Talk:Pressurized water reactor

Fukushima

The reactor at Fukushima used PWRs... after the tsunami they couldn't get the backup generators to run in order to restore cooling. I'd like to know which PWR reactors have a robust backup cooling system... do any? Clarafury (talk) 18:34, 25 April 2011 (UTC)clarafury

The reactors at Fukushima are BWR, not PWR. —Preceding unsigned comment added by 147.175.188.25 (talk) 11:09, 6 May 2011 (UTC)

• That and post accident cooling is a little easier in PWRs assuming that they don't draw the bubble into the core. Protonk (talk) 23:45, 6 May 2011 (UTC)

link on bottom of page

there should be a link on the bottom of the page to 'an article' about the different types of reactors - PWR, ..., ..., ... (if such an article exists) —Preceding unsigned comment added by 76.104.195.115 (talk) 04:41, 17 March 2011 (UTC)

Cleanup / Rewrite

I'm planing to do some cleanup and/or rewriting of this article over the next few days. In particular I'm planing to remove a lot of the stuff that is already covered in nuclear reactor leaving the bits that distinguish a PWR from other reactor types ( this appears to be the trend in other articles on special reactor types ). Please feel free to replace any material I remove if you feel it should remain. J.Ring 17:55, 10 September 2006 (UTC)

I'm going to remove the context tag as I think the introduction and general article is now a lot softer and more in line with other articles on various reactor types. J.Ring 20:19, 10 September 2006 (UTC)

I copied the Overview section's first sentence in again since it got lost in the Sep 10 edit. Was there a specific reason for that deletion? (Other than removing things covered elsewhere.) Mentioning the goal first makes the following details comprehensible, especially for non-technical readers. I think in a section called "Overview" it's justified to feature that information, especially if it's only one sentence. Removing redundance is good, but don't overdo it, we shouldn't drive people away by requiring them to read 4000 other articles first ;-) -- 193.99.145.162 19:06, 21 November 2006 (UTC)

I'd like to express my pleasure at seeing the temperature expressed first in kelvins, which is what matters in computing a heat engine's efficiency. DaveyHume (talk) 07:00, 29 November 2017 (UTC)

A lot to do

Quite a lot still to do on this page. For example, a PWR is *not* fully thermalised as this page currently claims. A CANDU is, and a graphite-moderated reactor is, but in a PWR or BWR the neutron loss from capture in the light water means that the core must be as compact as possible, so it's a compromise. Not quite sure how to put this simply.

Quite a lot of the information really belongs in the nuclear fission or nuclear reactor pages, as it applies to any reactor, not just to a PWR. Andrewa 06:19, 21 Aug 2003 (UTC)

• That's possibly true, but what strikes me the most is the paragraph about reaction control through delayed neutrons...that's something I didn't read anywhere else but is very pertinent.--Chealer 02:06, 2004 Nov 10 (UTC)

What's the difference between a VVER and a PWR? I've heard that VVER is the soviet design for PWR plants.

The Enrico Fermi 2 station is a BWR. I'm going to remove it from the list of PWR reactors. I suggest that someone check the other stations listed for similar problems.Sohlemac 18:30, 3 January 2006 (UTC)

The VVER is basically just the soviet disign for a PWR like you said. There may be small differences, but I am not sure what they are. Lcolson 19:25, 3 January 2006 (UTC)

The VVER is classified entirely as a PWR by definition, like the AP1000 is a PWR as well. It's just a matter different kinds of PWRs. theanphibian

I Have a question about reactors in general. I dont understand why the steam needs to be condensed/cooled before going through the cycle again. Really simple version of the entire thing, reaction heats water, water heats other water, steam goes to turbine, then gets cooled, that is the step I dont understand because if the water/steam is cooled there doesn't it just need to be heated again to go back through the turbine? I know going through the turbine it will cool down but why cool it more than it needs to be? doesnt this lower efficency of the reactor? sorry if this is a simple question just dont know much about it. Caleb rosenberg 05:12, 20 February 2006 (UTC)

