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March 20, 2011 • 0 Comments
Dry-Cask Storage vs. Spent-Fuel Pools| by Lisbeth Gronlund | nuclear power | nuclear power safety | Japan nuclear |
The nuclear crisis in Japan has started discussions about the safety and security advantages of storing spent fuel in dry casks (see photo) rather than spent fuel pools. UCS has long recommended that spent fuel be transferred from the pool to dry cask storage once the fuel has cooled enough, after about five years. This is a major issue in the U.S. because U.S. pools are becoming increasingly packed with spent fuel.
Here are some links for more information on this issue:
(1) Chapter 5, “Ensuring the Safe Disposal of Nuclear Waste,” from UCS’s report Nuclear Power in a Warming World (2007), which covers interim and long-term waste storage, and discusses why reprocessing is neither an effective nor desirable waste management strategy. (Note that the discussion of Yucca Mountain is out of date.)
(2) A 2003 paper Ed Lyman co-authored, followed by links to comments on the paper by the NRC, and the authors’ response:
“Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States,” by Alvarez, R., Beyea, J., Janberg, K., Kang, J., Lyman, E., Macfarlane, A., Thompson, G., and von Hippel, F.N., Science and Global Security, Vol 11, 1:1-51 (2003)
Here’s the abstract of the paper:
Because of the unavailability of off-site storage for spent power-reactor fuel, the NRC has allowed high-density storage of spent fuel in pools originally designed to hold much smaller inventories. As a result, virtually all U.S. spent-fuel pools have been re-racked to hold spent-fuel assemblies at densities that approach those in reactor cores. In order to prevent the spent fuel from going critical, the fuel assemblies are partitioned off from each other in metal boxes whose walls contain neutron-absorbing boron.
It has been known for more than two decades that, in case of a loss of water in the pool, convective air cooling would be relatively ineffective in such a “dense-packed” pool. Spent fuel recently discharged from a reactor could heat up relatively rapidly to temperatures at which the zircaloy fuel cladding could catch fire and the fuel’s volatile fission products, including 30-year half-life 137Cs, would be released. The fire could well spread to older spent fuel. The long-term land-contamination consequences of such an event could be significantly worse than those from Chernobyl.
No such event has occurred thus far. However, the consequences would affect such a large area that alternatives to dense-pack storage must be examined—especially in the context of concerns that terrorists might find nuclear facilities attractive targets. To reduce both the consequences and probability of a spent-fuel-pool fire, it is proposed that all spent fuel be transferred from wet to dry storage within five years of discharge. The cost of on-site dry-cask storage for an additional 35,000 tons of older spent fuel is estimated at $3.5–7 billion dollars or 0.03–0.06 cents per kilowatt-hour generated from that fuel. Later cost savings could offset some of this cost when the fuel is shipped off site.
The transfer to dry storage could be accomplished within a decade. The removal of the older fuel would reduce the average inventory of 137Cs in the pools by about a factor of four, bringing it down to about twice that in a reactor core. It would also make possible a return to open-rack storage for the remaining more recently discharged fuel. If accompanied by the installation of large emergency doors or blowers to provide large-scale airflow through the buildings housing the pools, natural convection air cooling of this spent fuel should be possible if airflow has not been blocked by collapse of the building or other cause. Other possible risk-reduction measures are also discussed.
