One of the most important safety questions in a nuclear power plant is: Can you cool the very hot nuclear fuel in an accident when normal cooling is disrupted? The scientific field best equipped to answer this question is called “thermal hydraulics.”
The first part of the term, “thermal,” relates to heat transfer, such as the movement of heat from the burner on a stove to the water in a pot via the metal of the pot. The second part, “hydraulic,” relates to the flow of a fluid such as water. The combination, “thermal hydraulics,” can be applied to systems where both the flow of fluid and the transfer of heat are important – such as a nuclear power plant.
I work in the NRC’s Office of Nuclear Regulatory Research as part of a team dedicated to expanding our understanding of thermal hydraulics and applying that understanding in nuclear power plant safety. Over time, we’ve put much effort into incorporating existing knowledge into the NRC’s thermal hydraulics computer simulation program, TRACE. This program allows NRC staff to construct computer models of the cooling systems of a nuclear power plant and then simulate accidents such as pipe breaks (but not wildly improbable events such as the considerable destruction caused near the end of a typical superhero action movie).
TRACE is constantly being pushed to become more accurate, reliable and versatile. Universities and test facilities around the world are conducting experiments and accident simulations to collect real-world data that can be used to determine TRACE’s ability to accurately predict specific phenomena. We use the outcomes to update the program as needed to make it more accurate and to better capture certain phenomena.
Sometimes, new safety issues may result in further investigation of certain scenarios and further evolution of TRACE. Ultimately, the goal of this work within the research arm of the NRC is to continuously expand our understanding of situations which may impact the cooling of the nuclear fuel. This knowledge can then be used to ensure that the public and the environment are protected in the unlikely event of an accident at an U.S. nuclear power plant.
U.S. NRC Blog 
Good article. Thanks for the information.
Great article, it is most helpful. I`v been asked to do a model simulation on the very same question: “Can you cool the very hot nuclear fuel in an accident?”
I’m having trouble believing the “unlikely” part of a disaster happening at a U.S. power plant.
Until the NRC talks about core meltdown, steam explosion and who knows what other make-believe processes and use incorrect models, defining the cladding temperature incorrectly for the steam bubble covered state and until the NRC pushes for Hydrogen recombiners for the supressed Hydrogen generation rates calculated by incorrect codes, instead of looking into the real ignition and burning of Zircaloy cladding in the steam there will be no safe nuclear power plant.
Even it would be relatively easy to achieve: the hot leg side injection ports could be used for the depressurization vent lines in the PWRs and the safety relief line in the BWRs. Whenever the state of the reactor is unknown, the forced circulation through the core is lost or the heat transfer to the ultimate heat sink is severed the operators should open the depressurization vent and allow the reactor to depressurize rapidly and the staged injections of borated water to operate. Only question remains: do we have sufficient gravity emergency core cooling water reserves?
If that is the case it sounds good. Arnie Gunderson has had concerns about AP1000’s earthquake stability, have they improved them from those that are being built now?
Each control rod has its own operating mechanism, they are operated in groups under normal conditions but in the event of a reactor trip each will insert into the core independently assisted by a spring and gravity.
What I am calling for is a direct rapid depressurization vent allowing the operators to vent the steam directly out of the upper part of the reactor before a stagnant steam bubble would extend down into the core.
In the PWR the accumulator injection (ECCS) ports connected to the hot leg side could be utilized for such vent, in the BWR the existing safety relief lines could be rerouted to the vent stack.
The use of this rapid depressurization vent is proposed in three cases:(1) the state of the reactor is unknown, (2) the forced circulation through the core is lost or (3) the heat transfer to the ultimate heat sink is severed
And please, add these to the existing plants too with sufficient gravity water reserves.
Hiddencamper, if You blow the steam into the torus and heat-up the water what You want to inject into the reactor, You will cause that a stagnant steam bubble extends down into the core, because You will not be able to pump the boiling water. Does not matter that You vent the steam or not, if You will deplete the water reserves, makes it unusable.
So this is why the proposed by me rapid depressurization vent is different from the existing solution: Blow the steam to the stack and have gravity water reserves for injection. (beside the now usable suppression pool water in the torus, not heated-up.)
Yes, please add some gravity water reserves to Mark I, too!
BWR plants have safety valves located in the steam lines, above the reactor. These valves will manually or automatically open to depressurize the reactor to release steam, and will fully depressurize the reactor under certain conditions (Low water level 1 alarm + low pressure injection systems available). I am talking about the steam IN the reactor steam dome, NOT the containment.