The most ideal cycle possible is the carnot cycle, which when plotted on a T-s diagram is a box. In order for the greatest second-law efficiency to be obtained we want a real cycle to be as close to this "carnot box" as possible. In a real rankine cycle the condenser forms the "bottom leg" of the box (which is essential what a PWR cycle is). More to the point, by definition a rankine cycle is a heat engine, which must have a hot and cold side. In the rankine cycle the condenser facilitates the heat transfer for the cold side. In laymens terms, if you don't have a condenser the cycle will continue to heat up until something breaks. Not only that, but very little energy would be created from a process like this, because the water would have to remain superheated vapor for the whole cycle, which completely defeats the purpose of a rankine cycle. HTH! Wizard191 01:28, 23 February 2006 (UTC)

Turbines operate most efficiently when the pressure at the inlet is much higher than the pressure at the outlet. When the steam is cooled it condenses to a liquid which lowers the pressure significantly. Thus by lowering the temperature of the water at the outlet, you can dramatically increase the efficiency of the turbine. Also, no turbine can convert 100% of the heat produced into electricity, thus if the reactor is not cooled it would eventually melt. This is called a loss of coolant accident, or LOCA. Btw Wizard, supercritical water is actually a very good working fluid precisely because it doesn't undergo phase changes. For this reason there is much work on Generation IV reactors cooled by supercritical water. 137.205.192.27 21:58, 3 September 2006 (UTC)

Pressure doesn't change as you condense the fluid. And objectively you want the condenser to operate at as low of a pressure as possible, which yes, means that temperature is lower. Now that pressure (the pressure of the condenser) is dictated by the temperature of your cooling water, colder coolant can condense water at a lower pressure, allowing for the use of a larger low pressure turbine. And if you used that idea of not using a condenser at all, then you would have a mixture at the exit of the turbine that was less dense than that coming into the turbine and it would take more energy to pump it back into the reactor that you got from the turbine to begin with. Believe me, EVERYONE has these questions when they take the first reactor heat transfer courses. theanphibian 06:01, 15 April 2007 (UTC)

As a note it is easer to pump water than to pump steam when you are designing a reactor. This makes components significantly cheaper to build. Greenwjam 7:35 29 October 2006

Boric acid

"This is an advantage for the BWR design because boric acid is very corrosive and the complex charging and letdown system is not required."

It is my understanding that boric acid is not all that corrosive (Boric acid calls it a mild acid), but that, over a long time, (very) tiny leaks in the CRDM nozzles (the Alloy 600 sleeves that the control rod drive mecahnisms move in) to the head drip enough boric acid to make a problem. Comments? --nbach 04:17, 5 April 2006 (UTC)

Boric acid is not too corrosive at standard temperature and pressure but at reactor operating temperatures it is. I guess the word "very" is not a very technical term ;-) Perhaps it would be better to state that when boric acid solutions leak onto reactor system components at operating temperatures, corrosion can be a problem. 205.188.117.70 16:12, 5 July 2006 (UTC)

I belive this should be included in how pressureized water reactors work...

Inside a Nuclear Power Plant To build a nuclear reactor, what you need is some mildly enriched uranium. Typically, the uranium is formed into pellets with approximately the same diameter as a dime and a length of an inch or so. The pellets are arranged into long rods, and the rods are collected together into bundles. The bundles are then typically submerged in water inside a pressure vessel. The water acts as a coolant. In order for the reactor to work, the bundle, submerged in water, must be slightly supercritical. That means that, left to its own devices, the uranium would eventually overheat and melt.

To prevent this, control rods made of a material that absorbs neutrons are inserted into the bundle using a mechanism that can raise or lower the control rods. Raising and lowering the control rods allow operators to control the rate of the nuclear reaction. When an operator wants the uranium core to produce more heat, the rods are raised out of the uranium bundle. To create less heat, the rods are lowered into the uranium bundle. The rods can also be lowered completely into the uranium bundle to shut the reactor down in the case of an accident or to change the fuel.

Pulled From http://www.howstuffworks.com/nuclear-power.htm its quite accurate for some apps.

Huh?

