needed for ADS activation were indeed cleared, the output of pumps for the system of removing residual heat and the system for spraying the inside of the reactor was insufficient (these pumps couldn't be operated due to loss of electric power).

The conclusion is that, logically speaking, the ADS system was not supposed to operate.

We investigated the possibility of ADS ending up operating, in spite of the fact that the logical procedure for its operation did not appear as having being achieved.

(moving on to page 35)

We thought, what if the conditions for the operation of ADS were in fact met. What about this possibility.

Due to the rise in pressure in the S/C (suppression chamber?), even though the pump(s) in the residual heat removal system were not functioning, the fact that a certain (significant) value of pressure on the output of this pump could be read might indicate that the conditions for ADS operation were in fact met.

(the diagram indicates that) S/C pressure reaches 0.455 MPa (abs) -> the pressure is transmitted -> pressure gauge measures a value that exceeds the 0.344 MPa needed for ADS activation

(moving on to page 36)

The actually measured data as well as analysis data were considered in relation with the decrease in reactor pressure.

The graph on this page shows that various actually measured parameters (the SRVs and the water level) are consistent with the hypothesis that the ADS had in fact been activated around 08:56 hours.

(moving on to page 37)

Considering the possibility that the cooling by water might have been insufficient, they are modifying the proposed graphic depiction of reactor 3 damage as shown in these drawings. Left is what they used to believe - right is what they think now.