Fukushima Japan Earthquake: nuclear plants Fukushima part 2

AI Thread Summary
A magnitude-5.3 earthquake struck Fukushima, Japan, prompting concerns due to its proximity to the damaged nuclear power plant from the 2011 disaster. The U.S. Geological Survey reported the quake occurred at a depth of about 13 miles, but no tsunami warning was issued. Discussions in the forum highlighted ongoing issues with tank leaks at the plant, with TEPCO discovering loosened bolts and corrosion, complicating monitoring efforts. There are plans for fuel removal from Unit 4, but similar structures will be needed for Units 1 and 3 to ensure safe decontamination. The forum also addressed the need for improved groundwater management and the establishment of a specialist team to tackle contamination risks.
  • #251
zapperzero said:
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.
 
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  • #252
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.
 
  • #253
etudiant said:
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.
 
  • #254
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.
 
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  • #255
Hiddencamper said:
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:

NRC said:
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
 
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  • #256
etudiant said:
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
 
  • #258
LabratSR said:
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.
 
  • #259
LabratSR said:
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.
 
  • #260
LabratSR said:
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...
 
  • #261
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
 
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  • #262
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
 
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  • #263
Gary7 said:
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.)
 
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  • #264
mscharisma said:
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).
 
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  • #266
I think the additional details and the confirmation of the Managing Director made this a page-one story yesterday.
 
  • #267
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!
 
  • #268
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.
 
  • #269
mscharisma said:
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.
 
  • #270
Hiddencamper said:
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.
 
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  • #271
Hiddencamper said:
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?
 
  • #272
zapperzero said:
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.
 
  • #273
Hiddencamper said:
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..
 
  • #274
gmax137 said:
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.
 
  • #275
Gary7 said:
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
 
  • #276
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.
 
  • #277
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!
 
  • #278
SteveElbows said:
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.
 
  • #279
Hiddencamper said:
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. :smile:
 
  • #280
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.
 
  • #281
Awesome Sotan! Thank you very much!
 
  • #282
RFI for Innovative Approach for Fuel Debris Retrieval"As fuel removal from spent fuel pool (SFP) at Unit 4 has started on November 18, 2013, “Mid-to-Long Term Roadmap (RM) on decommissioning of Fukushima Daiichi Nuclear Power Station” has shifted into the Phase-2.

The goal of the Phase 2 is to start fuel debris retrieval from reactor core expected in 2020. Necessary onsite works and associated R&D programs should be accelerated. From TMI-2 experience, retrieval of fuel debris is envisaged to be conducted by full submersion approach, in order to minimize workers radiation dose."

EDIT: Adding where this stuff is from http://irid.or.jp/fd/

Overview
http://irid.or.jp/debris/RFI_AFDR.pdf (I had to "Save Target As" to get this one to work)Technical Aspect of RFI 1
http://irid.or.jp/debris/TA_RFI1.pdf

Technical Aspect of RFI 2
http://irid.or.jp/debris/TA_RFI2.pdf
 
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  • #284
Sotan said:
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.

Alright, this is where I come in.

ADS (Automatic Depressurization System) is a part of each plant's ECCS (Emergency Core Cooling System) package. ADS does not directly cool the fuel in itself, but it is designed to automatically reduce reactor pressure to a level low enough that a low pressure ECCS system could inject water to the reactor. ADS works by sensing certain plant parameters, and, if those parameters are met, automatically lifting a select number of SRV (Safety Relief Valves) in their power operated relief mode to depressurize the reactor to the suppression pool.

The logic for an ADS initiation is roughly as follows:

Reactor water level < Level 1 (about 20 feet below normal, and a few feet above the fuel, this also starts the low pressure ECCS pumps)
AND
reactor water level < Level 3 (This is the low water level scram signal, about 2-3 feet below normal)
AND
any low pressure ECCS discharge piping has sufficient pressure for injection (signifies a low pressure ECCS pump is running)
AND
high drywell pressure (> 1.68 PSIG, this also auto starts all ECCS systems and puts the plant into LOCA mode)

Once all of the above are met, an alarm goes off in the control room for 105 seconds. If the operator does not manually inhibit the ADS system before the 105 seconds is up, ADS activates. ADS will blow down the reactor until the conditions are clear and the operator manually resets ADS.

If high drywell pressure is not present, but all the other conditions are present, the system will automatically activate after several (~6) minutes. High drywell pressure means a LOCA is in progress, so the ADS blowdown needs to happen as soon as possible. Without high drywell pressure, the 6 minute timer is long enough to give the operators time to restore cooling systems, and short enough that the fuel remains safe.