"Since the mass of a water molecule is very similar to the size of a neutron..." What on earth does this mean? First, I don't see how "mass" and "size" can be considered meaningfully similar in the first place. Second, a typical water molecule will contain eight neutrons and ten protons, which is obviously far heavier than a single neutron. I don't know what the article is trying to say here, but unfortunately do not have the background knowledge to correct or clarify it. Egomaniac 19:22, 11 December 2006 (UTC)

Ahh thats easy. Its atomic mass. Its like your takeing two bowling balls and smaking them together if they are both the same mass you dissapate the most energy. However if one is significantly larger than the other than the smaller of the two will simply bounce off. The modarator is the H2 in the water molocuele. Hydrogen atoms have close to 1 mass unit as do neutrons => when a neutron strkies a water molocuel and interacts with the hydrogen it sheds energy. => energy released in the form of heat => also slows the neutron => due to a slower moveing partical is more likely to interact with matter than a fast moveing partical you get more interactions => slows the neutron further and genarates more heat => assuming your not useing a fast fission reactor you'll get a slow moveing netron (thermalised) headed back into the core => more likely to interact with fuel in the core => probability states that when it interacts it will release more neutrons and continues the cycle.

This is a VERRY rudamentry decription but it works for the purpous of explanation. greenwjam 31 DEC 2006

Time between refueling

Look, I can't find anything here, in the article on nuclear power, or the article on nuclear plants that states the typical cycle time between refueling outages. THIS IS IMPORTANT. I will add later possibly, but either I just can't find stuff well or this is completely unacceptable to not include.theanphibian 17:41, 31 March 2007 (UTC)

Sub-cooled boiling?

A remark in section Overview reads: the gas laws guarantee that only sub-cooled boiling will occur in the primary loop. The edit comment for that was: There is always boiling in PWR under normal full power operating conditions. Bulk boiling is impossible and so is not boiling at all (15 Apr 07)

I'm reverting that because:

• I find the remark (and comment) incomprehensible. Sub-cooled boiling? Wikipedia doesn't have an entry on it. I think we can't expect the reader to know jargon that even Wikipedia hasn't heard of.
• The remark seems to contradict the 2nd sentence in the article, which says The primary coolant loop is kept under high pressure to prevent the water from boiling.

Please address these two things before putting the "sub-cooled boiling" back in. Thanks. --193.99.145.162 00:39, 26 May 2007 (UTC)

Sub-cooled boiling is a real phenomenon and indeed occurs in a PWR. This means that especially higher up in the core, it will get hot enough in at the surface of the fuel to boil. The bubble comes out off the fuel and recondenses into the bulk liquid. I agree that it is too technical for the article. The order goes like this (as I increase Rx Power):

• Subcooled boiling
• Nucleate boiling
• Critical heat flux (CHF) or Depature from Nucleate Boiling (DNB) or Leidenfrost Point
• Film boiling

When you reach DNB, the temperature increases dramatically. Thus melting/ burning the cladding. To describe this phenomenon as impossible is inaccurate. In fact, the OT delta T trip in Westinghouse reactors is included specifically to prevent this from happening.

I will update all references to boiling to specify film boiling and link the the Leidenfrost page as it has the nicest graphic. Alfredo22 03:58, 26 May 2007 (UTC)

I haven't heard of them listed in that order. As the first comment in this section pointed out, maybe Wikipedia doesn't have an article on sub-cooled boiling, but the important distintion is this:

• subcooled boiling
• bulk boiling

Granted, there are a large number of types of boiling that can occur, but the four you listed above I think all fall into the category of subcooled boiling, the channel itself will not go above the saturation point. Anyway, I think point I don't understand what's being said about film boiling in the reactor, b/c by your categorization above, it's kept at a high pressure to prevent CHF, and most every other kind of boiling for that matter. theanphibian 15:40, 27 May 2007 (UTC)

I'm having problems discerning your background, so forgive me if I start a too simple of a level. The first thing that it is important to say is that liquid water and steam coexist AT the saturation point for a LONG time. It takes as much (or more) energy to convert water to steam (latent heat) as it does to "warm up" the water to the saturation point (sensible heat). As such, the channel can be completely saturated AND completely liquid. This is 0% quality steam. Nucleate boiling would be process of converting the water to steam. Generally, this process would work well with bubbles/slugs/etc. The reactor would be safe in this condition-- in fact a BWR depends on this (I believe they go to about 20% quality-- but I'm not a BWR guy).