Statement: The AP1000 and ESBWR plant designs are generation 3+ designs which DO utilize gravity fed passive cooling systems, and are walkaway safe for 72+ hours.
Only if not jamed by earthquake or other even. Are many or all rods configured to one support or is each rod independed, knowing the extra cost for independent rods are a problem for profit I would guese not.
Hiddencamper, I am talking about venting the steam from the top of the reactor, above the core and not just venting the containment.
You also cited that the new design incorporates the passive injection. They should.
Why the NRC does not want to talk about upgrading the existing plants, also?
Zircaloy Mass in Fuel Cladding [kg / lb] 16,465/ 36,300 in the PWR and 40,580 /89,500 in BWR from NRC-2012-0022-0002 and NRC-2012-0022-0003.
Zr (91) + 2 H2O (36) = ZrO2 (123) + 2 H2 (4) + 5 MJ/kgZr
Water required for complete reaction for the PWR 16,465 * 36/91 = 6513,6 kg or about 6.5 m3 (available), it produces 16,465 * 123/91 ZrO2 = 22,255 kg zirconium dioxide and 16,465 * 4/91 = 723.7 kg Hydrogen and 82,325 MJ heat. For a 10 second firestorm duration it gives 8GW power… or twice the full power of the reactor…
Water required for complete reaction for the BWR 40,580 * 36/91 = 16053,6 kg or about 16 m3 (available), it produces 40,580 * 123/91 ZrO2 = 54,850 kg zirconium dioxide and 40,580 * 4/91 = 1784 kg Hydrogen and 204,250 MJ heat. For a 10 second firestorm duration it gives 20GW power… or five-six times the full power of the reactor…
Considering that NRC does not require a top of the reactor depressurization vent to prevent the zirconium firestorm in the reactor, the above back of the envelope calculated worst case scenario should be considered.
At Fukushima, they had to get permission for venting (different from US, where operators are required to vent by procedures). They also had problems getting their rupture disks to open, which is something the US plants to not use. They were TRYING to vent, and it failed due to their specific design of the containment vent. I think We both fully agree that venting containment is a critical strategy to protect the plant.
With BWRs, you have a steam void in the core at all times (it is a boiling reactor). There’s no incore temperature monitors. There are only water pressure and water level monitors.
ECCS is designed to automatically perform a rapid depressurization, followed by a transition to low pressure cooling systems. Existing plants do not have the required designs to support passive cooling systems. The RCIC/HPCI/IC steam powered cooling systems are designed to cool the core until power is restored or portable pumps are connected.
The AP1000 and ESBWR plant designs are generation 3+ designs which DO utilize gravity fed passive cooling systems, and are walkaway safe for 72+ hours.
nuclear guy, I thought I spelled out clearly that only in 3 situations I want to give an opportunity to the operators to intervene, prevent the core damage with that dedicated depressurization vent: “Whenever the state of the reactor is unknown, the forced circulation through the core is lost or the heat transfer to the ultimate heat sink is severed”.
In Mark I BWR unfortunately the torus is used for the suppression pool as well as for water reserve for injection and when the cooling of it was lost in Fukushima it resulted in the brake down of RHIC and failure of the core cooling (in a day or so). Here the venting of the steam starting immediately after the loss of the cooling was realized (which was not that radioactive before the core damage) to the dry well and to the vent stack even would help the situation. That is my proposal.
For the PWR the vent is equal to a large LOCA, which not suppose to cause such problems as You describe and vent to the environment should be a possibility, when there is no core damage. What I’m calling for is to provide staged passive injection reserves (or just check, that it is there) all the way to gravity injection, sufficient to cool the core for a relatively long time to prevent damage to the fuel rods.
I expected that someone objects that I did not mention the core exit temperature measurements, which are there to detect the steam bubble in the core. I do not want the operators to wait until they measure above the saturation temperatures there, when there is a real danger of the core damage from the above problems. Vent the steam bubble before it enters into the core, please!
I know that the ECCS is there to handle the design basis accidents. And yes, I think so that one size should fit all, the rapid depressurization and vent with well staged passive injection reserves all the way to the gravity injections should do that. Make the nuclear power plants, reactors unquestionably safe.
Just some information:
In BWRs, the automatic depressurization system will AUTOMATICALLY depressurize the reactor should water drop below level 1 (low-low-low alarm level) and a low pressure coolant pump is activated and ready for injection. This system is highly reliable and will automatically blowdown the core to the suppression pool for low pressure injection.
