nuclear engineer here. going through this document, there are a lot of things which are just flat out incorrect or misleading. I don't disagree that some parts of the AP1000 plant may need some more looking, but the article is not descriptive enough over what specific conditions there are issues with. Having read the entire design control document series for the AP1000 and knowing what type of organization fairwinds is (run by Arnie Gunderson, not the most reputable source for factual nuclear information), I'm a little skeptical.
First, the "radiation" didnt make the burning. If you were getting enough radiation to feel a 'burn' you would be losing your hair right now.
To put the numbers in perspective, you received a maximum of 1-1.5 Rem (10-15 mSv) of radiation. The average yearly background exposure not including medical is ~320 mRem/year (3.2mSv). Including medical: ~620 mRem (6.2 mSv). For a nuclear power plant worker the maximum allowed per year is 5 Rem (50 mSv). On average nuclear power plant workers get an additional 100mRem/year (1mSv), except for outage workers who average about an extra 300 mRem/year (3mSv). People working at Fukushima after the accident were authorized to receive 25 Rem (250mSv) to protect the plant and the public.
There is no statistical increase in liklihood of cancer until you pass 10 Rem (100mSv). Even using Linear No Threshold which is the most conservative accepted estimate and is used by the government for calculating deaths from radiation exposure to assign value to radiation, your increase in risk is.1% over your entire lifetime due to that procedure.
Again just FYIs and facts.
And none of those had to do with the age of the plant.
Additionally, probabilistic risk assessment (which is what generates the number 1 accident in X years) is a model, and a tool, used to determine what is the most safety significant components which you need to maintain. A tool, nothing more. People cite numbers believing that they are gospel truth that a plant will only have 1 accident every X years, but it is only a tool.
Additionally, PRA depends on your design basis being correct from the start. PRA does not model operator error very well as what happened in TMI (multiple operator errors combined with process errors, multiple latent failures, etc), and does not model bad designs like chernobyl. PRA also can't account for an incorrect design basis, like only 14ft tsunami vs. 45ft.
tl;dr, age of the equipment and strucutres has had nothing to do with any nuclear accident. Only human error at some level of the process (design, operation, etc).
In BWRs, the HPCI system (in BWR 2- some of the 5s) and RCIC system (found in BWR4+) are steam driven. Fukushima Daiichi unit 2 was cooled for 70 hours on RCIC using steam alone, and unit 3 on RCIC and HPCI for 36 hours. The limiting issue is cooling water for the pump, because the cooling water is the same water that is going in and out of the reactor and suppression pool, eventually the whole mass heats up beyond the maximum temperature of the pump.
PWRs have aux feedwater which is steam driven or direct diesel driven (or both depending on design) to provide feedwater to the steam generators for decay heat removal to the atmosphere in this case. Steam driven aux feedwater pumps are usually the same pump design as the RCIC pump used in BWRs.
My experience is in the last 5 years. the industry has been changing rapidly over the last 20 years and still is changing. it's hard to look at the 1989-1994 NRC and compare it to the 2006-2011 NRC
yes. it would likely add some level of fatigue to the containment. but the containment heat transfer function is likely only designed for a small number of passive core cooling events and will need to be heavily reanalyzed after each one.
Similar to existing BWRs which will emergency depressurize to atmospheric pressure to allow low pressure coolant systems to inject. Every time you use ADS (automatic depressurization system) you need to do a TON in order to show that your vessel didnt get over fatigued, and it will likely limit the total lifetime before you need to reanneal the vessel
The problem is the NRC notices this stuff. I'm going to take a guess you've never been questioned by the NRC, but I have (nuclear engineer). They get on top of even the smallest hint of bullshit or mistake in logic or even poor quality packages. They would have already known that you are missing a safety system which they REQUIRED you to have and you LEGALLY COMMITED to have and you would have your project stopped and reviewed again which would caost MUCH MORE than 15% to get the project moving forward again. We are told to never ever challenge our NRC commitments or requirements, because the cost of messing up is a LOT more than what you 'could' gain by cutting something.