If I tried to do that too quickly, then the hot surface would no longer be wetted-- steam would form before the bubble could be swept away. That's film boiling. The channel may still be liquid (or maybe not), but the surface would be covered by steam. That's what a PWR operator is trying to avoid. Once you get that steam blanket, the temperature can shoot up several orders of magnitude.

Now that we have discussed this-- I admit that there are some simplifications here. A so-so academic paper on some of this can be found at http://iron.nuc.berkeley.edu/~bdwirth/Public/NE104B/documents/Expt_T1.pdf

Boiling is necessary in a PWR, as it is such an effective way of moving heat (remember how much energy it takes to boil water? We use that to our advantage). Generally, our channels are below saturation, hence the reference to subcooled boiling. I don't know that all researchers differentiate subcooled and nucleate boiling when under forced flow. The channel isn't what matters so much as the surface of the fuel. Anything else?

Sorry for the confusion about my background, this is Wikipedia so it's an intentional effort by me to not argue by credibility. I understand what you're describing there, the sequence of different kinds of boiling that happens with increasing wall temperature. That's a boiling curve [1]. I'm not familiar with names for each of those regions, they've just used a,b,c,d nomenclature in all my old notes (which look a lot like your link). Your ordering of Subcooled boiling, Nucleate boiling, Critical heat flux, Film boiling still doesn't make sense to me. All those have specific definitions, but don't dictate any boiling sequence.

In terms of the article: both BWRs and LWRs only boil by nucleate boiling under normal operating conditions. The exit coolant from a BWR is a 20-30% saturated mix for the average channel, the exiting coolant for a PWR is a subcooled liquid for all channels. Are we on the same page here?

I also want to bring up the issue of bubbly/churn and whatnot. The boiling regimes are covered in most nuclear engineering curricula [2]. My understanding, however, is that the ONLY kind that ever happens in a BWR or PWR is bubbly flow due to the high pressure. I think that even annular flow is not reached in accident conditions because the film boiling doesn't extend over the entire surface of the fuel, just enough to fail the fuel.

To address the original question of this section, we probably need a boiling curve or something to illustrate this instead of making statements like this:

• The pressure in the primary coolant loop is typically 15-16 Megapascal, which is notably higher than in other nuclear reactors. As an effect of this, the water in the primary loop will not reach film boiling during normal operation and localized boiling will recondense promptly in the bulk fluid. By contrast, in a boiling water reactor the primary coolant is designed to boil.

Because I don't think this is 100% accurate. theanphibian 22:05, 30 May 2007 (UTC)

I'm certainly not trying to convince via credibility-- I certainly don't have a phd in this. I just want to make sure we are on the right page. The names I used were out of my old Heat Transfer book. The big thing with the boiling curve is that it is a power- not temperature- controlled curve. The experiment changes power while measuring temperature. So, you see subcooled boiling first, then nucleate, when you pass the critical heat _flux_, then liquid can no longer make it to the surface and temperature skyrockets. Still, whether we agree about the definition of subcooled boiling or not is irrelevant. I took the term "subcooled boiling" out of the article at the beginning of this. I also took out "boiling will never occur." Film boiling should never occur in a PWR. Agree? Any boiling in a PWR will recondense. Right? Boiling for a BWR is a design feature-- we want to make steam. Boiling for a PWR is just a fact of life. That's all I am trying to say. What about the bullet is not 100% accurate. And what can we say that improve it that people will understand?

My understanding of Annular flow during non-accident conditions is similar to yours. I think I might see it during an accident if I leave Reactor Coolant Pumps on with decreasing primary inventory. Annular flow should be enough to cool the core as the surface of the clad remains wetted. Think surface of clad must be near saturation temperature for the water in that case. After annular then it gets real bad.