Now there are some issues with your comment that simply depressurizing the core will fix all problems. First off, depressurizing with a hot containment/suppression pool will catastrophicaly damage your containment. This is undesirable and could greatly complicate the consequences (dose) to the public from an accident. Second, is depressurization will provide some steam cooling during the blowdown, but will also lower your water level much lower than it originally was. If a low pressure cooling system is not available, as was the case in Fukushima, then depressurizing has no benefit, and will just serve to hasten fuel fragmentation and progression of the accident. Depressurization also is only analyzed for 1 use in the life of the vessel, after which a full vessel reanalysis needs to be performed to see if the vessel is still capable of functioning. This isn’t something that you can just do willy nilly, it is only to be there to ensure design basis accidents can be automatically controlled using ECCS.
Loss of heat sink does not immediately cause a steam bubble. This “steam bubble” you talk of, in BWRs, happens when water drops BELOW 2/3rds top of active fuel for some extended period of time. If the normal heat sink is lost, the suppression pool becomes the interim heat sink, either through relief mode valve lifts or safety mode valve lifts. It’s not until most of the core is uncovered that you have an instance where the zirconium is even possible. In all cases, for design accidents, ECCS prevents this from ever happening.
There is only one reference to zirconium in NRC-2012-0022-0004: “These pellets are stacked and sealed inside long, slender, zirconium metal-based alloy (Zircaloy) tubes to form fuel rods”.
Zircaloy Mass in Fuel Cladding [kg / lb] 16,465/ 36,300 in the PWR and 40,580 /89,500 in BWR from NRC-2012-0022-0002 and NRC-2012-0022-0003.
Zr (91) + 2 H2O (36) = ZrO2 (123) + 2 H2 (4) + 5 MJ/kgZr
Water required for complete reaction for the PWR 16,465 * 36/91 = 6513,6 kg or about 6.5 m3 (available), it produces 16,465 * 123/91 ZrO2 = 22,255 kg zirconium dioxide and 16,465 * 4/91 = 723.7 kg Hydrogen and 82,325 MJ heat. For a 10 second firestorm duration it gives 8GW power… or twice the full power of the reactor…
Water required for complete reaction for the BWR 40,580 * 36/91 = 16053,6 kg or about 16 m3 (available), it produces 40,580 * 123/91 ZrO2 = 54,850 kg zirconium dioxide and 40,580 * 4/91 = 1784 kg Hydrogen and 204,250 MJ heat. For a 10 second firestorm duration it gives 20GW power… or five-six times the full power of the reactor…
Considering that NRC does not require a top of the reactor depressurization vent to prevent the zirconium firestorm in the reactor, the above back of the envelope calculated worst case scenario should be considered.
Nuclear guy and The Firestorm Fighter – the issue is that the real process when a stagnant steam bubble forms in the reactor core is an ignition and firestorm of the zirconium-steam reaction as it was modeled in the cited experiment as well as in the PBF SFD tests, most correctly in the scoping test of this later series. Which was indeed the real process in the TMI-2 accident, Chernobyl-4 accident, Paks 2 refueling vessel incident and in Fukushima 1, 2 and 3 reactors.
What I want to achieve is the addition of depressurization vent and operator response to depressurize the reactor and avoid the steam bubble formation in the core of BWR and PWR types, therefore exclude the possibility of the firestorm.
Until the NRC talks about core meltdown, steam explosion and who knows what other make-believe processes and use incorrect models, defining the cladding temperature incorrectly for the steam bubble covered state and until the NRC pushes for Hydrogen recombiners for the supressed Hydrogen generation rates calculated by incorrect codes, instead of looking into the real ignition and burning of Zircaloy cladding in the steam there will be no safe nuclear power plant.
Even it would be relatively easy to achieve: the hot leg side injection ports could be used for the depressurization vent lines in the PWRs and the safety relief line in the BWRs. Whenever the state of the reactor is unknown, the forced circulation through the core is lost or the heat transfer to the ultimate heat sink is severed the operators should open the depressurization vent and allow the reactor to depressurize rapidly and the staged injections of borated water to operate. Only question remains: do we have sufficient gravity emergency core cooling water reserves?
And yes, NRC is responsible for the Fukushima.