Sattilites use RTGs, not nuclear reactors. And RTG makes use of decay heat and the seebeck effect to generate a voltage difference. Very different from a nuclear reactor. As for nuclear power plants, the chain reaction is not "amplified", it is a chain reaction, nothing more or less. We actually control it using control rods and neutron absorbers. These plants can shutdown in less than 3 seconds, and only once has a plant failed to scram when called upon, and the backup scram system automatically did the job instead.
The pressurized coolant leak is the main design basis accident for nuclear power plants.
ALL nuclear power plants are designed to handle a double guillotine shear break in the largest coolant lines from the reactor, plus ANY single failure (including full failure of an entire train of safety systems), and prevent more than 1% core damage with radiation releases less than 10CFR100 requirements. A pressurized coolant means it is harder to inject, but it is not as bad of an issue as you would think. Pressurized coolant has advantages too. the two Fukushima units that had functioning HPCI and RCIC systems (passive steam powered cooling pumps) were able to keep cooled for 70 hours and 36 hours (for units 2 and 3 respectively) purely on steam with no DC power.
They werent designed with a maximum lifespan in mind. the "design lifespan" was based on financial decisions, but it was never the intent to make the plants only last 40 years.
additionally, they didnt have computers and methods to compute very specifically what 40 years would do to a plant or vessel, and we are finding that they went so far overboard in the conservatisms that we can easily get to 60 years and still maintain more than the required safety margins with no compromise in safety.
there are still operational challenges for some of the more advanced reactor types. The US has no generation 3 reactors producing power, we need to build Gen 3 and a lot of research on specific operational issues need to be completed on Gen 4 reactors, plus a big company like westinghouse or GE needs to pick up a gen 4 design and commercialize it.
it will happen eventually, may take another 20 years.
for normal power, when our generator is synced to the grid, we have 2 transformers which pull power in BEFORE it gets to the outside grid breaker. We don't pay the grid for this, since they never see it. One brings in 4160V power, the other brings in 6900V power.
We also have reserve transformers, and emergency reserve transformers, on top of diesel generators
The issue wasnt the DGs in the basement. they were put there because it was seismically more stable. The main issue was the building was designed as "flood tight" but not "flood proof". Basically it can handle a certain level of flooding, and that tsunamic went several times beyond that. Normal flooding would not have been an issue, but it isnt completely 100% water / flood proof. All the primary electrical switchgear are in basement locations as well.
Ive worked at 2 nuclear plants and visited 3 more, and at all of them in the US, their critical switchgear were in water tight areas above the flood plane (generally 30' or more above ground level)
Nuclear engineer here
The plant actually runs on generator power under normal conditions.
Nuclear plants have 4 AC power sources. The normal source is taking generator power BEFORE it goes out to the power grid in through the auxiliary transformers and then using internally for 4160 and 6900V power. Because this power hasn't gone to the grid yet, we don't "pay" for it. Additionally, when we are shut down, we can disconnect the generator and backfeed power in through the aux. transformers for power. This is typically an emergency/contingency action or an outage action to allow us to work on the reseve power system.
The standby source comes in from a different grid (or a different part of the same grid), and comes in from the reserve auxiliary transformers (sometimes called startup transformers). Because this is bringing power in from the grid, we "pay" for it (we get billed by the grid).
The emergency reserve transformer (sometimes called backup transformers) comes from a completey different grid than everything else. They power ONLY safety systems. Normal systems cannot use it.
The diesel generators are safety seismic and environmentally designed backup power systems. There is 1 DG for each primary safety division which has a decay heat removal function, and an additional DG for coolant injection. Most plants also have a fourth or fifth DG for DC power chargers only. There is enough fuel on site for a minimum of 1 week for all generators running 2% above maximum theoretical load of all equipment under worst case design conditions. The reality is you can probably get another 2-3 days past that since it assumes that like, air coolers and air heater are both on at the same time in the same area, and once you've stabilized an accident or emergency condition you can put most of the redundant safety systems into standby to conserve fuel.
that's not true.
after 3 days there the decay heat level has lowered to a point where you dont need the evaporative cooling effect to support passive decay heat removal.
the system is only ACCREDITED for 72 hours with no human interaction, however it can do more than 3 days. Additionally if you cant pump water into a non pressurized tank in 3 days you probably should be running a nuclear plant in the first place. ADDITIONALLY this doesnt include all of the non-safety critical systems such as aux feedwater which can run on steam alone and keep the core cooled for several days. (Fukushima Daiichi unit #2 was cooled for 70.2 hours on their steam pump alone)
It is a passive cooling system, once you activate it. And the activation is automated.
