LIGHTNING: Here is a response from INPO.

Click on the image to enlarge; hit return to get back here.
Lightning is covered in several entries in my other blog, http://nuclearpowerblog.blogspot.com/.

Right now, the system will not allow me to edit that blog, so please go there for the story of LIGHTNING that is relevant to this letter from INPO.

Below is my response to the above letter.

Robert H. Leyse
P. O. Box 2850
Sun Valley, ID 83353

August 17, 2008

Ronn K. Smith
INPO
Atlanta, GA

Dear Ronn:

Thank you for your letter of August 4, 2008, responding to my request for INPO SER 76-84. I received your letter on August 14, 2008. There was a delay because it was addressed to my street address instead of my P. O. Box.

My feeling is that the long-standing INPO policy is OK; however, your board should consider releasing documents that are aged and insensitive. Also, when NRC references a specific INPO SER in its public documents, that specific INPO SER should then be released to the public.

Now that I have told you how to run INPO, let’s get back to my narrow case. What has driven me nuts for decades is the INPO summary rejection of my NSAC/INPO SIGNIFICANT EVENT, SALEM 1, which was posted by NSAC on 12-AUG-2000. INPO, in a knee-jerk reaction, immediately (within hours) “suggested” the deletion of this entry.

I became aware of the INPO “suggestion” on 27 Aug 1980 and I told NSAC to “…send the completed form to INPO.” I never knew until 3-11-82 that NSAC had trashed my NSAC/INPO SIGNIFICANT EVENT, SALEM 1.

Ronn, maybe for now, INPO may answer the following question: Is the Salem 1 event of 06-08-00 included in INPO SER 76-84?

As an aside, NRC denied my request for the stuff under FOIA. I’ve appealed that and we’ll see what happens.

Where is she today? She was useful in 1957.

Click to enlarge.

Even with enlargement, the caption is hard to decipher. Joyce is, "one of the hippiest girls in the world." She is 18 and smart and goes to Drexel. The men are described as only a few years older than Joyce, "... members of the new generation that is unlocking door after door to reveal the secrets of atomic energy and harness them to useful purposes."

Maybe the youngest is Don Hughes at the far right. However, relative to age 18, those men are really old goats.

Advertising never changes, only the jargon -- hippiest?

Heat Transfer Experiments; Rectangular Channels

Heat Transfer Experiments: Rectangular Channels



General Electric screwed-up. Westinghouse did not. So, let's first look at the correct way to to the job.
Westinghouse (Bettis) issued the following report that includes the correct design for a test section to determine the heat transfer characteristics of a rectangular channel.

The cross section below is the Westinghouse (correct) design for the heat transfer test section.



And the illustration below shows a neat way of installing the test section in apparatus for testing at high pressure (note the insulation to prevent leakage of electric current).
Now that you have seen the correct design, you may read the following that describes the errors in the GE design that led to an erroneous burnout heat transfer correlation. I was at GE (Vallecitos) less than four months when I wrote the following:

My boss at Vallecitos advised me to keep this close to my chest. Well, I talked a bit. Too bad. Several years later when politics pushed me out of Vallecitos, I certainly had no chance for a softer job in San Jose. HOW THINGS WORK!

What I Should Have Said in Less Than Two Minutes

On February 20-21, 2007, the DOE's Nuclear Energy Research Advisory Committee held an open meeting in Idaho Falls, Idaho. They discussed several activities related to nuclear power developments worldwide. Time was allotted for public comments and I discussed Regulation by Myth for about eight minutes. However, following is what I should have said in less than two minutes.