“Even “Zirconium being one of the strongest reducing agents in the periodic table” page 148 of NRC-2012-0022-0002 “At the same time, Zr is also reacting with steam from concrete decomposition, producing hydrogen gas,”
Zr + 2H2O = ZrO2 +2H2
But the reaction heat of 5 MJ/kg Zr reacted is missing as well as the reaction of steam not from concrete, but the coolant itself! WHY? NRC does not allow the steam from coolant react with the zirconium, just with the steam from the concrete?! And the nature just follows the orders of NRC?! As we saw it in Fukushima Daiichi, indeed!” – cited one of my comments…
So let us get this straight – you say that NRC does not consider the effects of the auto-catalytic zirconium oxidation during severe accidents, and when asked for specific evidence of why you believe this, you cite for us a technical report about an experiment – sponsored by NRC – designed specifically to help further the NRC’s understanding of what happens to fuel rods during severe accidents and gather data for the purpose of improving the models in 2 of the NRC’s severe accident codes? Really?
I will admit that trying to decipher your logic is quite challenging., but I think I am beginning to see what you are really trying to convey. It sounds like what you want is to have the NRC change the assumptions of the design basis accident so that utilities have to design their plants to prevent or withstand severe accident events. That is quite a different thing than continuing to state that NRC needs to “allow consideration of the effects of the zirconium firestorm”.
All this time, it sounded like you were simply claiming that the NRC somehow was not, but should be considering the presence or importance of clad oxidation as a heatup mechanism and source of hydrogen during severe accidents – as if to imply their engineers don’t somehow know what happens during a severe accident. That seemed ludicrous to me.
Oh, one last bit of advice, and I will make this my last words for this thread. You really should tone down on the “firestorm” rhetoric. When you use words that are intentionally chosen so as to inflame, it makes it really hard for reasonable people to take you seriously.
And how I would change the regulations: 1. Perform experiments to verify that a well timed rapid depressurization in the light water (BWR, PWR) reactors can avoid the ignition (ballooning and burst) of cladding of nuclear fuel in the reactor
2. Design means of well timed rapid depressurization and subsequent passive prolonged flooding
of the nuclear reactor’s core and deployment of these means
3. Organization of international rapid response team and their national civil defense counterparts
for nuclear emergency response coordination
4. Review of channel type reactors to verify the impossibility of development of zirconium fire in
the steam (in CANDU), immediate shut down of RBMKs still operating on three sites in Russia
5. Development of evacuation procedures for the time of completion of no2 and perform mock
evacuations and public trainings
6. New nuclear power plants should be developed with underground sealed containment systems
as per Edward Teller’s suggestion
7. Mandatory release of data collected about the state of reactor during the accident to the public
The temperatures you are referring to, 1700K, is roughly 2600 degrees F.
This is GREATER than the maximum allowable temperature per 10CFR50.46 (2200F)
In other words, the NRC already REQUIRES reactor cores to NEVER EXCEED 2200F during worst case accident conditions. If they never exceed 2200F, they can never exceed 2600F, which is where you consider the start of your “zirconium firestorm”.
10CFR50.46 also REQUIRES that no more than 1% of maximum theoritical hydrogen products occur, and that cladding oxidation NOT exceed 17% of maximum.
Or in other words, your cited evidence does not call into question any deficiencies which can be readily observed in quantified ECCS performance data in US regulations.
Also, the US NRC had nothing to do with the Fukushima disaster, despite your somewhat unusual comment. Fukushima was in Japan, the US NRC does not govern Japan. Fukushima also was not caused by any “zirconium firestorm”. It was caused by a loss of decay heat removal.
I think you should just reword your entire argument to “I don’t like solid fuel nuclear power”, rather than try to construe some technical non-sense to make it sound like there is some big unknown issue that nuclear power experts have been “ignoring” for over 70 years.
The Firestorm fighter does not recognize the difference between the best estimate model with two sided oxidation of solid cladding and the real firestorm ignited and burned in the reactor cores, zirconium reducing the steam, resulting in hydrogen and zirconium dioxide reaction products. Sad, but the NRC longstanding position is the tragic situation, which led directly to the Fukushima disaster.
As what You are calling for an evidence…
“TEST RESULTS
Following the uncovering and dryout during the coolant boilaway, the rods heated at a rate of 2 to 5 K/s until peak cladding temperatures of 1700 K were attained, at which time the autocatalytic oxidation reaction resulted in a temperature excursion (at a rate of 10 to 50 K/s) and hydrogen generation. Peak local cladding temperatures are estimated to have exceeded 2600 K, based on information from thermocouples on the outside of the bundle liner.