Once it is activated, the reactor primary coolant circuit cools to a storage tank. That water boils into steam, then touches the outer wall of the containment, which cools it. The steam condenses and drips down into a collector around the containment and gets funneled back into the water storage tank.
On the other side of the containment, the chimney effect will move enough air to keep the containment cooled. For the first 72 hours, water needs to be poured on to help remove heat through evaporation, after that you can keep cooled indefinately with no human interaction.
it is truely a passive system.
Nuclear engineer here.
Decay is not Fission.
Fission is splitting the atom.
Decay is the act of a radioactive atom to reduce itself closer to a stable groundstate.
Fission is controllable and is directly related to neutron population, and if we stop neutron production with control rods, fission stops.
Decay is not controllable, and happens all the time no matter what until the material reaches a stable ground state.
All light water plants, except the AP1000, need active cooling. (The GE ESBWR doesnt need active cooling either, but its design isnt approved or even completed yet). After shutdown the core is still boiling about 600 gpm of water at 1000 pounds pressure (in a BWR). this is due to the radioactive WASTE products decaying. The fuel isn't doing anythign after shutdown, but the waste products are trying to become stable again.
a main generator of a NPP (and any other large base load plant) cannot handle small or transient loads. when you lose offsite power, you have no load, and the generator has to shut down, otherwise you could damage the generator or destroy the turbine catastrophically.
equate it to riding a bike. if you are going 20 mph on a bike in first gear and you are pedaling as fast as you can it is very unstable. (there isnt enough load for the energy your legs are producing). if you are in 6th gear you'll be fine because there's enough load to counter what your legs are trying to put in.
anyways...this is why there are at least 2 qualified AC circuits plus at least 2 diesel systems per plant.
there is some kind of goofy DRM even if you buy it on steam.
before I can play any DRM, my EA account needs to be logged in through dragon age. it's kind of shobby.
Every once in a while i cant load my saved games, so i log out of the ea account and back in and it works.
it doesnt reduce radiation dose.
gamma requires several feet of shielding to bring it down.
the suits are just there to prevent particle contamination from getting in/on their bodies.
they are discussing that.
they just dont tell the press about it because the press will spin it as "industry thinks their plants will have 100 mile fallouts everywhere and the chickens will mutate and become the new dominant species"
the pool of water below the reactor (in this case in the torus) is designed to quench vented steam and reduce pressure in the reactor.
under normal conditions, when the reactor is pressurized, there are 2 emergency core cooling systems available to the core. The first is HPCI (high pressure coolant injection), and is an active pump meant to get water on the fuel. the second is RCIC (reactor core isolation cooling), and uses steam from the reactor to run a turbine and pump water into the core. RCIC requires only batteries....and that your suppression pool is below boiling point so it can quench the steam.
If the plant had electrical power, and they had a leak or a failure of the high pressure systems, they would vent all their steam to the suppression pool and run their low pressure core spray and low pressure coolant injection systems. the plant i'm at has 3 LPCI pumps and 1 LPCS pump. these low pressure pumps are designed for flow, not pressure, and are capable of fully filling the reactor vessel in all but the worst pipe breaks (double guillotine shear in the recirc lines). and even in the worst condition, provided you still have electrical power, the water that is lost funnels down through pipes back into the suppression pool where it can get run through heat exchangers, cooled, and pumped back into the core repeatably.
if you dont have power, the core has relief valves that automatically lift in safety mode without power if the core has too much pressure. this vents to the suppression pool. i dont know if the pool has relief valves or the containment, but you then would need to manually vent the containment to keep the pressure down enough to inject water in.
during the accident, i believe unit 2 had too much pressure build up in the core and that stopped seawater injection for a while. they had to release the pressure to allow the pumper trucks to get water back in.
