Fukushima Japan Earthquake: nuclear plants Fukushima part 2

[SteveElbows]

I think it's simply from the contaminated water from the torus. That water is still from the first days of the accident, with Cs levels at the 10^6 range (or even higher).

To clarify this: I don't saying that there are *no* core debris in the torus. I'm just saying that the radiation levels are not sufficient to imply that there are.

I believe we need a lot more detailed mappings of radiation levels at a wide range of locations in the reactor 1 torus room. Especially since there is quite a large disparity between the first set of 'probe dangled on a wire' radiation readings at different heights within the reactor 1 torus room, and the second one which was used to make the graphic posted earlier. The large difference in radiation levels in that torus room compared to the others is of interest, but I agree that we should not jump to conclusions. The lack of reactor data during key stages of reactor 1 meltdown does not help. Nor does the failure to locate water leakage points at the other reactors.

Certainly before getting too carried away it is important to compare the several Sv/hr readings from the torus room with the multiple tens of Sv/hr we've seen from, for example, the last survey of the area approaching reactor 2 pedestal. Personally I lack the knowledge to appreciate the full potential of water shielding in the torus and the torus room, that may be an important factor when trying to reach any tentative conclusions.

I think that public awareness and discussion of this stuff has, like so many other aspects of the disaster, not been helped by the failure of various official narratives to really join dots, even tentatively, between possible events that happened and some of the specific data we get. For example the high radiation level at certain locations within the shared reactor 1/2 stack and associated pipework was not met, as far as I know, with a concise narrative about the various possible explanations for this. Throw in a potential lack of public awareness between corium/fuel and various other forms of radioactive elements that found their way into various parts of the reactors, and the crude state of narrative from certain anti-nuclear agenda driven sources (e.g. reactor 3 plutonium fuel fixation), and I am rather underwhelmed by the level of clarity offered to those looking for easily consumable explanations. We know its a long, slow journey to get enough solid data about all manner of things, but in the meantime far more could have been done to understand what the various realistic possibilities are, and to point out when something is discovered that tends to rule stuff in or, more often so far it seems, out. There have been all manner of occasions where accumulated knowledge shared on this forum has had the potential to offer narratives and tentative conclusions that far exceed that offered pretty much anywhere else in public. A summary of where we are at so far in relation to many things could be constructed from it and may be useful, but the level of collaboration required may be tricky, or considered too tedious given that so many question marks remain and that a prize of stumbling on some important revelation does not seem to be on offer at this stage anymore than it was during early photo-gazing.

Neither of those goals are really difficult. They will get some experience soon with freezing, as they trying to seal the trenches, and actually they are working on decontaminating the top of U3. They should be able to 'cut down' the turbine buildings and set up an acceptable working conditions on top of U3 on planned order.

I am far less optimistic about that, it is far from trivial to get the radiation levels down to acceptable levels anywhere near reactor 3 building. I think there is plenty still to be revealed about specific sources of radiation in and around that building. Decontaminating the upper levels is clearly important, but the wider area seems to still have some notable sources of radiation that make dose rates for workers in the entire region of the reactor 3 building rather impractical. How much they can do via remote control is likely to remain important - So far they've done quite an impressive job of removing debris from the upper floors of reactor 3, but the full challenges of dealing with the reactor 3 pool are yet to receive enough detailed public discussion. Likewise the survival of reactor 2 building presents some challenges with gaining access to that pool, and reactor 1 schedule has been lengthened by the need to undo the initial work they did constructing an outer shell for that building. If there were not so much fuel in reactor 4 building, and there had not been such incense concerns about that fuel pool at the height of the disaster, I suspect there would have been more public focus and concern on dealing with these other pools.

 

[zapperzero]

You can all pontificate all you want about how you think it should work, but you need to understand the design of these plants, along with the regulatory design requirements, to understand where the challenges are in just saying that some passive thing can be installed that will magically solve all your problems post accident.

Holy run-on sentences, Batman! Of course you can't pre-solve every possible issue. This doesn't mean passive safety features aren't better than active ones.

No matter what, it will take significant efforts by those at the plant to cope with a beyond design basis accident, with or without a filter.

"No matter what"? That's a sweeping generalization if I ever saw one. With a passive, filtered vent, it will be easier, because it removes the dilemma in which the operators of Fukushima 1 found themselves on the fateful night... to vent and blanket the environs with radio-iodine, or to NOT vent and risk even worse contamination? In the event, the choice was was made for them...

 

[a.ua.]

a radiation unit 3
After removal of the crane girders
http://ru.fotoalbum.eu/images1/200905/95064/273731/00000055.JPG

http://www.meti.go.jp/earthquake/nuclear/pdf/131031/131031_01g.pdf

For[/PLAIN] the "fuel debris retrieval
Fuel debris situation analysis "
j

 

[Hiddencamper]

Holy run-on sentences, Batman! Of course you can't pre-solve every possible issue. This doesn't mean passive safety features aren't better than active ones.



"No matter what"? That's a sweeping generalization if I ever saw one. With a passive, filtered vent, it will be easier, because it removes the dilemma in which the operators of Fukushima 1 found themselves on the fateful night... to vent and blanket the environs with radio-iodine, or to NOT vent and risk even worse contamination? In the event, the choice was was made for them...

The "no matter what", is because, by DEFINITION, a beyond design basis accident is one where all permanently installed onsite plant equipment fails to perform safety functions to prevent a core damaging event.

Under that definition, of a beyond design basis accident, that means the ONLY actions that will be effective are manual actions.

Even with a passive filtered vent, Fukushima operators found themselves unable to open their vents due to rupture disks that failed. They also didnt have portable equipment, plans, procedures, or leadership to ensure scrubbing was performed. Even if they had a passive filter, the fact that the rupture disks at Fukushima failed at multiple units means that passive filters wouldn't have helped at all if they were there.

 

[etudiant]

The "no matter what", is because, by DEFINITION, a beyond design basis accident is one where all permanently installed onsite plant equipment fails to perform safety functions to prevent a core damaging event.

Under that definition, of a beyond design basis accident, that means the ONLY actions that will be effective are manual actions.

Even with a passive filtered vent, Fukushima operators found themselves unable to open their vents due to rupture disks that failed. They also didnt have portable equipment, plans, procedures, or leadership to ensure scrubbing was performed. Even if they had a passive filter, the fact that the rupture disks at Fukushima failed at multiple units means that passive filters wouldn't have helped at all if they were there.

Is this correct?
I was not aware that the Fukushima site had filtered vents, or am I misreading the post?
I had thought that they delayed venting because they were concerned about unfiltered emissions from the failing reactors and that by the time they wanted to vent, they no longer could because there was no power.
The failure of the rupture discs is unsurprising to anyone who has worked in the electronics industry, it is always the high priced chip that gets fried, not the sacrificial diode or such that was supposed to protect it. The rationales posted for why these discs failed in this instance seem a little tortured, but I've not seen the official explanation or analysis, if it has been released. One would think that the Nordic system operators would be quite concerned about this aspect.

Your central point that managing a 'beyond design basis' accident really requires a trained operator staff who have a framework and appropriate tools to keep the beast in check is critical.
Fukushima shows what happens when these are not adequately provided.

 

[Hiddencamper]

Is this correct?
I was not aware that the Fukushima site had filtered vents, or am I misreading the post?
I had thought that they delayed venting because they were concerned about unfiltered emissions from the failing reactors and that by the time they wanted to vent, they no longer could because there was no power.
The failure of the rupture discs is unsurprising to anyone who has worked in the electronics industry, it is always the high priced chip that gets fried, not the sacrificial diode or such that was supposed to protect it. The rationales posted for why these discs failed in this instance seem a little tortured, but I've not seen the official explanation or analysis, if it has been released. One would think that the Nordic system operators would be quite concerned about this aspect.

Your central point that managing a 'beyond design basis' accident really requires a trained operator staff who have a framework and appropriate tools to keep the beast in check is critical.
Fukushima shows what happens when these are not adequately provided.

I'm really referring to the standby gas treatment system, which is a combination of HEPA filters and charcoal beds. SBGT is supposed to maintain a vacuum in the secondary containment to filter any leaks through the primary, and also has a backup function to vent the primary containment. As SBGT is a charcoal based filtration system, its effectiveness relies upon the ability to remove moisture from effluents (as well as active power to open the valves, dampers, run blowers, and run heaters/dehumidification modules). Like I said, it is not a passive filter, and is not very effective compared to wet scrubbing or a large dry filter like those in Europe.