It looks like the torus/wetwell pressure was high enough to make the system think the low pressure ECCS pumps were running. The low pressure ECCS pumps take pressure from the suppression pool, so it makes sense that the logic was made up.
 
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  • #285
So, Hiddencamper, what is the effect on reactor water level from this blowdown if no new water is injected?
 
  • #286
Most Curious said:
So, Hiddencamper, what is the effect on reactor water level from this blowdown if no new water is injected?

This is almost a trick question lol.

In terms of inventory, you end up with a very large reduction in inventory. I'll do a simulator run tomorrow and see about how much my plant loses, but I think its something on the order of 20000 gallons from a hot scram. This is why one of the permissive signals for automatic blowdown is that you have a low pressure ECCS lined up to inject water. However, in terms of core cooling, if you manually perform the blowdown at the right time, you can buy yourself an extra 20 minutes until core damage.

Right as your first few feet of fuel starts to get uncovered, you reach a point where the top of the fuel is going to overheat. If you initiate ADS manually at this point, the rapid flow of steam across the fuel is capable of providing adequate core cooling for an extra 20 minutes, so even though you uncovered fuel and are losing water inventory more rapidly you can delay the onset of core damage. This is useful if you know you can get a fire pump lined up or if you just need a little more time to get RCIC/HPCI reset or something. It also removes energy from the vessel and transfers it to the suppression pool, which can reduce the impacts to a core breach later on.

Another thing to remember, and the reason I say this is a trick question, is whenever you are blowing steam down you are going to have an increase in INDICATED water level, due to how the water level instruments work. This has tricked operators into thinking they have an overfill condition (this is what happened at Three Mile Island), and can also cause your high water level logic to trip your feed pumps and turbines, which can be very bad if you are in an failure to scram scenario and counting on feed pumps and turbines for injection and decay heat removal.
 
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  • #287
Ah, the detail I wanted.

I assumed substantial water would be lost but did not fully understand the rest of it. At what pressure is below that required to drive turbine for HPCI?

How many gallons of water does 1 foot of level represent in the core then above active fuel where I assume the "fill" of hardware is less?
 
  • #288
Most Curious said:
Ah, the detail I wanted.

I assumed substantial water would be lost but did not fully understand the rest of it. At what pressure is below that required to drive turbine for HPCI?

How many gallons of water does 1 foot of level represent in the core then above active fuel where I assume the "fill" of hardware is less?

HPCI is designed to run down around the 150 PSIG range, but I'm pretty sure you can go a bit lower if you bypassed the interlocks. After a loss of condenser, HPCI equipped plants will run HPCI to drop pressure down low enough for decay heat removal pumps to start, instead of lifting SRVs, and the decay heat removal interlocks are in the 100-150 PSI range (depending on plant). RCIC is capable of running down to like 50 PSIG. Remember that this is pressure across the turbine, as wetwell pressure goes up (the steam exhaust point), your inlet steam pressure requirements will also increase.

As for water level...it kind of depends. I've been taught for my BWR that the 'rule of thumb' is 200 gallons per inch, but this could be off quite a bit depending on where you are. I'll say this much, when water is dropping, it always feels like its less than 200 gal/inch, but when you're trying to fill with non-ECCS systems, it feels like much more than 200 gal/inch lol. A typical BWR will have roughly 20 feet between normal water level and 2/3rds core height (minimum required for adequate cooling post LOCA)
 
  • #289
I thought it might interest others so here’s my translation of the last page of the TEPCO report located at:
http://www.tepco.co.jp/cc/press/betu13_j/images/131213j0101.pdf

(This report presents progresses regarding some of the 52 “unconfirmed/unclarified aspects” which have been identified in the past in relation to the nuclear accident. Page 6 of the report shows that TEPCO has solved about 10 of these aspects – and they are the ones given in this report; the solving of these 10 items has, though, led to 2 additional items to be investigated in the future, which brings the total to 54. Also, 10 aspects have been flagged as priority items for future investigations; and there are 34 remaining ones which have not been assigned a priority level yet.)

The last page of the report lists the 10 unconfirmed/unclarified aspects which are to be given priority:

- Investigation of the operation of SRV valves after the occurrence of the damage of reactor cores
- Circumstances of the release of radioactive material after March 20
- A more precise evaluation of the amount of water actually poured into the reactor(s) by use of fire trucks
- An evaluation of the effect of the HPCI system operation at reactor 3 on the course of the nuclear accident
- Behaviour of melt core falling towards the lower plenum
- Establishing the cause of high level radioactive contamination of the RCW piping at reactor 1
- Investigation of the rise in pressure that occurred in reactor 2, after the forced pressure reduction was carried out
- Establishing whether the rupture disk on reactor 2 operated or not
- Elucidation of the cause of RCIC system stopping at reactor 3
- Investigation of the temperature (related) stratification in the forced suppression pool at reactor 3.