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nuclear engineer here. going through this document, there are a lot of things which are just flat out incorrect or misleading. I don't disagree that some parts of the AP1000 plant may need some more looking, but the article is not descriptive enough over what specific conditions there are issues with. Having read the entire design control document series for the AP1000 and knowing what type of organization fairwinds is (run by Arnie Gunderson, not the most reputable source for factual nuclear information), I'm a little skeptical.
First, the "radiation" didnt make the burning. If you were getting enough radiation to feel a 'burn' you would be losing your hair right now. To put the numbers in perspective, you received a maximum of 1-1.5 Rem (10-15 mSv) of radiation. The average yearly background exposure not including medical is ~320 mRem/year (3.2mSv). Including medical: ~620 mRem (6.2 mSv). For a nuclear power plant worker the maximum allowed per year is 5 Rem (50 mSv). On average nuclear power plant workers get an additional 100mRem/year (1mSv), except for outage workers who average about an extra 300 mRem/year (3mSv). People working at Fukushima after the accident were authorized to receive 25 Rem (250mSv) to protect the plant and the public. There is no statistical increase in liklihood of cancer until you pass 10 Rem (100mSv). Even using Linear No Threshold which is the most conservative accepted estimate and is used by the government for calculating deaths from radiation exposure to assign value to radiation, your increase in risk is .1% over your entire lifetime due to that procedure.
Again just FYIs and facts.
And none of those had to do with the age of the plant. Additionally, probabilistic risk assessment (which is what generates the number 1 accident in X years) is a model, and a tool, used to determine what is the most safety significant components which you need to maintain. A tool, nothing more. People cite numbers believing that they are gospel truth that a plant will only have 1 accident every X years, but it is only a tool. Additionally, PRA depends on your design basis being correct from the start. PRA does not model operator error very well as what happened in TMI (multiple operator errors combined with process errors, multiple latent failures, etc), and does not model bad designs like chernobyl. PRA also can't account for an incorrect design basis, like only 14ft tsunami vs. 45ft. tl;dr, age of the equipment and strucutres has had nothing to do with any nuclear accident. Only human error at some level of the process (design, operation, etc).
In BWRs, the HPCI system (in BWR 2- some of the 5s) and RCIC system (found in BWR4+) are steam driven. Fukushima Daiichi unit 2 was cooled for 70 hours on RCIC using steam alone, and unit 3 on RCIC and HPCI for 36 hours. The limiting issue is cooling water for the pump, because the cooling water is the same water that is going in and out of the reactor and suppression pool, eventually the whole mass heats up beyond the maximum temperature of the pump. PWRs have aux feedwater which is steam driven or direct diesel driven (or both depending on design) to provide feedwater to the steam generators for decay heat removal to the atmosphere in this case. Steam driven aux feedwater pumps are usually the same pump design as the RCIC pump used in BWRs.
My experience is in the last 5 years. the industry has been changing rapidly over the last 20 years and still is changing. it's hard to look at the 1989-1994 NRC and compare it to the 2006-2011 NRC
yes. it would likely add some level of fatigue to the containment. but the containment heat transfer function is likely only designed for a small number of passive core cooling events and will need to be heavily reanalyzed after each one. Similar to existing BWRs which will emergency depressurize to atmospheric pressure to allow low pressure coolant systems to inject. Every time you use ADS (automatic depressurization system) you need to do a TON in order to show that your vessel didnt get over fatigued, and it will likely limit the total lifetime before you need to reanneal the vessel
The problem is the NRC notices this stuff. I'm going to take a guess you've never been questioned by the NRC, but I have (nuclear engineer). They get on top of even the smallest hint of bullshit or mistake in logic or even poor quality packages. They would have already known that you are missing a safety system which they REQUIRED you to have and you LEGALLY COMMITED to have and you would have your project stopped and reviewed again which would caost MUCH MORE than 15% to get the project moving forward again. We are told to never ever challenge our NRC commitments or requirements, because the cost of messing up is a LOT more than what you 'could' gain by cutting something.
Sattilites use RTGs, not nuclear reactors. And RTG makes use of decay heat and the seebeck effect to generate a voltage difference. Very different from a nuclear reactor. As for nuclear power plants, the chain reaction is not "amplified", it is a chain reaction, nothing more or less. We actually control it using control rods and neutron absorbers. These plants can shutdown in less than 3 seconds, and only once has a plant failed to scram when called upon, and the backup scram system automatically did the job instead.