The safety logic in BWRs, when it sees an increase in effluents from the exhaust stack (depending on plant design, typically >10 mRem/hr from the secondary containment or >100 mRem/hr from primary containment exhaust), will automatically shut down all normal exhausting/ventillation systems and activate SBGT. All effluents are then routed through SBGT prior to exhaust to remove radioactive material. This is a standard feature in BWRs. This transfer to SBGT also automatically occurs if a LOCA signal is detected ( below the level 1 low water alarm setpoint or high drywell pressure). Again, all active logic, AC/DC power is required.

 

[Rive]

I am far less optimistic about that, it is far from trivial to get the radiation levels down to acceptable levels anywhere near reactor 3 building.

Definitely not trivial, but even if it'll take some years, it's not 'difficult'. They have everything they need for this task, so they *should* be able to start the manned work there within a year or two.

Should, would, will - that's a different matter, of course. We'll see.

IMHO the difficulties will start with removing the (fuel and other) debris from U3 pool. That'll be something different. The possibility that some fuel debris stuck to the FHM or construction material debris and gets to the surface with them...
Maybe they will cut it all to pieces underwater and cask it all.

If there were not so much fuel in reactor 4 building, and there had not been such incense concerns about that fuel pool at the height of the disaster, I suspect there would have been more public focus and concern on dealing with these other pools.

I'm actually using this as a kind of reliability check. If a source is more concerned about U4 pool/building than U3 pool/building, then it's most likely missed some important points in this story.

All the stuff you wrote about the public awareness and discussion is correct. Well said.

Ps.: some sources are really trying to keep the story going the same speed as in the start, whatever it costs. But it's different now, with much less drama, so these kind of efforts requires much and much 'inventions'.

 

[SteveElbows]

All this talk of rupture disc problems runs the risk of downplaying the other issues that delayed venting, and also runs the risk of making it sound like venting failed at all the reactors, as opposed to the apparent reality that it was mostly reactor 2 where the failure to vent story was allowed to play out in full. Thats certainly the only reactor where they seem rather unsure as to whether the rupture disc ever ruptured, and where no torus-scrubbed venting is thought to have taken place.

Having said that, the delays to venting at reactors 1 and 3 obviously caused additional delays in pumping water into those reactors.

Certainly to my mind the problem with the rupture disc-based system seems to be due to the fact that the initial design considerations for this system were heavily focussed on the idea of containment failures due to over pressurisation. Only in later theoretical loss of coolant accident analysis papers did other modes of containment failure, such as temperature-related failure of seals and penetrations, get due attention. Really the entire system seems inappropriate not just for venting under situations where other containment failures kept containment pressure below the level necessary to rupture the discs, but also for situations where there is a desperate need to reduce containment pressure much earlier on in order to ensure that the RPV itself can be depressurised via SRV's in order to allow pumping in of water.

 

[rmattila]

In the Fukushima design, the rupture disk is in series with the two closed valves that must first be opened in order for the pressure to work on the disk. If the valves are not opened in time, the containment will develop a leak and the pressure might never reach the disk burst pressure again.

This is not the way to design it. The valves in the rupture disk lines should be kept open, and there should be a manual by-pass to the rupture disks.

Back in the 80's, it took about 3 years after the Chernobyl accident to have the filtered vents designed and installed in European BWRs. Of course, a prerequisite for that was that everybody agreed on their necessity and no time was wasted on arguing whether or not they should be built.

 

[nikkkom]

It all depends on the time and distances,
at the moment: 2.6 years decay,
2-3 kg of nuclear fuel will give 2 Sv, at the distance of 1 meter, without shielding metal or concrete.

I have doubts about your numbers.
Almost all radiation from spent fuel comes from fission products and minor actinides. Let's check how much those emit.
IIRC French reprocess the fuel after about 5 years of cooldown.
This document:

http://www.wmsym.org/archives/2003/pdfs/194.pdf

says that at French reprocessing plant (best in the world) after reprocessing, vitrification of the fission products and minor actinides, and pouring of the resulting glass into 0.5 cm thick walled stainless canister, dose rate on contact with canister surface is 14000 Gy/h.

Granted, it is 500 kg of material, not 3. OTOH, with canister diameter of 43 cm there is substantial self-shielding, and canister's wall shields all betas and most of low-energy secondary gammas, while in your situation ("without shielding metal or concrete") there is no such effect.

And it is a contact reading, not 1 meter reading.

Still, 14000 Gy/h is vastly higher than measly 2 Sv/h (~=2 Gy/h) you provided. I think you are wrong by at least an order of magnitude.

 

[zapperzero]

Having said that, the delays to venting at reactors 1 and 3 obviously caused additional delays in pumping water into those reactors.

Iirc reactor 3 depressurized all by itself? The plots are somewhere in the mega-thread.

Certainly to my mind the problem with the rupture disc-based system seems to be due to the fact that the initial design considerations for this system were heavily focussed on the idea of containment failures due to over pressurisation. Only in later theoretical loss of coolant accident analysis papers did other modes of containment failure, such as temperature-related failure of seals and penetrations, get due attention. Really the entire system seems inappropriate not just for venting under situations where other containment failures kept containment pressure below the level necessary to rupture the discs, but also for situations where there is a desperate need to reduce containment pressure much earlier on in order to ensure that the RPV itself can be depressurised via SRV's in order to allow pumping in of water.

Obviously such situations can be handled by other means. For example you could have a vent path that is under operator control and feeds into the same filter. I don't see how you can argue against the necessity of a passive venting system by pointing out that there are failure modes it doesn't address. Shall we not install curtain (lateral) airbags in cars, because there are also lots of head-on collisions?

 

[a.ua.]

nikkkom

I have doubts about your numbers.

Should be considered not in the mass, but the number of terabecquerels.
For exact calculation is also required to know the power of fuel burn per day, amount of fuel enrichment.
Just keep in mind that the radiation power as a function of the distance does not vary linearly.
is the square of the distance.

The figures I quoted were made by an experienced expert in dosimetry.
not by me

 

[SteveElbows]

Iirc reactor 3 depressurized all by itself? The plots are somewhere in the mega-thread.

I'd need to go back and check. All the same I think they think reactor 3 venting did eventually happen via the stack. Obviously another problem they had was a lack of stack instrument functionality due to power failure, so they were using crude methods such as checking the webcam for evidence of emissions from the stack.

Obviously such situations can be handled by other means. For example you could have a vent path that is under operator control and feeds into the same filter. I don't see how you can argue against the necessity of a passive venting system by pointing out that there are failure modes it doesn't address. Shall we not install curtain (lateral) airbags in cars, because there are also lots of head-on collisions?

I was not arguing against various types of venting, just pointing out some of the flaws. In an ideal world the best solution would really be to close down all old reactors that, at a minimum, have the first type of containment design which has long been recognised as being inadequate, but obviously that isn't happening.

 

[zapperzero]

I'd need to go back and check. All the same I think they think reactor 3 venting did eventually happen via the stack. Obviously another problem they had was a lack of stack instrument functionality due to power failure, so they were using crude methods such as checking the webcam for evidence of emissions from the stack.

You are right:
http://www.tepco.co.jp/en/press/corp-com/release/11031310-e.html

Unit 3:[...]
In order to fully secure safety, we operated the vent valve to reduce the
pressure of the reactor containment vessels (partial release of air
containing radioactive materials) and completed the procedure at 8:41AM,
Mar 13 (successfully completed at 09:20AM, Mar 13. After that, we began
injecting water containing boric acid that absorbs neutron into the reactor
by the fire pump from 09:25AM, Mar 13.
Taking account of the situation that the water level within the pressure
vessel did not rise for a long time and the radiation dose is increasing,
we cannot exclude the possibility that the same situation occurred at Unit
1 on Mar 12 will occur. We are considering the countermeasure to prevent
that.

At the same time, TEPCO only has plant parameters starting from June 2011 on their website.

EDIT: found a semi-useful plot on page 16 of this report:
http://www.nsr.go.jp/english/data/dai-ichi_NPS_handouts2.pdf
the red line is the actual pressure, blue line is the imagination of report authors, augmented with some software, so you can ignore it. There is a small unexplained pressure dip near the start of the plot, but other than that, it seems that it was indeed depressurized through operator actions.