“Review of ‘Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States’,” by the Nuclear Regulatory Commission (NRC), Science and Global Security, Vol. 11, 2-3:203-212 (2003)
“Response by the Authors to the NRC Review of ‘Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States’,” by Alvarez, R., Beyea, J., Janberg, K., Kang, J., Lyman, E., Macfarlane, A., Thompson, G., and von Hippel, F.N., Science and Global Security, Vol. 11, 2-3:213-223 (2003)
(3) Ed also co-authored a paper addressing issues related to the potential effects of a radiation release from spent fuel:
“Damages from a Major Release of 137Cs into the Atmosphere of the United States,” by Beyea, J., Lyman, E., von Hippel, F. N., Science and Global Security, Vol. 12:125–136 (2004)
Here’s the abstract:
We report estimates of costs of evacuation, decontamination, property loss, and cancer deaths due to releases by a spent fuel fire of 3.5 and 35 MCi of 137Cs into the atmosphere at five U.S. nuclear-power plant sites. The MACCS2 atmospheric-dispersion model is used with median dispersion conditions and azimuthally-averaged radial population densities. Decontamination cost estimates are based primarily on the results of a Sandia study. Our five-site average consequences are $100 billion and 2000 cancer deaths for the 3.5 MCi release, and $400 billion in damages and 6000 cancer deaths for the 35 MCi release. The implications for the cost-benefit analyses in “Reducing the hazards” are discussed.
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March 19, 2011 • 2 notes • 1 Comment
Good Sources on Fukushima| by David Wright | nuclear power | nuclear power safety | Japan nuclear |
These sites are providing very useful, timely information on the situation in Japan:
-The Bulletin of the Atomic Scientists is posting periodic updates from Tatsujiro Suzuki, a Japanese expert on nuclear power, who is following events in Tokyo.
-Jeffrey Lewis at armscontrolwonk is posting daily updates from the Washington DC Office of the Federation of Electric Power Companies of Japan (FEPC) .
-The Japan Atomic Industrial Forum (JAIF) is posting updated status charts on the Dai-Ichi and Daini nuclear plants.
Translating times between Japan and the US:
JST = Japan Standard Time = GMT + 9
EDT = Eastern Daylight Time = GMT - 4
so H:00 JST = (H:00 – 13:00) EDT
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March 19, 2011 • 5 notes • 3 Comments
Possible Source of Leaks at Spent Fuel Pools at Fukushima| by Dave Lochbaum | nuclear power | nuclear power safety | Japan nuclear |
A current focus of concern in Japan now is the pools at the reactors where spent fuel is stored. Some of this spent fuel is still very radioactive since it was only removed from the reactors a few months ago, and it must be covered by water and cooled to keep from overheating. If the spent fuel rods get too hot, they can suffer damage and release significant amounts of radioactive gases into the atmosphere, and could eventually catch fire.
Since several of the reactor buildings that surround these pools have been damaged by explosions, the radioactivity released from the pools in those buildings would get directly into the atmosphere. Similar fuel damage within the reactor cores would be surrounded by the reactor’s primary containment so that a much smaller fraction would get out, unless there was a significant breech of the containment.
Water needs to be added to the spent fuel pools at Fukushima since heating by the spent fuel causes the water to evaporate and boil off.
In addition, reports from Japan say that the spent fuel pool at reactor Unit 4 is leaking, which further increases the need for additional water.
A possible source of the leak in the Unit 4 pool may be the seals around the doors (or “gates”) on one side of the spent fuel pool. These gates are shown in the diagram below. They are located between the pool and the area above the reactor vessel. They are concrete with metal liners, and are roughly 20’x 3’.
When fuel is moved between the pool and vessel, this whole region is filled with water, the gates are opened, and the fuel can be moved to or from the reactor core while remaining under water. The water not only keeps the fuel rods cool but acts as a radiation shield.
Boiling Water Reactor (BWR) Spent Fuel Cooling System
When the gates are closed, they are made watertight by an inflatable seal, similar to a bicycle innertube, that runs around the sides and bottom of the gates. Electric air pumps are used to inflate these seals and keep them inflated as air leaks out of them over time.
These pumps are powered by electricity from the power grid, and not by backup diesel power or batteries. So once the power grid in Japan was knocked out, these seals could not be inflated if they lost air over time. If these seals lost air they could lead to significant water loss from the pool, even if there were no direct physical damage to the pool from the earthquake or tsunami. This may be what happened at pool 4, and could affect the other pools as well.