Film boiling (DNB or >CHF) is a generally just a concern with Rx still on (or on w/i the past minute). Dryout is really the concern post-trip. Technically if water were to hit it in a dryout condition you would see a transient film boiling/ partial film boiling, but just like the drop on the pan in the stove, as the clad cooled down, the surface would soon become wetted again.

Fundamentals review is fun, but what can we communicate in wiki for the world?

Alfredo22 02:43, 1 June 2007 (UTC)

I'm switching back to left

I didn't mean to imply that you were arguing by credibility. Those are some very good comments. The statement that it is designed so that film boiling does not occur is very agreeable. Plus, "film boiling" will probably be a little more self-descriptive term to a layman.

I also like these and want to see them in the article:

• "Any boiling in a PWR will recondense."
• "Boiling for a BWR is a design feature-- we want to make steam. Boiling for a PWR is just a fact of life."

Possibly rephrased, I don't know. Whatever and whenever annular and the other kinds of flows occur, they don't seem highly relevant to this article, sorry for bringing that up. Anyway, I'll probably edit the article a bit more, that should keep this discussion focused. For one, I want to avoid implying that other reactors operate with film boiling (I'm pretty sure none do), which we both understand of course, but it could be read as such. Happy editing! theanphibian 05:55, 1 June 2007 (UTC)

Hussman36 (talk) 22:59, 14 March 2013 (UTC) A couple of points: It should be sub-cooled nucleate boiling. It is a real phenomenon. What happens is while the bulk water temperature is well below saturation, the water temperature at the fuel clad/water interface exceeds saturation temperature and a bubble forms on a nucleation site within the thermal boundary layer. As the bubble leaves the boundary layer, it collapses back into the coolant.

Moderator - other reactors

I appreciate the comparison to Chernobyl to illustrate the safety advantages - but does a CANDU reference belong here? (I cleaned it up so it correctly says positive void, not temperature coefficient). Given that there is no refence to BWRs or gas reactors, etc. it is my opinion that this is not needed, as it is covered under the void coefficient page. This is the PWR article, and I'm not sure other designs belong here unless we're going to do a full-run comparison. I'll leave this a bit for comment, but I plan to remove this line in a few days if nobody feels strongly otherwise. Revr J (talk) 20:44, 4 July 2008 (UTC)

Less fissile material than required for prompt critical

This claim doesn't look right - the delayed neutron fraction is only a percent or two, so the amount of fissile material would only have to increase by a comparable proportion, making this not a powerful safety feature. I think what is meant is that PWR fuel is not enriched enough to go critical without a moderator, unlike fast reactor fuel which must be around 20% fissile.

It's a safety feature but for a different reason. Reliance on delayed neutrons for criticality results in power and reactivity swings low enough to be controlled easily. Shutdown status without a moderator is a static safety feature, consider this more of a dynamic feature. Protonk (talk) 13:42, 8 September 2009 (UTC)

Yes, no doubt that delayed neutrons are a safety feature, what I'm questioning is the statement "do not contain enough fissile uranium to sustain a prompt critical chain reaction". It's not particularly the quantity or enrichment of the fuel that allows reliance on delayed neutrons. It's an inherent property of the fissile material and neutron spectrum. --JWB (talk) 15:12, 8 September 2009 (UTC)

Well, Beta is a function of the fuel choice. The DNF can change based one state of the reactor (kinda beside the point I guess). Let me take a look at the statement again. Protonk (talk) 16:00, 8 September 2009 (UTC)

thoughts? I'm a little rusty so please change as needed if I've made a grave error. Protonk (talk) 16:04, 8 September 2009 (UTC)

I think this gets rid of what I was objecting to (it might still be somewhere else in the article, not sure) though it wouldn't hurt to actually say beta is a function of the fuel nuclide and neutron energy, and not of enrichment, if that is the case. --JWB (talk) 17:52, 8 September 2009 (UTC)