 

[LabratSR]

At the same time, TEPCO only has plant parameters starting from June 2011 on their website.

Try here

https://fdada.info/EDIT: Found this there

https://fdada.info/docdata/accident_analysis/ES-Unit3-01.pdf

 

[turi]

You are right:
EDIT: found a semi-useful plot on page 16 of this report:
http://www.nsr.go.jp/english/data/dai-ichi_NPS_handouts2.pdf
the red line is the actual pressure, blue line is the imagination of report authors, augmented with some software, so you can ignore it. There is a small unexplained pressure dip near the start of the plot, but other than that, it seems that it was indeed depressurized through operator actions.

Aren't the blue and red dots the measured values and the red and blue lines the (software aided) approximations?

 

[Hiddencamper]

Iirc reactor 3 depressurized all by itself? The plots are somewhere in the mega-thread.



Obviously such situations can be handled by other means. For example you could have a vent path that is under operator control and feeds into the same filter. I don't see how you can argue against the necessity of a passive venting system by pointing out that there are failure modes it doesn't address. Shall we not install curtain (lateral) airbags in cars, because there are also lots of head-on collisions?

For a reactor to depressurize itself after its SRV accumulators have depleted means that the vessel was breached.

The SRV (safety relief valves) in GE BWRs are designed to open once or twice against 3/4 of containment design pressure. The first lift is assumed to be of all valves in relief mode, due to a load reject or MSIV fast closure. The second lift is ONLY the ADS (automatic depressurization system) valves, which then stay open until corespray comes in service to hold pressure low. About 1/2 of the valves in a BWR are ADS valves. (This is likely a little different for the unit 1 BWR, as some BWRs use EMRVs and ADS valves, but the overall concept is similar).

The accumulators are typically either 20-25 gallon (for normal SRVs) or 55 gallon (for SRVs that also utilize the ADS feature).

These air accumulators, against no containment pressure, only have a handful of lifts each. Many plants have backup air bottles, which can be used to refill and get up to 100 lifts out of the SRVs in their relief move. With no electrical power, there is no way to refill the SRV accumulators through normal means. The air lines into containment automatically isolate on a loss of power, a level 1 water level (about 2 feet above the fuel), high drywell pressure, or a loss of air pressure (they use air pressure pilot valves to hold them open).

 

[Hiddencamper]

All this talk of rupture disc problems runs the risk of downplaying the other issues that delayed venting, and also runs the risk of making it sound like venting failed at all the reactors, as opposed to the apparent reality that it was mostly reactor 2 where the failure to vent story was allowed to play out in full. Thats certainly the only reactor where they seem rather unsure as to whether the rupture disc ever ruptured, and where no torus-scrubbed venting is thought to have taken place.

Having said that, the delays to venting at reactors 1 and 3 obviously caused additional delays in pumping water into those reactors.

Certainly to my mind the problem with the rupture disc-based system seems to be due to the fact that the initial design considerations for this system were heavily focussed on the idea of containment failures due to over pressurisation. Only in later theoretical loss of coolant accident analysis papers did other modes of containment failure, such as temperature-related failure of seals and penetrations, get due attention. Really the entire system seems inappropriate not just for venting under situations where other containment failures kept containment pressure below the level necessary to rupture the discs, but also for situations where there is a desperate need to reduce containment pressure much earlier on in order to ensure that the RPV itself can be depressurised via SRV's in order to allow pumping in of water.

Need to also remember the SRVs require DC power and pressurized air to operate in their relief mode, regardless of containment pressure. There were cases at units 2 and 3 of SRVs drifting closed or failing to open, between loss of DC power or pressurized air.

Typically, venting containment during a casualty is to help you flood the containment more than flood the core.

 

[nikkkom]

For a reactor to depressurize itself after its SRV accumulators have depleted means that the vessel was breached.

The SRV (safety relief valves) in GE BWRs are designed to open once or twice

For me, an outsider, this is a shocking revelation. A revief valve which is *not* designed for at least hundreds of actuations?

The accumulators are typically either 20-25 gallon (for normal SRVs) or 55 gallon (for SRVs that also utilize the ADS feature).

These air accumulators, against no containment pressure, only have a handful of lifts each.

A relief valve which requires *consumables* to work??

 

[westfield]

I'm really referring to the standby gas treatment system, which is a combination of HEPA filters and charcoal beds. <snip>.

I'm confused. Why are you referring to the SGTS filters?

If we are talking venting via the "hardened vent" systems at fukushima I don't understand why SGTS filters are involved in the conversation. The "hardened vent" systems don't run through the SGTS filters at fukushima daiichi, they go straight to the stacks, unfiltered apart from the scrubbing from the torus water in the case of the SC "hardened vent" path or no filtering whatsoever in the case of the drywell "hardened vent" path.

The inadequacy of the SGTS in an emergency venting scenario is precisely the reason why "hardened vent" systems were retrofitted to these types of plant. It was realized early on that the ducting of the SGTS systems in these types of plant would be highly likely to fail under an emergency venting scenario and would fill secondary containment with steam and combustible gases.

 

[Hiddencamper]

For me, an outsider, this is a shocking revelation. A revief valve which is *not* designed for at least hundreds of actuations?



A relief valve which requires *consumables* to work??

They are dual function valves. The relief function is used for manual or automatic control to open the valve, and uses air pressure and DC power. You can control pressure in almost any range with the relief mode, and can blow the reactor down with these. ADS (automatic depressurization system) works by using the relief mode solenoids to lift the valve and blow down the reactor. The relief mode solenoids is actuated by either logic systems, which respond to overpressure, or manually by throwing a control switch to energize the valve solenoid directly.

The safety mode is spring loaded and typically actuates about 100 PSIG above the relief mode. The spring mode is completely passive, but can only maintain pressure around its setpoint. It only reduces reactor pressure by no more than 100 PSIG from it's lift setpoint.

For my BWR, my lowest SRV has a logic that lifts it at 1103 PSIG, and reseats it at 926 PSIG. This is the automatic relief mode, and the plant's control systems will maintain my reactor pressure between those two points for me while I manage other, more important parts of the accident (like starting or overriding ECCS, getting feedwater back, making the turbine safe, getting aux steam running, or restarting feedwater). Now if my relief mode fails, or I lose DC power/air, this same valve's safety mode (spring) lifts around 1165 PSIG, and seats when you no longer have adequate force to overcome spring pressure. It will lift to maintain my pressure between 1165 and about 1065. It's more harsh on the equipment, because it results in more valve lifts, and increases the risk of a valve getting stuck open, but it will keep the pressure vessel safe.

If I want to or need to blow down the reactor, I need air and DC power, but if I don't have those, the safety mode will protect the vessel from exceeding its ASME code limit (typically in the 1300-1400 PSIG range), even with a full power ATWS.

The reason this is considered acceptable, is because for the design basis accident, you assume you only need two valve lifts. The first lift is on the initial load reject/MSIV closure, all valves lift once in the relief mode. The non ADS valves are assumed to have utilized their entire air inventory. Then the loss of coolant/loss of feedwater/loss of high pressure injection accident would eventually pick up the logic to activate ADS, and all the SRVs that have an ADS mode would use their remaining air inventory to bring pressure down to minimum. As pressure drops, the core spray system would actuate and spray the core. Core spray actually reduces pressure greatly, and as long as core spray is running, the pressure remains low enough for all the low pressure ECCS systems to inject to the core.

tl;dr - MSIV fast closure + loss of feedwater = first SRV lift. failure of high pressure ECCS (single failure required for design basis accident) leads to low-low-low alarm level 1 water level, which automatically performs an ADS blowdown. When pressure gets low enough, the core sprayers start up to keep pressure low enough for low pressure coolant injection to run. Instrument air to refill the valves is non-safety, and is not available during the accident.

If you read GE's design basis documents, they assume 1 lift on non-ADS SRVs, 2 lifts on ADS-SRVs, then their requirements for maintaining safe shutdown are 1 core spray and 1 coolant injection ECCS to maintain the core safe once its been depressurized. After you've blown down, the SRVs no longer are required to perform a safety function.