The pressurized coolant leak is the main design basis accident for nuclear power plants. ALL nuclear power plants are designed to handle a double guillotine shear break in the largest coolant lines from the reactor, plus ANY single failure (including full failure of an entire train of safety systems), and prevent more than 1% core damage with radiation releases less than 10CFR100 requirements. A pressurized coolant means it is harder to inject, but it is not as bad of an issue as you would think. Pressurized coolant has advantages too. the two Fukushima units that had functioning HPCI and RCIC systems (passive steam powered cooling pumps) were able to keep cooled for 70 hours and 36 hours (for units 2 and 3 respectively) purely on steam with no DC power.
They werent designed with a maximum lifespan in mind. the "design lifespan" was based on financial decisions, but it was never the intent to make the plants only last 40 years. additionally, they didnt have computers and methods to compute very specifically what 40 years would do to a plant or vessel, and we are finding that they went so far overboard in the conservatisms that we can easily get to 60 years and still maintain more than the required safety margins with no compromise in safety.
there are still operational challenges for some of the more advanced reactor types. The US has no generation 3 reactors producing power, we need to build Gen 3 and a lot of research on specific operational issues need to be completed on Gen 4 reactors, plus a big company like westinghouse or GE needs to pick up a gen 4 design and commercialize it. it will happen eventually, may take another 20 years.
for normal power, when our generator is synced to the grid, we have 2 transformers which pull power in BEFORE it gets to the outside grid breaker. We don't pay the grid for this, since they never see it. One brings in 4160V power, the other brings in 6900V power. We also have reserve transformers, and emergency reserve transformers, on top of diesel generators
The issue wasnt the DGs in the basement. they were put there because it was seismically more stable. The main issue was the building was designed as "flood tight" but not "flood proof". Basically it can handle a certain level of flooding, and that tsunamic went several times beyond that. Normal flooding would not have been an issue, but it isnt completely 100% water / flood proof. All the primary electrical switchgear are in basement locations as well. Ive worked at 2 nuclear plants and visited 3 more, and at all of them in the US, their critical switchgear were in water tight areas above the flood plane (generally 30' or more above ground level)
Nuclear engineer here The plant actually runs on generator power under normal conditions. Nuclear plants have 4 AC power sources. The normal source is taking generator power BEFORE it goes out to the power grid in through the auxiliary transformers and then using internally for 4160 and 6900V power. Because this power hasn't gone to the grid yet, we don't "pay" for it. Additionally, when we are shut down, we can disconnect the generator and backfeed power in through the aux. transformers for power. This is typically an emergency/contingency action or an outage action to allow us to work on the reseve power system. The standby source comes in from a different grid (or a different part of the same grid), and comes in from the reserve auxiliary transformers (sometimes called startup transformers). Because this is bringing power in from the grid, we "pay" for it (we get billed by the grid). The emergency reserve transformer (sometimes called backup transformers) comes from a completey different grid than everything else. They power ONLY safety systems. Normal systems cannot use it. The diesel generators are safety seismic and environmentally designed backup power systems. There is 1 DG for each primary safety division which has a decay heat removal function, and an additional DG for coolant injection. Most plants also have a fourth or fifth DG for DC power chargers only. There is enough fuel on site for a minimum of 1 week for all generators running 2% above maximum theoretical load of all equipment under worst case design conditions. The reality is you can probably get another 2-3 days past that since it assumes that like, air coolers and air heater are both on at the same time in the same area, and once you've stabilized an accident or emergency condition you can put most of the redundant safety systems into standby to conserve fuel.
that's not true. after 3 days there the decay heat level has lowered to a point where you dont need the evaporative cooling effect to support passive decay heat removal. the system is only ACCREDITED for 72 hours with no human interaction, however it can do more than 3 days. Additionally if you cant pump water into a non pressurized tank in 3 days you probably should be running a nuclear plant in the first place. ADDITIONALLY this doesnt include all of the non-safety critical systems such as aux feedwater which can run on steam alone and keep the core cooled for several days. (Fukushima Daiichi unit #2 was cooled for 70.2 hours on their steam pump alone)
It is a passive cooling system, once you activate it. And the activation is automated. Once it is activated, the reactor primary coolant circuit cools to a storage tank. That water boils into steam, then touches the outer wall of the containment, which cools it. The steam condenses and drips down into a collector around the containment and gets funneled back into the water storage tank. On the other side of the containment, the chimney effect will move enough air to keep the containment cooled. For the first 72 hours, water needs to be poured on to help remove heat through evaporation, after that you can keep cooled indefinately with no human interaction. it is truely a passive system.