The high-temperature oxidation reaction began at the 2.4- to 3.04-m elevation and formed a localized burn front that moved quickly downward as far as the 1.2-m elevation and then steadily upward. The burn front reached the top end caps (3.80m) and ceased 15 min before the end of the test. The oxidation reaction consumed 75% of the total Zircaloy or almost 100% of the Zircaloy in the path of the burn front. The remaining 25% of the Zircaloy was always below or near the bundle water level. The amount of hydrogen generated was 300±30 g, close to the total conversion of the 1.26-g/s makeup coolant flow within the 45-min high-temperature period. The hydrogen flow fluctuated during the 45-min high-temperature period in response to similar fluctuations (10% to 20% relative) in the bundle coolant flow. The peak hydrogen flow was 190 mg/s, which corresponded to an oxidation power of 28 kW. ” FULL-LENGTH HIGH-TEMPERATURE SEVERE FUEL DAMAGE TEST #5 D. D. Lanning N. J. Lombardo W. K. Hensley D. E. Fitzsimmons J. K. Hartwell @ EG&G-Idaho F. E. Panisko April 1988 – Completion Date September 1993 – Publication Date Prepared for U. S. Nuclear Regulatory Commission Under U.S. Department of Energy Contract DE-ACO6-76RLO1830 PNL—6540‖ cited at
http://www.osti.gov/energycitations/product.biblio.jsp?query_id=2&page=0&osti_id=10188341 With description: „Post-test visual examination of one side of the fuel bundle revealed no massive relocation and flow blockage; however, rundown of molten cladding was evident.” – contradicts to the above description of authors that the cladding burned off above the water level with a rate allowed by water flow. A very typical misrepresentation regarding Zircaloy fires by NRC…
Nothing that you say provides any evidence as to why you believe the NRC does not consider the effects of zirconium oxidation. Please point to a specific regulation that you believe should be changed and how you would change it. The fact is that the NRC DOES consider the effects of the “zirconium firestorm”, as you so affectionately, but incorrectly, characterize it.
Have you ever looked at the regulations in 10CFR 50.46? Paragraph 50.46(b)(1) requires that the calculated maximum temperature of fuel element cladding not be ~ greater than 2200’F. (A limit established so as to prevent both runaway clad oxidation and clad embrittlement). In addition, paragraphs 50.46(b) (2) through (b) (5), which contain required limits for calculated maximum cladding oxidation and maximum hydrogen generation, require that calculated changes in core geometry remain amenable to cooling and that long-term decay heat removal be provided.
In addition, there is Regulatory Guide 1.157, which describes models, correlations, data, model evaluation procedures, and methods that are acceptable-to the NRC staff for meeting the requirements for a realistic or best-estimate calculation of ECCS performance during a loss-of-coolant accident. It says the following on page 6:
3.2.5 Metal-Water Reaction Rate
The rate of energy release, hydrogen generation, and cladding oxidation from the reaction of the zircaloy cladding with steam should be calculated in a best-estimate manner. Best-estimate models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analyses. For rods calculated to rupture their cladding during the loss-of-coolant accident, the oxidation of the inside of the cladding should be calculated in a best estimate manner.
3.2.5.1 Model Evaluation Procedure for Metal-Water Reaction Rate.
Correlations to be used to calculate metal-water reaction rates at less than or equal to 1900’F should:
a. Be checked against a set of relevant data,
and
b. Recognize the effects of steam pressure, pre-oxidation of the cladding, deformation during oxidation, and internal oxidation from both steam and U02 fuel.
The data of Reference 11 are considered acceptable for calculating the rates of energy release, hydrogen generation, and cladding oxidation for cladding temperatures greater than 1900 ‘F.
Not in mine or in the one I taught in…
I’m looking forward to additional posts from you – Salute!
Getting far more qualified people involved and especially professionals from outside of the NRC and most importantly from outside of Region IV, is the first step toward answering basic reactor fatigue safety questions that we now know, affect the entire US Nuclear Fleet. If we learned nothing else from the Fukushima tragedy, we now know that when it come to reactor safety, the widest possible public review can only help insure against future nuclear accidents.
Since you are an engineering professional I urge you to read :
“Press Release 13-01-22 ATHOS Validity Questioned, Qualifying Investigation Required”
Validity of ATHOS computer model requires NRR Qualifying Investigation. (3 Pages)
https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=1ltCb57ciXRaOkhK1rhc2BaB0ACXf7MwcSDZZyEAkFDI
or this one for much more in-depth technical information:
“Response to NRR RAI #32 – Technical”
The SCE cannot provide an ACCEPTABLE operational assessment to the NRR, therefore NO RESTART IS POSSIBLE and here ARE THE TECHNICAL REASONS WHY (50 Pages)
https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=0BweZ3c0aFXcFX05DMWxKNmZXUTA
and/or the even the longer paper:
“SCE NRC Presentation analysis + 14 Questions 12-12-17”
Technical document includes 14 questions affecting US Reactor SAFETY, that the NRC, NRR and RES Regulators need to ask SCE at their 12/18/12 NRR/RES Meeting. (78 Pages)
https://docs.google.com/folder/d/0BweZ3c0aFXcFZGpvRlo4aXJCT2s/edit?docId=0BweZ3c0aFXcFRzBqZUJROWRYNlE
Thank You.