Obviously, you can get several lifts of an SRV without refilling the air, but it will depleate over time (whether or not you use it), and it obviously depletes much faster if you use it. The ADS backup air bottles allow you to refill those accumulators for more lifts, or to blow the reactor down, if you didnt need to blowdown early in the event, but now conditions have changed and you do need it

 

[Hiddencamper]

I'm confused. Why are you referring to the SGTS filters?

If we are talking venting via the "hardened vent" systems at fukushima I don't understand why SGTS filters are involved in the conversation. The "hardened vent" systems don't run through the SGTS filters at fukushima daiichi, they go straight to the stacks, unfiltered apart from the scrubbing from the torus water in the case of the SC "hardened vent" path or no filtering whatsoever in the case of the drywell "hardened vent" path.

The inadequacy of the SGTS in an emergency venting scenario is precisely the reason why "hardened vent" systems were retrofitted to these types of plant. It was realized early on that the ducting of the SGTS systems in these types of plant would be highly likely to fail under an emergency venting scenario and would fill secondary containment with steam and combustible gases.

I'm trying to discuss what the plant ALREADY has installed to meet its design basis requirements. SGTS is not inadequate for design basis accidents, its only inadequate in an extended total loss of power with damage to your permanently installed plant systems. This means a filtered vent is not required to maintain the public safe during design basis accidents. In no case during a DBA would you need a passive filtered vent to make the plant safe. The installation of a passive filtered vent does not help you at all for any design accident, and provides very little if any net benefit. From an engineering/reactor designer perspective its more of a warm fuzzy, because you already have nuclear safety grade equipment which performs that function. (Now if we were designing a new plant, you sure as **** can bet that I would design a passive filter in, but talking about existing plants, you already have something for that)

Now for beyond design basis accidents, you have to assume going into the BDBA that all your permenently installed equipment failed. This makes sense, because in order to get into a BDBA, you had to lose all your onsite equipment. So in that case, can you actually honestly believe that the passive filtered vent will function when all your other safety grade equipment failed? The most likely strategy to succeed (in my opinion) is one which utilizes off-site portable equipment to spray and wet scrubbing. I don't see how you can, in engineering space, claim a passive filter along with all the valves, discs, etc, is any more likely to be functional than an active system. And with the EPRI study on decontamination factors for wet spray/scrubbing and the suppression pool being available (the suppression pool is a major scrubbing source, and on its own achieves the appropriate level of DF), you can meet the same quantitative goal with > 1000 DF using portable equipment which you can guarantee will function post accident.

That's my view on it as a plant design engineer.

 

[zapperzero]

This means a filtered vent is not required to maintain the public safe during design basis accidents. In no case during a DBA would you need a passive filtered vent to make the plant safe.

Circular reasoning much?

The installation of a passive filtered vent does not help you at all for any design accident, and provides very little if any net benefit.

The historical record shows that so-called "beyond design basis" accidents do happen to NPPs, with alarming frequency even (~1% of population). The assumptions built in the design basis need a bit of challenging, iow.

Now for beyond design basis accidents, you have to assume going into the BDBA that all your permenently installed equipment failed.

Nonsense. Lots of equipment was and still is functional inside Fukushima 1. Useful stuff, even - isolation condensers, vent stacks, SGTSs for at least units 1 and 3 and many other such things. And yet, as soon as the tsunami swept through, the plant was in a BDBA (or so Tepco would have us believe).

So in that case, can you actually honestly believe that the passive filtered vent will function when all your other safety grade equipment failed?

It is not a matter of belief. You can ensure the passive filtered vent will only fail in harsher conditions, by designing it properly. Fewer moving parts, simpler control logic...

The most likely strategy to succeed (in my opinion) is one which utilizes off-site portable equipment to spray and wet scrubbing.

You are assuming you can get it onsite in time. Again, recent history shows that's not always the case - causes can be as trivial as a padlocked gate, or a worker who is not able to go in an unknown radiation field.

I don't see how you can, in engineering space, claim a passive filter along with all the valves, discs, etc, is any more likely to be functional than an active system.

I can assign some non-zero probability to the event of the active system losing power and/or control...

And with the EPRI study on decontamination factors for wet spray/scrubbing and the suppression pool being available (the suppression pool is a major scrubbing source, and on its own achieves the appropriate level of DF),

What if the pool water level drops too low for effective scrubbing for some reason? Say, I dunno, too high of a temperature and pressure?

you can meet the same quantitative goal with > 1000 DF using portable equipment which you can guarantee will function post accident.

Err... what? Are you seriously suggesting that having to bring a filter from somewhere else is just as reliable a strategy as already having it onsite?

That's my view on it as a plant design engineer.

It is a bit troubling.

 

[Hiddencamper]

Circular reasoning much?


The historical record shows that so-called "beyond design basis" accidents do happen to NPPs, with alarming frequency even (~1% of population). The assumptions built in the design basis need a bit of challenging, iow.


Nonsense. Lots of equipment was and still is functional inside Fukushima 1. Useful stuff, even - isolation condensers, vent stacks, SGTSs for at least units 1 and 3 and many other such things. And yet, as soon as the tsunami swept through, the plant was in a BDBA (or so Tepco would have us believe).


It is not a matter of belief. You can ensure the passive filtered vent will only fail in harsher conditions, by designing it properly. Fewer moving parts, simpler control logic...


You are assuming you can get it onsite in time. Again, recent history shows that's not always the case - causes can be as trivial as a padlocked gate, or a worker who is not able to go in an unknown radiation field.


I can assign some non-zero probability to the event of the active system losing power and/or control...


What if the pool water level drops too low for effective scrubbing for some reason? Say, I dunno, too high of a temperature and pressure?


Err... what? Are you seriously suggesting that having to bring a filter from somewhere else is just as reliable a strategy as already having it onsite?


It is a bit troubling.

The IC was NOT functional (if it was, unit 1 would not have failed in a few hours, as the IC had some ridiculous amount of cooling available to it). The IC would have provided adequate core cooling to unit 1, and could have been easily made up with a simple fire pump truck. Remember, the IC has to be cycled on and off under normal conditions to prevent exceeding the 100 degree F per hour cooldown rate (I think its 40 deg C/hr for Japan) for the vessel. It was cycled off when power was lost. The internal isolation valves are DC valves, and the external valves are AC valves. The operators typically cycle the IC discharge outboard isolation open and closed to turn the IC on and off. On the loss of power, the IC was already in the off position, and it is believed that some of the other valves may have went closed under an invalid isolation signal during the flooding of the safety related MCCs. The operators erroneously determined the IC was functioning when it wasnt. They possibly could have sent an operator to the IC outboard isolation valves, manually opened them, and got cooling, but they didnt even think to try (unfortunately they were not well trained on the system, and nobody in the control room at the time had any experience using it). So yes, it may have been capable of helping, but it was not functional at the time of the event, and may not have been available even.

SGTS was not functioning at Fukushima (need power to open the dampers, need power for the pre-heaters and dehumidification). First, the SGTS has fail close dampers on the containment isolation side. these dampers are controlled with fail close hydramotors. Hydramotors are throttleable or 2 position hydraulic actuating units for positioning valves and dampers. When a hydramotor loses power, its relief solenoid loses power, which drains pressure from the accumulator, and causes the valve or damper to fail to a specified state on loss of power. SGTS containment isolation valves fails closed on loss of power.

The vent stack is not an active component.

The rupture disks failed to break, those are passive components (which goes to show that you cannot count on your on-site equipment)

As for the suppression pool level, in a normal accident you have RHR to remove heat from it. In a beyond design basis accident, you can lose level if the suppression pool itself (torus) breaks. You can deal with this by flooding the basement using portable or fire pumps, which, while you lose containment capability, you still have wet scrubbing capability. A passive vent wouldn't help you in this case as your pressure boundary broke. As for temp/press, remember that pool is an enclosed system. Inventory has no place to go while the system is sealed. It's not going to just disappear for Mark I/II containments (for Mark III containments, it can lower due to the very large volume of the containment. Mark III plant designs utilize passive gravity fed suppression pool makeup systems to deal with that, which will automatically dump when pool level drops about 4 feet, to ensure the drywell vents are adequately covered). The only time level should lower is when you are venting the wetwell, which does require a hardened vent, but the effluents have already been scrubbed by the pool, and by venting the containment you can now make up the pool to maintain your scrubbing.