I don't know how much we want to go down that road. Beta is a function of the fuel choice. Beta_eff is a function of the fuel choice, load and geometry (because we have to consider absorption and other cross sections). Like I said I'm operating on distant memory right now. I'll try to have something more authoritative to say later. Protonk (talk) 18:27, 8 September 2009 (UTC)

 Done I think that gets the point across (along w/ reminding me that the link is through effective neutron lifetime). :) Protonk (talk) 01:10, 9 September 2009 (UTC)

OK, I think that is a well phrased description of how it works - but is it a particular characteristic of PWRs (or LWRs or thermal reactors in general) vs. some other kind of reactors? In this section, would expect comparison of PWRs to others. --JWB (talk) 02:45, 9 September 2009 (UTC)

Ah. That's a good point. It's probably more of a characteristic of the fuel load, which isn't necessarily unique to PWRs. Protonk (talk) 02:51, 9 September 2009 (UTC)

Could move General LWR material to LWR article

Some of the material in this article applies to LWRs in general. At least if highly detailed, we should move it to that article and retain WP:Summary style here. --JWB (talk) 18:55, 10 September 2009 (UTC)

• I'll try to move some. I'm going to make some changes to some of the info here over the next few weeks. Please give me a hand by reducing some of the LWR details, but I would prefer if you comment out blocks of text so I can dig through them without looking in the history. I think eventually this article can expand much more on the LOCA problems PWRs have (versus BWR) and the passive safety differences (temp. vs. void coefficient). Protonk (talk) 19:00, 10 September 2009 (UTC)

Are you thinking of working on the BWR article as well? --JWB (talk) 20:11, 10 September 2009 (UTC)

No, I wasn't. Protonk (talk) 20:16, 10 September 2009 (UTC)

Dense water gets less dense -> this reduces reactivity?

In the Moderator section the explanation seems to go

• Water slows down the fast neutrons
• The more dense the more the neutrons are slowed down
• When the temperature goes up the water expands and becomes less dense
• So if the reactivity in the reactor goes up the chain reaction will be slowed down
• This makes the PWR safe

I don't get how the 4th bullet point, in my rephrasing, follows from the second

Kestasjk (talk) 10:37, 11 April 2010 (UTC)

I agree that the source text is somewhat confusingly worded, but it is true, remember that fast neutrons cannot be absorbed by U-235 nuclei to start the chain-reaction. So although it's a bit odd under conventional logic it makes sense in this context. - Tequila Monster (talk) 22:02, 29 April 2010 (UTC)

No, that's not true, U-235 has a fast neutron fission cross section of a few barns, similar to other actinides. What makes the difference is that the fission cross section goes up to over 100x that value for thermal neutrons, while the absorption cross section of U-238 stays low.

I think Kestasjk is not understanding that this is a negative feedback loop. --JWB (talk) 14:57, 30 April 2010 (UTC)

Error in subtitles animation movie

There is an error contained in the subtitles of the animation video featured in this article (An animation of a PWR power station with cooling towers). There is mention of fusion whereas in a PWR (as for other nuclear reactors) the proces is called fission. I hope signaling this helps in improving the quality of the article. — Preceding unsigned comment added by 109.130.183.62 (talk) 20:54, 6 May 2014 (UTC)

Pressure in the primary coolant loop

The pressure in the primary coolant loop is given as 15-16 megapascals but I question this. If the water is not to boil, wouldn't the pressure have to be above the critical point of 22 megapascals? Biscuittin (talk) 14:01, 25 December 2014 (UTC)

I think I understand. The water actually does boil, quote: "localized boiling occurs and steam will recondense promptly in the bulk fluid". Biscuittin (talk) 14:10, 25 December 2014 (UTC)

Could we have a list of PWR grid power plants

Could we have a list of PWR grid power plants ? - If not in this article then where ? - Rod57 (talk) 13:45, 28 April 2019 (UTC)

Could we have a list of licenced PWR designs

Could we have a list of licenced PWR designs - eg the Westinghouse 4-loop (used at Watts Bar), the three coolant loop Hualong One design, and any others ? - Rod57 (talk) 13:48, 28 April 2019 (UTC)

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