So they will keep building plants not knowing the risks till they gather data on failures to be plugged into TRACE. This to me is scary that they do not know the potential outcome of these plant designs but go ahead with the risk. They seem to ignore what they do not understand and hope for the best results. Is this proper engineering and design practices? Not in the Engineering school I went to.
Lots of talk. The reality is the atmosphere, the Pacific Ocean, and the dairy pastures on the west coast are all now contaminated. Here’s a question: are the pumps at Fort Calhoun submersible, since they are on a flood plain? Here’s another: What’s the NRC done about the faked quake test results that whisteblowers reported? Here’s another: How many babies died in the womb, or shortly after birth from Fukushima fallout?
TRACE is a systems analysis code, so it is not used to predict FEI or structural failure. So the issues you discuss are not addressed by TRACE.
Scott Krepel
The assessment database used to determine TRACE’s validity includes “separate effects” tests and “integral effects” tests. The former are experiments that are performed to isolate specific thermal hydraulic phenomena and develop correlations for a variety of different conditions. The latter are run at scaled test facilities that represent specific nuclear plant designs, including all of the relevant systems. These integral effects tests are simulations of accident scenarios, which can be used to develop confidence that TRACE is adequately predicting relevant events during different plant scenarios. The design and operational parameters for specific plants can then be used to generate plant-specific models and investigate accident scenarios. The intent of TRACE is to provide NRC staff with an independent tool to assess applicants’ analysis results. The manuals that describe the modeling and primary uses of TRACE can be found in NRC’s public ADAMS system under ML120060403, including the assessment manuals that describe various assessments performed on TRACE.
I am not currently doing work associated with San Onofre, so I can’t answer these questions. This type of concern would probably be better addressed by the San Onofre review board. As such, they have been passed along, but all members of the public are encouraged to raise any concerns during public hearings, public comment periods, and other opportunities that the NRC provides to solicit feedback.
Scott Krepel
Information on how to obtain TRACE can be found here: http://www.nrc.gov/about-nrc/regulatory/research/obtainingcodes.html .
Scott Krepel
TRACE is a systems analysis code, and is not used to determine the potability of drinking water near nuclear plants.
Scott Krepel
TRACE is a systems analysis code, so it is not used to predict FEI or structural failure. So the issues you discuss are not addressed by TRACE.
Scott Krepel
The key process in 1979 TMI-2 accident, the 1986 Chernobyl-4 accident, the 2003 Paks 2 refueling pond washing vessel incident and the Fukushima Daiichi 1, 2 and 3 reactors March 2011 accidents. A few of us, nuclear engineers were, are fighting for lifetime for the consideration of real processes in the reactor severe accidents.
As I formulated in a comment to US NRC: Consideration of the zirconium-steam reaction and the ignition and intense firestorm in nuclear reactor fuel rods is well overdue. Reevaluating the evidence provided by the TMI-2 reactor accident, Chernobyl-4 reactor accident, and Paks Unit 2 fuel washing incident, (Fukushima Daiichi units 1-4 fuel damage) with consideration of this intense fiery process, will bring us closer to an ultimately safe nuclear power plant design.
Click to access ML103340250.pdf
Also, I called two years ago for a review: If the hydrogen which is generated in the reactor core from the reaction of the steam (coolant) with the zirconium alloy (or other low neutron absorbing metal cladding and other fuel bundle elements) explodes inside the building surrounding the reactor, this detonation still will not cause a break of the pressure boundary of the containment.
Thirty years after the TMI-2 accident and 23 years after the Chernobyl disaster, I feel obligated to formulate this guideline in order to protect the public from further irradiation from the use of nuclear power. The Chernobyl type reactors (RBMK), which are still operating, have to be shut down immediately because they do not satisfy this guideline. Other nuclear reactors operating and future designs shall be reviewed for compliance to this key requirement and the result of such review shall be defining for their future.
http://aladar-mychernobyl.blogspot.com/
Returning to the comment to US NRC http://pbadupws.nrc.gov/docs/ML1033/ML103340250.pdf : „It is a much overdue duty of NRC and IAEA to evaluate the evidence provided by the TMI-2 accident, Chernobyl-4 accident, Paks-2 incident, and related experiments. Evaluating this evidence, one can see that the ignition of the zirconium fire in the steam occurs at a local temperature of the fuel cladding of around 1000-1200’C, [[and that a self-feeding with steam due to the precipitation of eroded fuel pellets and zirconia reaction product from the hydrogen stream into the water pool, causes intense evaporation.]]