As for the 'circular reasoning'. I don't see circular. During DBAs (things that are IN the design basis), your active filtering is all that you need. You don't NEED a passive vent. That's not circular at all, its saying active is already installed and works, passive could also work but you don't need it. Passive filters only have added benefits for beyond design accidents, which I argue you might not even have them because whatever nasty accident took out your active systems could have damaged your passive filter as well.

 

[zapperzero]

The IC was NOT functional

http://www.tepco.co.jp/en/nu/fukushima-np/images/handouts_111122_03-e.pdf
says otherwise. Both trains available and functioning, but not at full capacity

The operators erroneously determined the IC was functioning when it wasnt.

According to the document above, the operation was confirmed by observing steam coming out of the appropriate place.

They possibly could have sent an operator to the IC outboard isolation valves, manually opened them, and got cooling, but they didnt even think to try.

" At 21:30. the operator conducted open op
eration of valve 3A and confirmed
generation of steam. "
(from the same cited document)

SGTS was not functioning at Fukushima (need power to open the dampers, need power for the pre-heaters and dehumidification).

I... what? The point I was making was that it was not damaged in any way - yet did not get used in the event.

The rupture disks failed to break

What rupture disks?

As for the suppression pool level, in a normal accident you have RHR to remove heat from it. In a beyond design basis accident, you can lose level if the suppression pool itself (torus) breaks. You can deal with this by flooding the basement using portable or fire pumps, which, while you lose containment capability, you still have wet scrubbing capability. A passive vent wouldn't help you in this case as your pressure boundary broke. As for temp/press, remember that pool is an enclosed system. Inventory has no place to go while the system is sealed. It's not going to just disappear for Mark I/II containments (for Mark III containments, it can lower due to the very large volume of the containment. Mark III plant designs utilize passive gravity fed suppression pool makeup systems to deal with that, which will automatically dump when pool level drops about 4 feet, to ensure the drywell vents are adequately covered). The only time level should lower is when you are venting the wetwell, which does require a hardened vent, but the effluents have already been scrubbed by the pool, and by venting the containment you can now make up the pool to maintain your scrubbing.

You don't like temp/pressure? Fine. Let's say an earthquake damaged a steam downcomer, so that there is now a big crack in it, above the water level? Now you can't scrub your steam, although there is plenty of water.
I have a lot of doubt about your claim that the wetwell provides sufficient scrubbing, too. I seem to remember dramatic spikes in readings of the counters at plant boundary, corresponding to venting operations.
The operators of the plant were not convinced either, as I recall there was much wringing of hands before venting was even attempted, as there was explicit concern at TEPCO over the pace/effectiveness of the evacuation effort. Venting was delayed too much, actually.

As for the 'circular reasoning'. I don't see circular. During DBAs (things that are IN the design basis), your active filtering is all that you need. You don't NEED a passive vent. That's not circular at all, its saying active is already installed and works, passive could also work but you don't need it. Passive filters only have added benefits for beyond design accidents, which I argue you might not even have them because whatever nasty accident took out your active systems could have damaged your passive filter as well.

You are basing this belief on what, exactly?

 

[Hiddencamper]

http://www.tepco.co.jp/en/nu/fukushima-np/images/handouts_111122_03-e.pdf
says otherwise. Both trains available and functioning, but not at full capacity


According to the document above, the operation was confirmed by observing steam coming out of the appropriate place.


" At 21:30. the operator conducted open op
eration of valve 3A and confirmed
generation of steam. "
(from the same cited document)


I... what? The point I was making was that it was not damaged in any way - yet did not get used in the event.


What rupture disks?


You don't like temp/pressure? Fine. Let's say an earthquake damaged a steam downcomer, so that there is now a big crack in it, above the water level? Now you can't scrub your steam, although there is plenty of water.
I have a lot of doubt about your claim that the wetwell provides sufficient scrubbing, too. I seem to remember dramatic spikes in readings of the counters at plant boundary, corresponding to venting operations.
The operators of the plant were not convinced either, as I recall there was much wringing of hands before venting was even attempted, as there was explicit concern at TEPCO over the pace/effectiveness of the evacuation effort. Venting was delayed too much, actually.



You are basing this belief on what, exactly?

I really appreciate a link. I want to add first off, if the IC was truly functional, unit 1 would not have had an accident. I also want to add that the official report from Japan's national diet concludes that the "IC systems were acknowledged to have largely lost their cooling function." (see page -80- of the following link). That is non-functional. Just like how HPCI was non-functional at unit 1, due to the loss of electric power causing the system to be failed in a state where it could not operate, IC was also failed at unit 1, due to the loss of electric power causing the system to be failed in a state where it could not operate.

http://www.cas.go.jp/jp/seisaku/icanps/eng/02Attachment1.pdf

They additionally state, in their report on the accident, that "The other isolation valves, which had been fully open until that time, were fully or almost fully closed as a result of the fail-safe function triggered by total loss of AC and DC power." On page 34 of the following link.

http://www.cas.go.jp/jp/seisaku/icanps/eng/03IIfinal.pdf

There's also the fact that if you stopped IC for long enough, you lose the ability to have natural circulation due to the generation of various gases and the like. But that's neither here nor there, I just know about this because I know Oyster Creek has safety analysis about it.

With regards to rupture disks, I'm talking about unit 2's rupture disc not operating, which made the accident at unit 2 worse than it needed to be. You can see this on the validated timeline in INPO 11-05 (the publicly available US industry document on the accident), which states on 13-Mar at 1100, the rupture disk failed to break. This is important because a rupture disk would be the primary method of activating a passive filter. Again on the 14th at 1130, they could not break it. Later on the 15th around midnight, when pressure was 40 psi above the rupture disk break pressure, it still did not break.

As for an earthquake damaging a downcomer. Are we talking about a steam downcomer or an SRV downcomer? For a steam downcomer, that's primarily important for LOCA, or immediately after the vessel breaches. With a broken steam downcomer, you would have already opt to flood the containment due to the loss of all ability to cool the core, which would obviate the need for it. The steam downcomers are designed to ensure high pressure/temperature steam is vented to the suppression pool for quenching, to prevent containment damage. If your downcomer breaks, you are likely to damage your containment due to the loss of pressure suppression capability, and you would end up breaching it, making your passive filter useless, and wet spraying and scrubbing, along with containment flooding, more useful.

Spikes in radiation measurements will happen, when you melt fuel, and that fuel then melts through the vessel into the drywell, where it then causes over pressure, such that you now have escaping noble gas inventory being ejected. Appropriate response with portable pumping systems would have directed containment drywell injection prior to the hot debris ejection event (my plant's SAMGs do, and they are nearly identical to every US BWR). Spraying would also be in progress through portable pumps. Ideally though, you would have used your portable equipment to prevent the core damaging event in the first place, but even assuming you failed at that (maybe because your SRVs were depleted...), running containment spray using portable equipment, venting from the wetwell (not the drywell) initially and making use of the vacuum breakers to siphon drywell radionuclide inventory through the pool, those would be useful. There are some cases where drywell filtering may be needed, and the NRC agrees with that, but it's not the only way to skin the cat.

Fully agree venting was delayed much too much though. Unfortunately they did not have the resources, plans, training, or equipment to handle a multi-unit event of this magnitude.

 

[zapperzero]

I really appreciate a link. I want to add first off, if the IC was truly functional, unit 1 would not have had an accident. I also want to add that the official report from Japan's national diet concludes that the "IC systems were acknowledged to have largely lost their cooling function." (see page -80- of the following link). That is non-functional. Just like how HPCI was non-functional at unit 1, due to the loss of electric power causing the system to be failed in a state where it could not operate, IC was also failed at unit 1, due to the loss of electric power causing the system to be failed in a state where it could not operate.

http://www.cas.go.jp/jp/seisaku/icanps/eng/02Attachment1.pdf

They additionally state, in their report on the accident, that "The other isolation valves, which had been fully open until that time, were fully or almost fully closed as a result of the fail-safe function triggered by total loss of AC and DC power." On page 34 of the following link.

http://www.cas.go.jp/jp/seisaku/icanps/eng/03IIfinal.pdf

This refers to the time immediately after the tsunami hit, flooding junction boxes.
But on the very same page they also state:

"It cannot be determined that, between the time of the earthquake and the arrival of the tsunami, there was such damage to the IC lines and tanks as to degrade the cooling function of the IC"

The operator (as stated in the document cited by me above), did later manage to open valve 3A and to confirm that the IC 1 was functioning.