Nuclear engineer here. Decay is not Fission. Fission is splitting the atom. Decay is the act of a radioactive atom to reduce itself closer to a stable groundstate. Fission is controllable and is directly related to neutron population, and if we stop neutron production with control rods, fission stops. Decay is not controllable, and happens all the time no matter what until the material reaches a stable ground state. All light water plants, except the AP1000, need active cooling. (The GE ESBWR doesnt need active cooling either, but its design isnt approved or even completed yet). After shutdown the core is still boiling about 600 gpm of water at 1000 pounds pressure (in a BWR). this is due to the radioactive WASTE products decaying. The fuel isn't doing anythign after shutdown, but the waste products are trying to become stable again.
a main generator of a NPP (and any other large base load plant) cannot handle small or transient loads. when you lose offsite power, you have no load, and the generator has to shut down, otherwise you could damage the generator or destroy the turbine catastrophically. equate it to riding a bike. if you are going 20 mph on a bike in first gear and you are pedaling as fast as you can it is very unstable. (there isnt enough load for the energy your legs are producing). if you are in 6th gear you'll be fine because there's enough load to counter what your legs are trying to put in. anyways...this is why there are at least 2 qualified AC circuits plus at least 2 diesel systems per plant.
all floors in nuclear power plants are posted by elevation. there isnt a 'first floor' 'second floor' 'basement'.
there is some kind of goofy DRM even if you buy it on steam. before I can play any DRM, my EA account needs to be logged in through dragon age. it's kind of shobby. Every once in a while i cant load my saved games, so i log out of the ea account and back in and it works.
it doesnt reduce radiation dose. gamma requires several feet of shielding to bring it down. the suits are just there to prevent particle contamination from getting in/on their bodies.
and the tectonic plates. its just as much their fault too ^_^
in a BWR loss of water actually slows/stops the reaction. water is used to moderate neutrons and without it you cannot maintain the nuclear reaction.
they are discussing that. they just dont tell the press about it because the press will spin it as "industry thinks their plants will have 100 mile fallouts everywhere and the chickens will mutate and become the new dominant species"
the pool of water below the reactor (in this case in the torus) is designed to quench vented steam and reduce pressure in the reactor. under normal conditions, when the reactor is pressurized, there are 2 emergency core cooling systems available to the core. The first is HPCI (high pressure coolant injection), and is an active pump meant to get water on the fuel. the second is RCIC (reactor core isolation cooling), and uses steam from the reactor to run a turbine and pump water into the core. RCIC requires only batteries....and that your suppression pool is below boiling point so it can quench the steam. If the plant had electrical power, and they had a leak or a failure of the high pressure systems, they would vent all their steam to the suppression pool and run their low pressure core spray and low pressure coolant injection systems. the plant i'm at has 3 LPCI pumps and 1 LPCS pump. these low pressure pumps are designed for flow, not pressure, and are capable of fully filling the reactor vessel in all but the worst pipe breaks (double guillotine shear in the recirc lines). and even in the worst condition, provided you still have electrical power, the water that is lost funnels down through pipes back into the suppression pool where it can get run through heat exchangers, cooled, and pumped back into the core repeatably. if you dont have power, the core has relief valves that automatically lift in safety mode without power if the core has too much pressure. this vents to the suppression pool. i dont know if the pool has relief valves or the containment, but you then would need to manually vent the containment to keep the pressure down enough to inject water in. during the accident, i believe unit 2 had too much pressure build up in the core and that stopped seawater injection for a while. they had to release the pressure to allow the pumper trucks to get water back in.