(On page 6 it is stated that they will try to provide answers for this issues within 2 years.)
 
  • #290
I did a simulator run. From normal water level post scram an ADS blowdown left me with the top fuel just starting to get uncovered. When I did the same thing starting from top of active fuel, I ended with the core uncovered. I then started a 5800 gpm ECCS pump. In the fuel zone it was very low to flood up, possibly because I drained the downcomer to help lower level and it needed to be refilled. After that, came up about 30" a minute until we got near the dryer skirt (close to normal range), then it was like 50"/ minute. This is a 1050 MWe bwr with a rather small core for its size. I wouldn't take it as a direct comparison to a Fukushima type plant, but qualitatively it seems like if you use ads you will uncover the fuel.
 
  • #291
Perhaps I should mention that another, much longer, document came out on the same date, I assume exploring the same topics in more technical detail. But as my attempts to machine-translate it have given very poor results, I haven't tried to work out if there is anything of note in it.

http://www.tepco.co.jp/cc/press/betu13_j/images/131213j0102.pdf
 
  • #292
Hiddencamper, thank you for that effort. Helps me better understand the water level issues of an ADS event. No doubt the Japanese did not start with a normal level in the RPV initially so would have had even more uncovery of the fuel. With a lower flow rate to refill, (very likely) they were in deep doo doo no matter what they did!

I assume the steam would provide adequate cooling of the uncovered core but not for long? Seems "pulling the trigger" on ADS one needs to be confident of the ability to refill at high flow rate, which they did not have. I would think they may have had little choice once high pressure feed was no longer available but almost assured of fuel damage, at best, if they did blowdown to use low pressure feed. Heck of a choice to be faced with!

If radiation levels were already high when the HP injection pump failed, is it safe to assume fuel damage had ALREADY occurred? They were really up the creek with few, if any, options left it seems to me.

No doubt operational errors occurred, but even had everything been done exactly right, did they have ANY chance to prevent at least 2 if not 3 meltdowns with station blackout? TMI stopped just short of melt-out from the RPV - inside 4 hours - and they HAD power available!
 
  • #293
Most Curious said:
Hiddencamper, thank you for that effort. Helps me better understand the water level issues of an ADS event. No doubt the Japanese did not start with a normal level in the RPV initially so would have had even more uncovery of the fuel. With a lower flow rate to refill, (very likely) they were in deep doo doo no matter what they did!

I assume the steam would provide adequate cooling of the uncovered core but not for long? Seems "pulling the trigger" on ADS one needs to be confident of the ability to refill at high flow rate, which they did not have. I would think they may have had little choice once high pressure feed was no longer available but almost assured of fuel damage, at best, if they did blowdown to use low pressure feed. Heck of a choice to be faced with!

If radiation levels were already high when the HP injection pump failed, is it safe to assume fuel damage had ALREADY occurred? They were really up the creek with few, if any, options left it seems to me.

No doubt operational errors occurred, but even had everything been done exactly right, did they have ANY chance to prevent at least 2 if not 3 meltdowns with station blackout? TMI stopped just short of melt-out from the RPV - inside 4 hours - and they HAD power available!

With regards to ADS, in the BWR Owners Group EOPs (Emergency Operating Procedures), down the EOP-1 Inventory control leg, you are supposed to save blowdown until you reach top of active fuel, even if you have a low pressure source available. Once you hit TAF, you blowdown, then try injecting using all available low pressure sources. As I said in an earlier post, I think this is the first time ADS has ever actuated in a GE BWR outside of testing.

As for saving the other units, I have an opinion that if they were better prepared, if they had SAGs (severe accident guidelines), prestaged portable equipment, and severe accident strategies, they probably could have saved unit 2 or 3. Unit 2 had over 70 hours of RCIC, and unit 3 had like 36 total. Unit 3 in particular was coming down in pressure on HPCI and had a portable pump aligned to take over, but they secured HPCI prior to starting injection, didn't understand their SRVs were not functional at the time, lost pressure control, and exceeded the portable pump shutoff head. (The indicating lights for SRVs come from AC power, but the actual SRV solenoids use DC power, so you can get the light but not have the valve lift. This is why its important to verify proper RPV pressure response, SRV tailpipe temperature, and SRV acoustic detection, to ensure your SRV did in fact operate)
 
  • #294
SteveElbows said:
Perhaps I should mention that another, much longer, document came out on the same date, I assume exploring the same topics in more technical detail. But as my attempts to machine-translate it have given very poor results, I haven't tried to work out if there is anything of note in it.

http://www.tepco.co.jp/cc/press/betu13_j/images/131213j0102.pdf

Massive report, 327 pages... I am on the road now with limited time and internet access, but I intend to look through it eventually and will give at least a rough translation of contents. So that if anyone's interested in certain particular aspects I will try to find those chapters and respond.