We saw an example of this in the US at the Hatch nuclear plant in Georgia in December 1986. This reactor is very similar to the reactors at Fukushima. In the Hatch case, the line supplying air to the inflatable seal was accidentally closed, the seal lost pressure and created a leak, and by the time the problem was identified several hours later some 141,000 gallons of water leaked from the pool—about half the water in the pool Fortunately, the source of the problem was discovered and fixed before the water level uncovered the fuel.
An NRC document on the leak gave this description of the event:
A valve in the single air supply line to the seals was mistakenly closed. Although water level dropped about 5 feet and low-level alarms in the spent fuel pool worked, the leak was not specifically identified for several hours because a leak detection device was valved out and none of the seals were instrumented to alarm on loss of air pressure.
The NRC document goes on to note that if the water level had gotten low enough to expose the fuel the high radiation level around the pool would have made it difficult for workers to fix the problem.
The closed air line in the Hatch case had the same result that lack of electric power the air pump inflating the seals in Japan could have.
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The spent fuel pool appears on the right side of this diagram. The reactor vessel and its reactor core appear on the left side. The refueling platform is used to move fuel assemblies one at a time between the reactor core and the spent fuel pool through an opening in the spent fuel pool wall created by removal of a gate.
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Overhead view of an irradiated spent fuel bundle being transferred from the reactor core (lower right) to the spent fuel pool (upper left) through what is called the “cattle chute” at the Browns Ferry Nuclear Plant in Alabama. The spent fuel pool gate has been removed to connect the spent fuel pool water with the water in the reactor well area above the open reactor pressure vessel.
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Looking down at the fuel transfer canal at the Grand Gulf Nuclear Station in Mississippi. Grand Gulf is a BWR with a Mark III containment. It features a fuel transfer canal that does not exist in the BWR Mark I and Mark II containment designs. However, all three Containment designs feature gates that are removable from the spent fuel pool walls to allow underwater transfer of spent fuel assemblies. In this picture, the fuel transfer canal is in the center, the spent fuel pool is to the left, and the cask loading area is to the upper right and is used when fuel is transferred from spent fuel pool to casks.
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Looking down into the spent fuel pool at the Grand Gulf Nuclear Station in Mississippi before the plant commenced operation. The spent fuel pool is drained of water. The fuel storage racks can be seen in the lower region of the spent fuel pool. The beams used to hold the racks in place against forces from an earthquake can been seen between the racks and the pool walls. In the lower right portion of the picture, the opening in the fuel pool wall created by the removal of the spent fuel gate can be seen.
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A cross-section fuel of a typical BWR spent fuel pool. The fuel pool gate appears on the left side of the pool. The bottom of the opening created when the gate is removed (or its seals leaking) is about 5 feet above the top of spent fuel in the storage racks at the bottom of the spent fuel pool.
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March 18, 2011 • 2 notes • 1 Comment
Transcripts of Press Briefings on Fukushima by UCS Technical Experts| by David Wright | nuclear power | nuclear power safety | Japan nuclear |
We have been providing daily phone briefings for reporters by our technical experts on the evolving situation surrounding the crippled reactors in Japan, and will continue these briefings through the weekend.
To make this information available to the public, we have been transcribing and posting these briefings, including the Q&A sessions. The transcripts, and voice recordings of the opening remarks remarks by the briefers, are available here, and we will add new transcripts daily to that same site.
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March 18, 2011 • 1 note • 3 Comments
More on KI Pills| by Lisbeth Gronlund | nuclear power | nuclear power safety | Japan nuclear |
We’ve gotten some questions asking for clarification about our statement on potassium-iodide (KI) pills. In particular, why are KI pills effective in the case of inhalation of radioactive iodine, but not considered an effective countermeasure to ingesting it via, for example, milk?
According to the 2004 National Academy of Sciences study on Distribution and Administration of Potassium Iodide in the Event of a Nuclear Incident:
Exposure to radioactive iodine is possible through the ingestion pathway, so it is important that plans address this situation. Monitoring of the environment and food products controls this route of exposure. Removing contaminated products from the market and isolating contaminated products until the radioactive iodine decays to safe levels are the most effective way to eliminate radiation exposure and damage to the thyroid. That also eliminates the need for the use of KI by the general public as a protective action.