With regards to rupture disks, I'm talking about unit 2's rupture disc not operating, which made the accident at unit 2 worse than it needed to be.

You can see this on the validated timeline in INPO 11-05 (the publicly available US industry document on the accident), which states on 13-Mar at 1100, the rupture disk failed to break. This is important because a rupture disk would be the primary method of activating a passive filter. Again on the 14th at 1130, they could not break it. Later on the 15th around midnight, when pressure was 40 psi above the rupture disk break pressure, it still did not break.

from the document you cite:

"The motor-operated containment vent valve (MO - 271) was opened at 0810 on March 13 (T plus 41.4 hours). At the time, containment pressure indicated approximately 50.8 psia (0.35 MPa abs). At 0855, indicated containment pressure reached 52.9 psia (0.365 MPa abs), below the design pressure of 55.1 psig (0.38 MPa gauge), then began to lower. The venting lineup was not yet complete. At 1015 (T plus 43.5 hours), the site superintendent directed operators to vent the Unit 2 containment (see Figure 7.4- 5). Workers used the small generator in the control room, which had been installed to restore some lighting, to energize the solenoid for the large air-operated suppression chamber vent valve (AO-205). At 1100 (T plus 44.2 hours), the vent lineup was completed but indicated containment pressure was lower than the 62 psig (427 kPa gauge)
pressure necessary to open the rupture disk and allow venting, and the rupture disk remained intact."

It worked as intended, iow. Perhaps the set point was too high, yes. This does not in any way invalidate the principle of using a rupture disk...
Later on, we have active equipment failing for lack of power, despite emergency equipment having been brought on site and activated:

"On March 14 at 1101 (T plus 68.3 hours), a hydrogen explosion occurred in the Unit 3 reactor building. The explosion damaged the temporary power supply used to open the Unit 2 suppression chamber vent valve (AO-205), causing the valve to fail closed."

Even later, due to various events, the rupture disk again functions as intended:

"at 2100 (T plus 78.2 hours), operators opened the small suppression chamber air - operated vent valve (AO - 206), establishing the venting lineup (other than the rupture disk).
Indicated containment pressure remained slightly lower than the 62 psig (427 kPa gauge) working pressure of the rupture disk, so venting did not occur."

even later, just before the unexplained explosion-like event, we have another failure of powered equipment:

"Two minutes after midnight on March 15, the operators opened the small air-operated drywell vent valve (AO - 208). The vent line lineup was complete, except for the rupture disk that remained closed.
Containment pressure remained stable at approximately 109 psia (750 kPa abs). The operators rechecked their lineup and found that the small air-operated drywell vent valve had already failed closed."

welp.

As for an earthquake damaging a downcomer. Are we talking about a steam downcomer or an SRV downcomer?
For a steam downcomer, that's primarily important for LOCA, or immediately after the vessel breaches. With a broken steam downcomer, you would have already opt to flood the containment due to the loss of all ability to cool the core, which would obviate the need for it. The steam downcomers are designed to ensure high pressure/temperature steam is vented to the suppression pool for quenching, to prevent containment damage. If your downcomer breaks, you are likely to damage your containment due to the loss of pressure suppression capability

Unless there was some means to vent safely...

Spikes in radiation measurements will happen, when you melt fuel, and that fuel then melts through the vessel into the drywell, where it then causes over pressure, such that you now have escaping noble gas inventory being ejected. Appropriate response with portable pumping systems would have directed containment drywell injection prior to the hot debris ejection event (my plant's SAMGs do, and they are nearly identical to every US BWR). Spraying would also be in progress through portable pumps. Ideally though, you would have used your portable equipment to prevent the core damaging event in the first place, but even assuming you failed at that (maybe because your SRVs were depleted...), running containment spray using portable equipment, venting from the wetwell (not the drywell) initially and making use of the vacuum breakers to siphon drywell radionuclide inventory through the pool, those would be useful. There are some cases where drywell filtering may be needed, and the NRC agrees with that, but it's not the only way to skin the cat.

Other ways may or may not be practicable, as it turns out. Venting through the wetwell can fail (as shown at Fukushima 1-2) repeatedly, portable pumps can be damaged by wholly unrelated events, radiation levels can be to high in the vicinity of manually-operated valves etc etc. These are all things that have happened at Fukushima, not hypotheticals. You propose that there is no need to have a way to deal with them happening again elsewhere.

Fully agree venting was delayed much too much though. Unfortunately they did not have the resources, plans, training, or equipment to handle a multi-unit event of this magnitude.

If only there had been a way for the venting to take place safely, without operator intervention and in the absence of power and instrumentation!

 

[nikkkom]

I really appreciate a link. I want to add first off, if the IC was truly functional, unit 1 would not have had an accident.

By now it is firmly established that IC did not save Unit 1 because TEPCO never considered extended SBO, including EDG failure, to be possible.

To be more precise: (1) operators had no training what to do, and accident manuals had no description what to do in such situation, and (2) valves leading to/from IC weren't designed so that they don't fail close without power in a way which makes impossible for them to be opened, even manually.

Both of these errors are easy to fix.

 

[nikkkom]

Now for beyond design basis accidents, you have to assume going into the BDBA that all your permenently installed equipment failed. This makes sense, because in order to get into a BDBA, you had to lose all your onsite equipment.

It reads as if you are unaware about Fukushima disaster. Which can't be true, so you must be willfully ignoring it. Pretty scary. It means that you need to experience another meltdown, somewhere in US this time, to see the light.

"All your permenently installed equipment" does not need to fail for plant to get into a serious accident. It is enough for it to be merely without power! IT IS EMPIRICALLY PROVEN NOW! How many Fukushimas need to happen for you to admit it?

Fukushima had shown that there must be passive systems, which need no power and no operator intervention at all, or can be actuated manually (meaning with bare hands, as in a valve which can be opened by rotating a handle). Filtered vent is one such system.

Why do you fight it? Because, gasp, it needs some significant paperwork?? THAT is more important than preventing thousands of square miles and millions of people from being dusted with Cs-137?

 

[Rive]

If only there had been a way for the venting to take place safely, without operator intervention and in the absence of power and instrumentation!

It's an interesting question.

With the leaks, the pressure (therefore: the boiling point) were kept high and so the main of the water were still there to act as a heat puffer. Heat were removed with high pressure, high temperature steam.

With a vent through a rapture disk the boiling point would fall to 100 degree -> almost all the water would had gone within hours, at low pressure, low temperature (therefore along with much less heat).

PS.: BTW the first cask left U4.
http://translate.googleusercontent.com/translate_c?depth=1&hl=en&ie=UTF8&prev=_t&rurl=translate.google.com&sl=ja&tl=en&u=http://photo.tepco.co.jp/date/2013/201311-j/131121-01j.html&usg=ALkJrhg1vKOQzhihL7wG9YCHGgHS-d0B7Q

 

[zapperzero]

http://photo.tepco.co.jp/library/131121_01/131121_03.jpg
is censored
what the heck?

 

[Rive]

On every photo (in the previous galleries too) some parts of the casks and the handling machinery is blurred because of some safety reasons.

(Sounds stupid for me, actually.)

 

[Hiddencamper]

It reads as if you are unaware about Fukushima disaster. Which can't be true, so you must be willfully ignoring it. Pretty scary. It means that you need to experience another meltdown, somewhere in US this time, to see the light.

"All your permenently installed equipment" does not need to fail for plant to get into a serious accident. It is enough for it to be merely without power! IT IS EMPIRICALLY PROVEN NOW! How many Fukushimas need to happen for you to admit it?

Fukushima had shown that there must be passive systems, which need no power and no operator intervention at all, or can be actuated manually (meaning with bare hands, as in a valve which can be opened by rotating a handle). Filtered vent is one such system.

Why do you fight it? Because, gasp, it needs some significant paperwork?? THAT is more important than preventing thousands of square miles and millions of people from being dusted with Cs-137?

I agree that you can get into a significant accident without all of your permanently installed equipment failing, however, the definition of a BDBA that we are required by regulation to design to requires us to assume that we've lost pretty much all on site permanently installed equipment. That's the starting point from the regulator's perspective. Even many passive components are assumed to fail due to the extreme common mode failure phenomenon. I'm looking at this from the perspective of what the regulator is requiring us to design to. Under the assumptions we are given, if I put something in that doesn't meet the regulator's definitions, that means its not going to "work" in a severe accident. Remember all of this goes back to what the regulator is willing to accept, and when the regulator starts off from day 1 saying that a BDBA assumes failure of all onsite permanent equipment with certain exceptions, that's how you have to start from a design perspective. For example, under those definitions and regulations, a passive filter on its own may not work, or even worse, it could affect how my plant responds to its design basis accidents.