There are insignificant differences in the progression of the firestorms that occurred in the TMI-2 reactor severe accident, Paks washing vessel incident, and Chernobyl-4 reactor accident; the later defined only by the amount of zirconium available for the reaction. At the mean time, there are significant similarities in the processes leading to the ignition of the firestorm. In all three of the compared cases, it took several hours of ill-fated actions or in-actions of the operators to cause the ignition condition. Also, there are similarities in the end result of the firestorm; namely, that the extent of the fuel damage is much less than it was predicted from any other severe fuel damage causing scenarios, introduced for explanations. Therefore the fraction of released fission products is significantly less than was anticipated from the fuel melting or a so called “steam explosion” scenario. Also, the fiery steam-zirconium reaction results in a much higher than anticipated (from any other scenarios) rate of Hydrogen production, which in turn requires a review of containment designs.”
Statement: ‘Over time, we’ve put much effort into incorporating existing knowledge into the NRC’s thermal hydraulics computer simulation program, TRACE. This program allows NRC staff to construct computer models of the cooling systems of a nuclear power plant and then simulate accidents such as pipe breaks’
Until TRACE is perfected, why did they build a plant that is not 100% proven safe, surely they did a mock up of the cooling system to test its use, so what went wrong, is it that it costs too much to do it right or is there no other way of cooling.
I think you may be a little confused about hoe a pressurized water reactor plant works.
the dry steam pockets occur in the secondary side of the steam generators, The primary side where the coolant pumps are is pressurized and keeps the water in the system sub cooled. Steam pockets will not occur in the primary side of the S/G tubes under normal operating conditions.
@Aladar Stolmar
Where do you get the notion that there will be a “zirconium firestorm”? It sounds to me like you have been watching too many of the “typical superhero action movies” that the blog post author mentioned.
Science and technology are amazing servants of mankind; they help us to understand the real world, not the imaginary one created by people who do not understand how to perform math or how materials function in action conditions.
Can MHI ATHOS predict the effects of Fluid Elastic Instability (FEI) and Flashing Feedwater HELB Jet Impingement on the steam generators tubes in SONGS Unit 2, when the steam generator is depressurized due to a main steam line break outside containment along with the failure of MSIV to close. If it can, what is the accuracy of prediction plus minus…..Dr. Pettigrew says it is 20 to 50 percent. Also a 2011 research paper shows that cross-flow fluid velocities at the 90-degree portion on both sides of the u-bend in the hot/cold leg compared with the straight portion of the horizontal leg is double during FEI in large U-bends. This velocity overcomes any resistance provided by the straight flat, curved or honey-comb anti-vibration bars MHI is currently testing in the out-of-plane and/or in-plane direction. SONGS 3 had 8 failed tubes at MSLB test pressures with a vapor fraction of 99.6% and the flat bars failed to prevent the tube-to-tube/AVB wear. Low Contact Forces And Manufacturing Dispersion SCE Theory for SONGS Unit 3 FEI is Bogus and is contradicted/disputed by Westinghouse, AREVA, Dr. Pettigrew, John Large and Bill Hawkins. SCE convinced and sweet-talked MHI in building SONGS large CE Replacement Steam Generators with high heat flows, (MHI was forced to accept numerous untested and unanalyzed design changes to accommodate these high steam flows) in a rush and cheap cost. Now MHI is getting beat by NRC and SCE to cover their mistakes. I am sorry to say, MHI accepted this contract and it is time for MHI to get wise and be out of it rather than try to replace the SONGS steam generators free of cost to please SCE/NRC. MHI in order to maintain their business reputation needs to slow their marketing/manufacturing efforts for US APWR and focus on research/testing/validation in 100% mock-up units before going any further.
Dr. Pettigrew, World’s foremost Expert on fluid Elastic Instability stated in 2006, “In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are axisymmetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability.”