Gentlemen, you are likely eager to bail out of here after the foregoing endless sets of slide presentations. Today I have only virtual slides. Slide one lists eight nuclear power plants that have applied ultrasonic fuel cleaning. Slide two shows the ultrasonic fuel cleaning equipment. Slide three has two photographs of fouling on nuclear power plant fuel rods. Slide four shows that with fouling, light water reactors have operated with fuel cladding surface temperatures in the range of 1200 degrees Fahrenheit and above (instead of the range of 550 to 600 degrees Fahrenheit that is the design basis for long term operation). Slide five is fantastic and non-existent; it shows the growth of Zirconium dioxide scale and also the increase in dissolved oxygen concentration in the cladding with time at six operating temperatures: 600, 800, 1100, 1500, 2000 and 2500 degrees Fahrenheit. To produce slide five, more experiments are needed with Zircaloy cladding, as called for by the AEC Commissioners decades ago.

A Real Small Break LOCA that could not be isolated

It was a good thing that I was at the operating station about 53 years ago at National Reactor Testing Station (NRTS) in the desert near Idaho Falls, Idaho. The Materials Testing Reactor (MTR) was about a year or so in operation and I had the job of installing and starting up the first pressurized water loop in the MTR for prototype Nautilus fuel. With about three years of solid engineering experience in the nuances of in-pile testing, I inherited this job somewhat by default. The Nautilus was running as well as prototype reactors in the Idaho desert, so why test prototype fuel? However, the project was underway when I went to work at Argonne National Laboratory near Chicago during November 1952 and testing of the prototype fuel proceeded by inertia.

We completed fabrication and crude shakedown testing of the loop at Argonne, Chicago, and shipped the gear to NRTS. With expert assistance and design improvement by key personnel from Phillips Petroleum Company, the operating contractor for MTR, installation and startup of the equipment proceeded in record time. So the loop was operating with the prototype Nautilus fuel in place, when I had my baptism in managing an accident. A 1/4 inch diameter sensing line at the main flow metering orifice severed and blow down from 2500 PSI and 570 degrees F. began. I was at the control panel and the first indication of trouble was a loss of flow signal that let to switching of the primary pump to a standby pump, followed immediately by a scram (fast shutdown) of the MTR.

I very quickly knew where the break was, and I knew that it could not be isolated. I decided that the best approach would be to allow the primary pumps to automatically shift from A to B to C and back to A, etc., as the blowdown proceeded. An option would have been to turn off the pumps, but that would have led to a real loss of cooling for the prototype Nautilus fuel while it still had a relatively high level of decay heat. Of course, this option was not an option forever. At some point, the pressurizer would be empty and voiding and real loss of flow would quickly follow.

My first step was to turn on a low capacity piston pump that delivered about one gallon per hour to the loop. I valved this flow to the liquid level reference sensing standpipe for the pressurizer. I knew that the loop was operating with a concentration of dissolved hydrogen for corrosion control. I also knew that the cold reference sensing line acted as a cold thumb and the the hydrogen concentration on the standpipe would be at saturation for 2500 PSI and its temperature of under 100 degrees F. And I knew that as pressure was reduced during blowdown the hydrogen would begin to outgas in the standpipe and that water would thus be bubbled out of the standpipe. The liquid level signal would then be false and the pressurizer would be empty even though the instrumentation would indicate otherwise. I also knew that the modest flow into the standpipe would not upset the accuracy of the calibration of the level sensing system.

So, I allowed the loop to blow down and depressurize, and as the level approached 10 percent of full, in quick order I proceeded as follows: I turned off the primary pumps. Next, with assistance from Fred McMillan of Phillips Petroleum Company, I isolated the in-pile pressure tube assembly from the main loop (two valves) and then opened the cooling of the in-pile assembly to once-through cooling by process water (two more valves).

Of course, it is not always great for one's career to be on the scene of an accident. And it makes little difference if one's moves were somewhat lifesaving. A few months later I was back at the home office at Argonne, Chicago. At one point I was asked what good I thought a gallon per hour of injected flow would do in the blowdown situation. Well, I did not answer.

Decades later, the value of a degassed reference standpipe (or the safety problem with a hydrogen saturated standpipe) was not recognized by some very highly paid consultants as well as the equipment suppliers of huge nuclear power plants.