In my translation the title sounds like this: "Estimation of the state of reactor 1 ~ 3 cores and containment vessels at Fukushima Daiichi Nuclear Plant and investigation of unsolved aspects. First progress report - 13 Dec 2013, Tepco".

It is a very detailed report which starts with the earthquake and the tsunami and their effects on the nuclear plant; then goes to analyze the accidents at reactors 1, 2 and 3, with detailed timelines of events and little conclusions chapters for each reactor; then there's an overall estimation of the state of each reactor core and containment vessel, from 1 to 3 (less than 1 page for each of them). Indeed, everything is discussed in more detail than in the other report published in the same day. And this brings us to page 54. The rest, up to page 327, are additional documents - something I haven't yet read at all, about the results of a MAAP (Modular Accident Analysis Program) analysis of the accident sequence; lots of graphs, each of them probably worth a separate discussion, and generally speaking lots of data which probably deserve more than what I am saying here.

It's an overwhelming document and I wouldn't know where to start (translating it all is close to impossible due to limited time). Perhaps there are some specific points you are interested in and would like to see what Tepco says about that; I could search for it and translate that portion. I'll be back home on Sunday. Even if there are no requests I will still post fragments that I find interesting (even though my poor knowledge of the field is not going to help much).
 
  • #295
If this document gets translated, if anyone sees a translation of it online, please post it. This contains a lot of critical detail towards understanding and analyzing the sequence of events. My Japanese is barely good enough to understand what graphs I'm looking at lol.
 
  • #296
Good news, I think:
http://www.tepco.co.jp/en/press/corp-com/release/2013/1233101_5130.html

It is stated that

"Attachment:
-‘Report on the survey and study results of unconfirmed and unexplained events of the Fukushima nuclear power plant accident - First Progress Report' (Summary) (PDF 2.15MB)PDF

-‘Report on the survey and study results of unconfirmed and unexplained events of the Fukushima nuclear power plant accident - First Progress Report' (Full Edition) (PDF 14.3MB)PDF

*English translations of the full edition report is now being developed and it takes a while to complete them.
We will post the translations one by one when it is prepared.
(The documents written in Japanese below will be replaced by English translations.)
We apologize for this inconvenience caused."
 
  • #298
Sotan said:
Looks like the shorter (summary) report has been translated in English:
http://www.tepco.co.jp/en/press/corp-com/release/betu13_e/images/131213e0101.pdf

Read it. Still would be nice to know why the HPCI stopped functioning. It's possible they just didnt have enough decay heat to drive it at the time.

Another piece they talk about is the worker observations of water spraying. They claim it was from the spent fuel pool. GE's SFP (spent fuel pool) design has air vents directly above the pool. These vents provide suction to ensure any radionucleides that offgas from the pool get captured by the plant's HVAC/filtering system. It appears the earthquake caused sloshing in the pool, which allowed the water to enter these air ducts, and that the design of the system is to drain that water out to prevent duct damage.

My plant has overfilled these ducts before (due to poor operation of the system). And during the 2008 Earthquake in Japan, several BWRs had water slosh out of their spent fuel pools. So this seems plausible in my opinion.
 
  • #300
Hiddencamper said:
Read it. Still would be nice to know why the HPCI stopped functioning. It's possible they just didnt have enough decay heat to drive it at the time.

The main body of the full report has now been translated. Regarding the RCIC stop, I found these [STRIKE]two[/STRIKE] three paragraphs which indicate that they still don't have an answer to that:

Page 2: "On other hand, there are still unclear issues, e.g., the reason why the reactor core isolation cooling (RCIC) system of Unit-2 lost its functions still remains unknown, and some observed phenomena cannot be interpreted yet."

Page 25: "The assumptions made in the analysis could reproduce quite well the reactor pressure changes, but why the RCIC stopped is unknown. It is necessary, therefore, that the RCIC shutdown mechanism consistent with those assumptions in the analysis be investigated (Unit-2/Issue-2). "

Page 36: "The RCIC stopped automatically at 11:36 on March 12th and thereafter its status of shutdown was confirmed on-site but its rest art-up failed after all. It was found upon an on-site check that the latch for the trip mechanism of the RCIC turbine trip throttle valve had been detached and the valve had been closed, but the background to this and reasons remain unknown and are subjects for continued examination (Unit-3/Issue-1)."
 

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