Potassium iodide can only reduce the risk from radioactive iodine that has entered the body, not eliminate it. People in the radioactive plume do not have the option of not breathing, so taking KI is an effective countermeasure against inhalation. However, people have the option of not drinking contaminated milk or eating other contaminated food products. In comparison, taking KI would be less effective.
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March 18, 2011 • 20 notes • 17 Comments
Possible Cause of Reactor Building Explosions| by Dave Lochbaum | nuclear power | nuclear power safety | Japan nuclear |
Dramatic videos show the explosions that severely damaged the reactor buildings at first Unit 1 and then Unit 3 at the stricken Fukushima Dai-Ichi nuclear plant in Japan. The explosions are attibuted to the ignition of hydrogen gas that collected within the reactor buildings. This was early in the crisis, and before the spent fuel pools are thought to have lost water and started producing hydrogen.
The hydrogen was likely produced by damaged fuel rods in the reactor core. To reduce pressure in the reactor vessel, some of that hydrogen was released from the vessel into the primary containment structure of the reactor.
A key, unsolved riddle is how a significant amount of hydrogen escaped from the primary containment into the reactor building, and how this low-probability event would have happened in mulitple reactors.
How Hydrogen Got into Primary Containment
Figure 1 shows a cross-sectional view of a boiling water reactor with a Mark I containment like that at Fukushima Dai-Ichi. The reactor core is housed within a metal reactor vessel. The reactor vessel is enclosed within the primary containment structure. The reactor building completely surrounds the containment structure. The reactor building walls are made of 18 to 30 inch-thick concrete up to the elevation of the refueling platform. The walls are made of metal from that elevation to the roof.
Figure 1
The hydrogen gas most likely came from a chemical reaction between water and the metal cladding of fuel rods in the reactor cores when the water level inside the reactor vessels dropped low enough to expose at least the upper core regions.The hydrogen gas initially collected in the reactor vessel.
To cool the fuel in the reactor, workers attempted to pump seawater into the reactor vessel. As pressure inside the reactor vessel increased, it kept water from flowing into the reactor. Periodically, workers opened valves to vent steam and gas from the reactor vessel to into the pressure suppression chamber (also called the torus). The gas, including hydrogen, collected in the torus and periodically equalized with the air space in the drywell.
When pressure in the primary containment (the combination of the drywell and the torus) rose too high, workers vented the containment to the atmosphere. This vent piping passed through the reactor building, but discharged well outside of it, and should not have led to a hydrogen buildup inside the building.
How Hydrogen May Have Gotten from Primary Containment into the Reactor Building
The destruction of the Unit 1 and 3 reactor buildings appears to have been caused by hydrogen explosions. As noted above, an unanswered question is how the hydrogen got into the reactor buildings. A little-known test performed decades ago at the Brunswick nuclear plant in North Carolina may hold the key to answering that question.
To satisfy a requirement in the American Society of Mechanical Engineers (ASME) code for prototype containment designs, workers performed a structual integirty test on the reactor at Brunswick in the 1970s.
The primary containment structure at Brunswick was designed to withstand an internal pressure of 62 pounds per square inch (psi). The ASME code required it to be tested at 71 psi. This test involved pumping air into the containment structure until the pressure rose to 71 psi. The pumps would then be turned off and the pressure would be monitored for several hours to verify that it remained fairly constant, indicating that the primary containment was intact and not leaking. During this time, workers would record data from strain gauges and other instrumentation to verify that structural loads were properly distributed.
But as workers increased the containment pressure they encountered a problem. The pressure stopped increasing and remained constant at 70 psi. The pumps continued to push air into the containment, but its pressure just stopped increasing. This unexpected plateau started a hunt for air leaking from the containment somewhere.
A hissing sound attracted workers to the top of the containment structure. They identified air leaking through the drywell flange area (see Figure 1). The metal drywell head (see Figure 2) is bolted to the metal drywell with a rubber O-ring between the surfaces to provide a good seal fit.