Based on this paper, any knowledgeable Steam Generator (SG) Expert in SG with high heat steam flows such as SONGS, should have designed the steam generators to prevent the adverse effects of fluid elastic instability and flow-induced random vibrations by excluding the following components :
1. Low frequency retainer bars, and
2. Use of flat AVB bars supports to restrain vibrations of the U-tubes in the out-of-plane and in-plane direction.
SCE Certified Design Specifications signed by a California Licensed Professional engineer did not specify fluid elastic instability and Mitsubishi did not manufacture steam generators to prevent the adverse effects of fluid elastic instability and flow-induced random vibrations. This could be very well the items, Barbara Boxer is talking about MHI Root cause and MHI/SCE/NRC are covering up under the false pretense of proprietary information. Changing these items or admitting deficiency in these items would have delayed the SGRP Project, was a very expensive change and triggered a NRC Licensing Amendment Process.
Hello Mr. Krepel, Can Trace predict the effects of Fluid Elastic Instability (FEI) and Flashing Feedwater HELB Jet Impingement on the steam generators tubes in SONGS Unit 2 , when the steam generator is depressurized due to a main steam line break outside containment along with the failure of MSIV to close. If it can, what is the accuracy of prediction plus minus…..Dr. Pettigrew says it is 20 to 50 percent. Also a 2011 research paper shows that cross-flow fluid velocities at the 90 degree portion on both sides of the u-bend in the hot/cold leg compared with the straight portion of the horizontal leg is double during FEI in large U-bends. This velocity overcomes any resistance provided by the straight flat or curved bars in the out-of-plane and/or in-plane direction. SONGS 3 had 8 failed tubes at MSLB test pressures with a vapor fraction of 99.6% and the MHI famous flat bars failed to prevent the tube-to-tube/AVB wear. Thanks
The nuclear industry had 70+ years and over 400 plants built to make them safe, I see it taken another 100+ or more years to make them safe only if the cost does not prevent it, this industry seems to reduce costs over safety, no one can trust the owners, designers or contractors.
To: Scott Krepel I urge you to describe how NRC modeling has been able to do anymore that provide educated guesses especially since many other “Experts” have told the NRC that current computer modeling technology cannot provide anything other than information on laboratory “tests” that do not begin to simulate all the variables of the real world! I believe it was Professor Michel Pettigrew that told Commissioner Magwood that same thing at the last NRC San Onofre special meeting held earlier this month in Rockville, MD. Case in point, MHI has built over a hundred SG’s and until San Onofre they have all preformed as designed for the most part and they used their own modeling system that has proven successful at least until San Onofre. Others, including AREVA and Westinghouse/Hitachi said that they had no problems with their own modeling, so what we are left with is that the NRC is approving both designs and operational parameters that now leave too much up to chance because they are not making their specific parameters and/or data available and without those, all calculations are suspect. In short, we have been very lucky to date, San Onofre should serve as a wake up call to the NRC that it needs to tighten up its regulations and their enforcement before the USA suffers a nuclear incident or worse a nuclear accident like Fukushima!
I would be interested to know if TRACE has been used to model multiple (aka cascade, not just a single tube) of SG tube failures along with a MSLB at San Onofre and if not, why not, since it almost happened for real? Remember Unit 3 had 8 RSG tubes fail in-situ pressure testing along with 1 RSG tube in Unit 2 which was in service (until it was shutdown for refueling) at the same time that had 90% wear, well above the 35% safety limit. If these tube all failed due to a large EQ and or MSLB, it would posed a real threat to San Onofre and all those that live nearby!
As a mechanical engineer who performs nuclear power plant design work, I am curious whether TRACE is publicly available? If so, is it free and could a link be provided here in the comments?
Thanks,
After watching the fiasco at SONGS, I’ve been asking “why do main coolant pumps draw suction from the steam generators?” This virtually assures that dry steam pockets will occur. If suction were from the free-flowing RV and discharge to the generators, resistance in the 3/4″ tubes would suppress dry steam pockets.
Scott Krepel,
I would like to know if this safety program TRACE is above and beyond or part of the planned VC Sumner’s Westinghouse AP1000 unit I have heard and read about ? Regarding the use of the water source for the new VC Sumner reactors, I personally am concerned about the potability of the water used to cool the reactors. From what I have read, the concern for the potability of the drinking water supply appears to be secondary to the safety of the reactors in the case of an accident such as a bursting pipe.
James C. Matthews
Columbia, SC
Once the NRC allows the consideration of the real process, the zirconium firestorm in the core, reducing the steam and generating large amount of hydrogen, it will become pressing that the steam bubble covering of the fuel has to be avoided at any cost. Which is possible by venting and depressurizing the reactor and providing sufficient water reserves for gravity injection.