Unfortunately what you think things should be, and the practical side of things, do not work that way in nuclear.

side bar: I take personal offense to your claim that I would need to experience another meltdown. I'm here like you all are to have discussions about Fukushima, nuclear accidents, etc, and the hostility here is ridiculous.

 

[nikkkom]

I'm looking at this from the perspective of what the regulator is requiring us to design to.

I propose that you look at it from a different perspective: "let's assume that some failure modes *demonstrated* at Fukushima may happen on our plants too".

This includes:
- unfiltered venting
- failure to vent before meltdown
- inadequate radiometers
- lack of robust emergency lighting
- ...(read accident destription and you can easily add a few more)...

If I would see nuclear industry taking active steps to improve the areas which *demonstrably* failed (we don't need to guess "can this happen?" - we KNOW it can), I would have hope the industry has a future.

I'm seeing something else instead: pointless discussions "what is 'design basis accident' and what is beyond that" (as if anyone cares how you call it!), procrastination with implementing even the most obvious fixes...

Frankly, by this point I prefer endless fields of PV panels in Arizona to this mess.

 

[zapperzero]

I agree that you can get into a significant accident without all of your permanently installed equipment failing, however, the definition of a BDBA that we are required by regulation to design to requires us to assume that we've lost pretty much all on site permanently installed equipment.

There's something like a cognition bug here. You design to deal with design basis accidents, by definition. Beyond design basis means just that, stuff that your design isn't really expected to deal with. At the very most, you can make an effort to fail gracefully, whatever that means in context...

Now, the class of design basis accidents SHOULD include Fukushima-like events (prolonged loss of onsite power, both DC and AC) but doesn't, apparently. This doesn't seem to trouble you at all.

side bar: I take personal offense to your claim that I would need to experience another meltdown. I'm here like you all are to have discussions about Fukushima, nuclear accidents, etc, and the hostility here is ridiculous.

When you come up with gems like this, it's also a bit justified:

Unfortunately what you think things should be, and the practical side of things, do not work that way in nuclear.

Unfortunately? That's all you have to say?

 

[LabratSR]

side bar: I take personal offense to your claim that I would need to experience another meltdown. I'm here like you all are to have discussions about Fukushima, nuclear accidents, etc, and the hostility here is ridiculous.

Well I, for one, really appreciate your input.

 

[nikkkom]

It's an interesting question.

With the leaks, the pressure (therefore: the boiling point) were kept high and so the main of the water were still there to act as a heat puffer. Heat were removed with high pressure, high temperature steam.

With a vent through a rapture disk the boiling point would fall to 100 degree -> almost all the water would had gone within hours, at low pressure, low temperature (therefore along with much less heat).

Boiling of water takes about the same amount of energy as heating it to 500 C. You don't lose that much cooling capacity by letting it boil at lower temperature.

If water would have been allowed to boil inside RPVs at low pressure, it would still cool them (as long as it lasted), and the steam would have drastically fewer contaminants (basically, it would be on par with usual BWR first loop water).
After it all boiled away, the fuel would melt "dry", not generating highly contaminated steam as was observed oozing out of Unit 3 for days.

(In truth, since even depressurized RPVs would have some residual water in the lower head, there still would be highly contaminated steam after fuel melts, just less of it).

 

[etudiant]

Well I, for one, really appreciate your input.

Add me to that list as well, hiddencamper!
Your inputs, coming from a base of real world experience rather than theory/ideal, are a major resource.
Don't let it get to you when others lash out, they are simply frustrated by the regulatory and hardware mechanisms.

 

[Most Curious]

Hiddencamper, add me to the list of those who appreciate your perspective and "insider" knowledge of plant systems and design.

I am here to learn and understand, mostly from folks far more knowledgeable than myself, not to sort through anti-nuclear agendas. The media does a fine job of presenting inaccurate agenda driven gibberish so no more is needed.

I will certainly agree some things should have been done differently at Fukushima such as locations of emergency power systems, better use of the IC of unit 1 and better operator training. (Which can ALWAYS be improved at any major installation, nuclear or not.) Clearly, dispersion of hydrogen from melting reactors did not go as planned, adding greatly to plant damage and in radioactive release. Of course, had the meltdowns been prevented, many of these issues would not have come up at all...

Those of us not really in the know as to how changes in plant design are made can only guess at the complexity added by regulation - as only ONE factor. Hopefully, there will be some value come from the comments made by those of us too ignorant to know what can't be done - much like the bumblebee being ignorant of the fact it can't fly.

I have no doubt the events of Fukushima will be studied for at least the next century; lessons will be learned and improvements made. That is great and everyone will pat themselves on the back for a job well done - until the next event which will point out additional areas in need of improvement, and possibly indicate previous "improvements" weren't such a good idea after all. So goes the scientific process and increase in mankind's knowledge, long after personal agendas are forgotten.

To gain maximum benefit for us all in matters nuclear, we need to attract folks directly involved in the industry to make the rest of us more useful in adding our limited brainpower to so many complex problems. In order to attract the comments from such people we need to welcome them and try not to be offensive. In reality, they are here for the same reason as the rest of us - to learn. We should be pleased they are also willing to educate the rest of us.

Just my 2 cents worth.

 

[a.ua.]

Well I, for one, really appreciate your input.

+1...

 

[Hiddencamper]

Sorry to make a rift here everyone.

As for zapper, you mention that I don't seem concered enough that a "fukushima" accident isn't in the DBA.

First off, if I got to a Fukushima accident, it probably means my DBA probably wasn't determined correctly. The DBA is supposed to include the worst case environmental impacts to the plant. So if I got to Fukushima, then it means that I never determined my DBA right. I then have to ask, how do I know putting a "Fukushima" accident in the license requirements is going to actually cover a Fukushima accident, when I couldn't even determine my normal accidents correctly. This is why Fukushima needs to be covered as a beyond design accident.

The DBA for a nuclear plant is essentially as follows: Worst case initial conditions (reactor overpower, lowest lake level, hottest temperaturs, lowest emergency generator fuel storage, etc etc), all safety systems in service, initiating accident, single limiting failure, no human action for 30 minutes, plant is automatically stabilized/made safe, cold shutdown achieved within 36 hours and maintained for 30 days. No core damage if it is an anticipated event. Minimal release is allowed for abnormal events (once in the life of the plant type events). Only postulated events like a LB-LOCA allow for any fuel damage or release approaching the limits of your license.

A fukushima accident requires assumptions that go far beyond the DBA definition. As such, it really fits in with the other accidents, that are non-DBA. Examples of these are station blackout and ATWS. Things that have a high liklihood of occurring, or an unacceptably high consequence if it did occur. Under beyond dba, my initial conditions are what the regulator tells me. Unlike a DBA, I don't need to use the most limiting conditions, instead I only need to demonstrate reasonable assurance that I can protect against the event. This means I'm allowed to use portable equipment, manual operator actions, I'm allowed to assume I start from realistic conditions, I'm allowed to violate my operating license (if it is required for the health and safety of the public), I'm allowed to repurpose equipment as necessary. The goal is to meet the requirement of the accident. For SBO, I have to survive my coping time without violating any design limits of the plant. For BWR ATWS, I have to be able to reduce power independent of the scram system to a point where the plant can survive without violating its safety or design limits long enough for boron injection to complete. My initial conditions and success criteria of the event are what the regulator tells me.

A Fukushima event requires something beyond the definition of the DBA to get there. It fits in best with the select DBAs which have a high liklihood or consequences.

As for DBAs and design criteria for plants, I personally am a huge fan of re-validating, using present day methods, the DBAs for all plants. In the US, plants are revalidating their seismic/structural/flooding, and I think that's a huge step in the right direction. If Fukushima has shown us anything, its that as your methods change, you may find hazards you did not originally expect (or design for)

 

[rmattila]

Just to keep some international perspective, here's my post from two years ago:

Even though the design bases in pretty much all Western nations were initially based on the NRC:s criteria from the 1960's, the definitions have since diverged.