Figure 2
Workers found that the containment pressure of 70 psi pushing upward against the inner dome of the drywell head lifted it off the drywell flange enough to provide a pathway for air to leak from the containment. That air leaked into the area labeled refueling cavity in Figure 1. The refueling cavity is located outside the primary containment but inside the reactor building.
At Brunswick, workers tightened the drywell head bolts beyond the amount specified in the reactor plans in order to reduce the leak rate and continue the test. While workers conducted pressure tests at all nuclear reactors prior to initial startup and periodically thereafter, these tests were performed at or below the containment design-pressure of 62 psi. So none of them reached the pressure that caused the leak around the drywell head.
In other words, had Brunswick not featured a prototype containment design, its initial and recurring pressure tests would have been conducted at 62 psi, not 71 psi. Leaking from the drywell head was not observed until the containment pressure rose to 70 psi.
How does this Brunswick containment testing experience relate to the reactor building explosions experienced at Fukushima Dai-Ichi Units 1 and 3?
Like Brunswick, the containment design at those reactors features a drywell head bolted onto the lower portion of the drywell. Workers at these reactors faced siginficant problems cooling the reactor cores. The combined effects of the earthquake and tsunami left the reactors without ac electrical power. The only dc-powered (i.e., battery-powered) backup system was lost when the batteries were exhausted. Workers turned to their only remaining option: injecting sea water into the reactor vessels to cool the reactor cores.
The pumps used to pump seawater into the vessel operated at low pressure. When seawater entered the reactor vessel, it was heated by the hot reactor core to the point of boiling. Steam produced by the boiling increased the pressure inside the reactor vessel. To prevent this rising pressure from hindering seawater from being pumped into reactor, workers periodically vented the reactor vessel. This carried steam and gas, including hydrogen, into the primary containment. This flow in turn increased the pressure inside containment. When containment pressure rose too high, workers vented the containment to the atmosphere.
The workers properly sought to minimize the amount of gas they vented from containment to the atmosphere to lessen the amount of radiation released. They did this by allowing the containment pressure to rise as high as tolerable between ventings.
It is possible that the containment pressures rose high enough to replicate the Brunswick experience by lifting the drywell head enough to allow hydrogen and other gases to leak into the refueling cavity and reactor building. If so, hydrogen could build up to an explosive mixture.
This tragedy will be closely examined for its causes. That scrutiny must determine how hydrogen got into the reactor building early in the crisis. The drywell head pathway may be that answer.
Answering this question is critical to prevent hydrogen explosions at the other reactors at Fukushima.
If this mechanism is the cause of the leak, it could be averted easily and effectively simply by changing the venting procedures so that workers vent the containment pressure to the atmosphere more frequently and do not let it build up to such high level. Taking such action might moderately increase the amount of radioactive gases vented into the atmosphere, but could eliminate a source of hydrogen inside the reactor buildings that could cause another explosion.
Authorities should launch an investigation to pinpoint the source of the hydrogen leak to eliminate this risk in the future. But in the meantime, since the Brunswick test showed that this containment is vulnerable to high-pressure leaking, Tokyo Electric Power Co. can and should take immediate steps to avoid creating such a leak by changing its procedures to vent the containment before it builds up to such high pressure (70 psi).
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March 17, 2011 • 1 note • 38 Comments
Fuel Amounts at Fukushima| by David Wright | nuclear power | nuclear power safety | Japan nuclear |
This post was revised 11:45pm Thursday.
Based on Japanese press stories, we have compiled a table of the amount of fuel in the cores of the reactors and the spent-fuel pools in the 6 reactors at the Fukushima Dai-Ichi nuclear facility.
While BWR fuel comes in various sizes, the last column assumes 170 kg per assembly. Each fuel assembly consists of roughly 60 fuel rods.
Thanks to readers for confirming that the fuel rods in Unit 4 had been moved from the core to the spent fuel pool during maintenance.
Units 5 and 6 were reported to be producing power in January, but are now shut down for mainenance. Reports say that 130 assemblies from each core were recently transfered to the pools, but those were included in the previous numbers in the table for the spent fuel for those reactors. If anyone has additional information about these reactors, please let us know.
A New York Times article states that 32 assemblies in the spent fuel pool of Unit 3 are MOX. The MOX fuel rods were stored in the pool but TEPCO announced they were being loaded into the core last fall, so we think those are currently in the core.
The same article says that a total of 11,125 spent fuel assemblies are stored at the Fukushima Dai-Ichi facility. However, not all of those are stored in the pools in the reactor buildings. Several hundred are currently in dry cask storage, and more than half of the total are stored in a common storage pool.
Thanks to Masa Takubo for his help in compiling these numbers.
Sources:
http://www3.nhk.or.jp/news/genpatsu-fukushima/
http://astand.asahi.com/magazine/judiciary/articles/2011031600001.html
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March 17, 2011 • 21 notes • 20 Comments
Radiation Risk to the US| by Ed Lyman | nuclear power | nuclear power safety | Japan nuclear |
Given the fact that Japan is thousands of miles from the United States, it is highly unlikely that Americans would be exposed to radioactive material from direct inhalation of a plume from the Fukushima nuclear complex.
While wind patterns will likely carry the radioactive plume eastward, radioactive material will be so diffuse by the time it reaches Hawaii, Alaska, or the mainland United States that it is highly unlikely to create significant health concerns.
Related to this, UCS just released a statement about potassium-iodide pills:
The people of Japan should be given priority access to potassium iodide (KI) pills used to protect against thyroid cancer following inhalation of radioactive iodine.
Given the fact that Japan is thousands of miles from the United States, it is highly unlikely that Americans would be exposed to radioactive iodine from direct inhalation of a plume from the Fukushima nuclear complex. Direct inhalation is the kind of exposure that potassium iodide pills would be most effective against.
Regardless, there are reports that global supplies of potassium iodide pills are being depleted because Americans are buying them, prompting fears that there will not be adequate supplies in Japan in the event of a larger radiological release.
Besides inhalation, another way Americans could be exposed to radioactive iodine is if agricultural products were contaminated. Radioactive iodine could be ingested by dairy cows, for example, and then would be concentrated in milk. Potassium iodide, however, would not be an effective countermeasure in that situation. Moreover, federal and state health authorities would test for such contamination and could take products off the market if necessary.
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March 17, 2011 • 2 notes • 15 Comments
Nuclear “Station Blackout”| by Dave Lochbaum | nuclear power | nuclear power safety | Japan nuclear |
The combination of an earthquake followed by a tsunami in Japan initiated a sequence of events that ultimately led to damage to the reactor cores at Fukushima Dai-Ichi Units 1, 2, and 3 caused by inadequate cooling.
Can’t happen here? Perhaps not by the same method, but definitely with the same consequences.
The earthquake caused the normal supply of electrical power—that from the electrical grid—for the Fukushima nuclear plant to be lost. Per design, the emergency diesel generators at the site automatically started and provided power to essential emergency equipment.
Then the tsunami arrived and disabled the emergency diesel generators. This left the plant without alternating current (ac) electrical power. This condition with no ac electrical power is called a station blackout (SBO).
Per design, batteries provided direct current (dc) electrical power for a bare-bones minimal subset of emergency equipment. The dc power enabled a steam-driven turbine connected to a pump of the reactor-core isolation cooling (RCIC) system to supply cooling water for the reactor cores. The steam was being produced by the decay heat from the shut down reactor cores.
In June 1988, the NRC adopted a new safety regulation (10 CFR 50.63, “Loss of all alternating current power”) that required the owners of US reactors to take steps to assure their facilities could safely withstand an SBO event lasting 4 or 8 hours, depending on site-specific parameters.