Here in Finland, for example, severe accidents were included in the design bases in the 1980's, with specific criteria for failure assumptions (pretty much all "normal" safety systems and instrumentation assumed lost), containmet loads, equipment qualification for the core meltdown conditions, allowable releases (100 TBq Cs-137) etc., and backfittings (filtered ventings, passive containment flooding systems etc.) were made at the old plants. For new plants, a more robust core catcher has been required since the early 1990's.

A more recent development has been a systematic approach to so called "design extension conditions" (DEC), which were outside the original design bases. These conditions include e.g. situations with a common cause failure in any of the safety systems, other complex accident sequences or very rare natural events, and the category has its own design rules and acceptance criteria (to be demonstrated when applying a construction or operating permit and ever 10 years during operation).


So all in all, the design basis of plants consists of three event categories based on the conservatively estimated frequency of the initiating event:

1. the "old-fashioned" design basis conditions
DBC1, normal operation
DBC2, anticipated operational occurrences, f > 1e-2/a
DBC3, Class 1 postulated accidents, 1e-2/a < f < 1e-3/a
DBC4, Class 2 postulated accidents, f < 1e-3/a

2. Design extension conditions, events with an estimated frequency between 1e-4/a and 1e-7/a
DEC A, DBC2-3 with a CCF in a safety system
DEC B, complex accident sequence (=multiple failures)
DEC C, very rare events (such as a collision of a large passenger aircraft)

3. Severe accidents, events exceeding the acceptance criteria for DECs
total sum of all severe accident even trees shall be lower than 1e-5/a.

Summing up, the cutoff frequency for events to be considered in the design is of the order of 1e-7, and there's the additional reuirement that the sum for all such events shall be lower than 1e-5. And the severe accident systems shall be able to fulfill their design basis so that the probability for exceeding the acceptance criteria for severe accidents is lower than 5e-7/a.

Since all these event categories contain explicit design rules and acceptance criteria, it is natural to include them all in the concept "design basis" of the plant. I have the impression that many other countries are also taking steps in this direction, so it may become internationally more common to redefine the "design basis" to go beyond the traditional DBC2-4 events with a single (or double in some countries) failure.

 

[Hiddencamper]

Just to keep some international perspective, here's my post from two years ago:

fascinating thank you!

 

[Hiddencamper]

Hey guys, with Fukushima pulling spent fuel out, there is a lot of noise in the media about "inadvertent criticality" in the SFP.

As far as I know, the SFPs utilize boron plated racks and the fuel assemblies are positioned to ensure keff < 0.95 at all times. I mean, in all seriousness, not only should it not occur, but even if there was a threat, that could be dealt with simply by adding boron to the SFP inventory prior to moving rods.

What doesn't make any logical sense, is the fact that they are claiming that removing the rods may cause criticality. Yes moving rods means you are shifting the local reactivity profile, but the overall net effect of removing a rod would be to reduce reactivity in that cell of the SFP. I also don't see how removing or moving any individual fuel assembly would be capable of defeating that < 0.95 keff in the SFP. The fuel had to be placed in that position originally, so removing it should not put you even close to such an event. So logically it seems completely out of the picture.

What I'm curious is if anyone knows where this claim came from. Was this simply the type of stuff from Arnie Gunderson or a handful of others who have a tendency to exaggerate claims, or was there some official source that mentioned there was a possibility here? Has TEPCO or any official agencies (Japan or international) mentioned anything here?

In my searches, I haven't seen anything, and I think it's likely a quack claim, but I'm interested in seeing where it came from.

Thanks

 

[LabratSR]

TEPCO video of fuel extraction from Fukushima Daiichi pool 4

http://youtu.be/rLvu57X474M

 

[jadair1]

Hey guys, with Fukushima pulling spent fuel out, there is a lot of noise in the media about "inadvertent criticality" in the SFP.


Tepco is not removing spent fuel rods at this time, they are currently removing unused assemblies and this should take until some time in the new year.

As far as I know, the SFPs utilize boron plated racks and the fuel assemblies are positioned to ensure keff < 0.95 at all times. I mean, in all seriousness, not only should it not occur, but even if there was a threat, that could be dealt with simply by adding boron to the SFP inventory prior to moving rods.

There is some speculation that the Boron plated racks have been degraded by salt water and high heat in the pool, I do not know how credible this information is.

What doesn't make any logical sense, is the fact that they are claiming that removing the rods may cause criticality. Yes moving rods means you are shifting the local reactivity profile, but the overall net effect of removing a rod would be to reduce reactivity in that cell of the SFP. I also don't see how removing or moving any individual fuel assembly would be capable of defeating that < 0.95 keff in the SFP. The fuel had to be placed in that position originally, so removing it should not put you even close to such an event. So logically it seems completely out of the picture.

I don't believe anyone is claiming that the succseful removal of rods will increase the chance of a criticallity in the pool, rather it is the chance that an unsuccessful extration could lead to a criticality some how.

What I'm curious is if anyone knows where this claim came from. Was this simply the type of stuff from Arnie Gunderson or a handful of others who have a tendency to exaggerate claims, or was there some official source that mentioned there was a possibility here? Has TEPCO or any official agencies (Japan or international) mentioned anything here?

Several sources have made this claim, some more credible than others. Do you actualy believe Tepco or the NSA are credible sources?


In my searches, I haven't seen anything, and I think it's likely a quack claim, but I'm interested in seeing where it came from.

Thanks

I'm sorry, it's late and I'm tired so I'm not going to find the links for you but it is not only Gunnersan and Busby making these claims but several other nuclear engineers with experience in spent fuel.

 

[Rive]

Boiling of water takes about the same amount of energy as heating it to 500 C. You don't lose that much cooling capacity by letting it boil at lower temperature.

You lose mass as it boils (and goes away). That lost mass will carry away heat belonging to only 100 C, instead of 500C. It's quite a difference, especially because the difference will boil away even more water at low temperature.

What I'm curious is if anyone knows where this claim came from. Was this simply the type of stuff from Arnie Gunderson or a handful of others who have a tendency to exaggerate claims, or was there some official source that mentioned there was a possibility here? Has TEPCO or any official agencies (Japan or international) mentioned anything here?

In my searches, I haven't seen anything, and I think it's likely a quack claim, but I'm interested in seeing where it came from.

Thanks

As I recall, in the early days it was considered as a worst case scenario: if the pools are partially out of water then the cladding gets fire and the fuel breaks down. Then the damaged geometry might lead to criticailty in the remaining water. At that point it was taken seriously.
As the water level was secured, the story evolved to the 'the whole building breaks down and so' stage as the catastrophe it would cause was too tempting to let it drop -> the story went gundersened.



From officials, as I recall the possibility of criticality of U3 pool (where the fuel geometry might be severely damaged) in the early days is still on the table. This idea also has some supporting evidences, as thermal images and the excess radiation measured above the upper parts of U3.

I, personally, think that there are other explanations too. However as the geometry there might be affected by removing the debris, this line should not be dropped easily.

 

[nikkkom]

You lose mass as it boils (and goes away). That lost mass will carry away heat belonging to only 100 C, instead of 500C.

Not quite. It will carry away thermal heat *and* "heat" of vaporization, which is quite substantial for water.

More to it: as water heats up, heat of vaporization goes down (eventually reaching zero at critical temperature where difference between liquid and gas disappears). So to boil a liter of water at 200 C does not take as much energy as to do so at 100 C.

 

[rmattila]

I suppose the criticality issue arises from someone trying to explain the concept of criticality to a listener who has watched too much Ghostbusters, and hence the "if two rods touch each other" meme was born..

Qualitatively, there are three issues that might reduce the margin to criticality: loss of absorber plates in between the assemblies, grid distortion and dropping of an assembly on top of the rack. Based on the footage,#1 is not an issue in unit 4. And in the case of dense racks, the magnitude of #2 and 3 is at most a few hundred pcm, insignificant compared to the min 5000 pcm initial margin. I don't know about the Japanese practice, but at least here the 5000 pcm margin is required even if the entire rack was filled with the most reactive one-year-spent assemblies; so in real life, there probably is a few thousand pcm extra margin.

 

[nikkkom]

Hey guys, with Fukushima pulling spent fuel out, there is a lot of noise in the media about "inadvertent criticality" in the SFP.

I think it is usual uninformed hysterionics.