Long after the plant owners implemented all the modifications to their reactors and revisions to operating procedures required to comply with the SBO regulation, the NRC evaluated the effectiveness of the new requirements and published the results in a report called NUREG-1776. Table B-2 of this NRC report summarizes results for each reactor operating in the US (see below for help in reading these tables).
Eleven reactors have batteries designed to supply dc power for up to 8 hours should an SBO occur. The Fukushima reactors were also equipped with 8-hour battery capacities; they were insufficient to meet the challenge. Ninety-three US reactors are designed with batteries lasting half that long.
What are the odds of a SBO leading to disaster at a US reactor?
Higher than you might think.
For example, the NRC’s report shows that the risk of an SBO at the Brunswick nuclear plant in North Carolina leading to reactor core damage is nearly twice the risk from all other causes combined. Brunswick’s batteries are sized to last only 4 hours—half the capacity of the batteries that failed to preserve safety at Fukushima.
And, as Figure 1 below shows, Brunswick is not so unusual. This plot shows that for many US reactors, SBO accounts for a large fraction of risk that events would lead to core damage.
Figure 1: This plot shows the core-damage frequency (CDF) due to SBO as a fraction of the core-damage frequency due to all causes (“Plant CDF”). The bars show how many of the US reactors had a value of this ratio falling in the range 0-5%, 5-10%, etc.
Rather than stand behind the empty reassurance that our reactors are not susceptible to SBO caused by an earthquake/tsunami one-two punch, the US government should review the SBO risk from two perspectives:
1) How to increase the reliability of the electrical grid being restored and emergency diesel generators being repaired so as to improve the odds of exiting a SBO condition before the batteries are depleted.
2) How to lessen the chances of a reactor meltdown should the batteries be depleted before ac electrical power is restored.
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Reading the NUREG-1766 Tables:
For those interested in the details of the NUREG-1766 tables, here is a brief description of the entries and terminology in the table:
CDF: Core Damage Frequency
Plant CDF: Risk of reactor core damage per year from all credible scenarios. In other words, the risk of reactor core damage from a pipe break, from a station blackout, from the unexpected shut down of the main turbine, and all other credible causes are added up to yield the total risk per year.
SBO CDF: Risk of reactor core damage per year from only a station blackout event.
Percent SBO CDF of Plant CDF: The percentage of overall risk of reactor core damage from a station blackout event.
Coping time: The number of hours the batteries are designed to provide dc power / the reliability of the onsite emergency diesel generators (1.0 means 100 percent reliable, .95 means 95 percent reliable) / time in minutes needed to establish alternate ac power / severe weather category
Modification Summary: Physical changes made to the plant for compliance with the SBO rule
PRA Loop initiating event frequency: Chances per year that the ac power from the electrical grid will be lost.
Number of LOOP events since commercial operation: Actual number of times that ac power from the electrical grid has been lost due to plant problems, due to weather problems, and due to grid problems.
LOOP event recovery times > 240 minutes: The time, in minutes, when the electrical grid has been lost for longer than 4 hours.
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March 17, 2011 • 2 notes • 1 Comment
NRC’s Record in 2010: A UCS Assessment| by David Wright | nuclear power | nuclear power safety | Japan nuclear |
We had prepared and were planning to release a new report this week that assesses how the US Nuclear Regulatory Commission (NRC) did in 2010 at its job of ensuring that US nuclear power plants are operated as safely as possible. Since this assessment seems more relevant than ever, we decided to go ahead with the release.
The report, The NRC and Nuclear Power Plant Safety in 2010: A Brighter Spotlight Needed, was written by UCS’s Dave Lochbaum, a nuclear engineer by training who worked at nuclear power plants for 17 years. This is the first of these assessments, which we plan to do annually.
The report looks at 14 “near-misses” at US nuclear plants during 2010 and gives examples of both effective and ineffective responses to them by the NRC. It concludes the NRC can be effective, but that there are a variety of shortcomings, such as inadequate training, faulty maintenance, poor design, and failure to investigate problems thoroughly, that need to be fixed to make reactors operating in the US as safe as possible.
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