Blind faith in single tube tests in the production of TRACE

The following text in italics is copied from the transcript of the full ACRS meeting on February 1, 2007. It is a very small part of the part of the transcript that covers TRACE, however, it reveals more than the previous lengthy discussion of the TRACE activities.

MEMBER ABDEL-KHALIK: But philosophically, if you had a perfect code, and you understand the physics, then it doesn't matter what the scale is because you're verifying phenomena. And therefore, by this process, you're essentially saying the code is nothing more than an empirical fitting tool for the experimental data. Is that true?

MEMBER BANERJEE: It cannot predict new phenomena.

MEMBER ABDEL-KHALIK: Because you are limiting the range of applicability of the code, essentially, to a rather narrow range around where the experiment is. So the code, you philosophically by doing this, you're viewing the code as nothing more than an empirical fitting tool.

MR. BAJOREK: I think that's an accurate statement.

MEMBER POWERS: Do you really want to say that though? I think that's what he was getting at.

MEMBER BANERJEE: It's not predictive of new phenomena.

MR. BAJOREK: That's the -- these codes are not based on first principles. They are based on and held together by closure relations which are based on sub-scale experiments. A lot of those correlations come from single tube tests and you are using that at faith when you start to look at larger and larger scales. Assessment helps to benchmark and let you know whether those correlations are truly applicable with those other conditions but going back to the experiments, we all in integral tests in particular, you want to try to establish a basis for that system global-wide behavior and is it going to behave much like you'd expect in something with much larger scale. But the smaller scale test, that's all you have to run the full test.

MEMBER BANERJEE: As we come to full scale tests.

MR. BAJOREK: If we had full scale tests the --

MEMBER BANERJEE: The assemble system, we can do it in components.

MR. BAJOREK: Components, yes. That's all

The complete transcript from which the above was extracted may be found at:

http://www.nrc.gov/reading-rm/doc-collections/acrs/tr/fullcommittee/2007/ac020107.pdf

However, all of the single tube tests, and also the larger scale tests were conducted with clean (unfouled) heat transfer surfaces. Faith in those tests is blind faith. More later on this.

It is not a TRACE; it is a mamouth SMOKESCREEN

Somewhat periodically the NRC's ACRS reviews activities in the production of the massive so-called thermal hydraulics code, TRACE. The most recent review by the ACRS Thermal-Hydraulics Phenomenon Subcommittee , December 5, 2006, page 9, includes the following remark by MEMBER WALLIS: We recommended in our research report that TRACE becomes the tool for the agency. We recommend TRACE should actually become the mature code used by the agency all over the place and we wanted to see it mature and you say it's going to be universal documentation in 2007, but was sent to us to review seemed to be a hodgepodge of all kinds of stuff. What I want to review is a draft final document, not a hodgepodge of stuff which I have to figure out - not even dated. I don't even know whether some of the documents are old or new or what they are. That's not very helpful to us.

The entire transcript may be viewed at:

http://www.nrc.gov/reading-rm/doc-collections/acrs/tr/subcommittee/2006/th120506.pdf

Going back to January 11, 2001, The ACRS issued a letter, Issues Associated With Industry-Developed Thermal-Hydraulics Codes. It is a lengthy tome with a very lengthy appendix.

http://www.nrc.gov/reading-rm/doc-collections/acrs/letters/2001/4781926.html

The appendix includes the following section; I am quoting only the heading and the last sentence:

4. Codes have evolved, but the development process is hard to trace.

This situation supports the need for the staff to have its own code and to maintain a clear record of why design choices were made in its development.

Now, the TRACE racket has been proceeding under various guises for decades. Fortunes have been cast to the winds, not only in the software extravaganzas, but in the vast array of American as well as international test programs. The connections and relevance of the test programs to TRACE is obscure at best. The most recent disclosure of a link between testing and TRACE is from Staudenmeier in his slide presentation to the ACRS Thermal-Hydraulics Phenomenon Subcommittee on February 15, 2005. Two slides follow:

The four test series in the slide above are a very small sample of the vast test programs that have been conducted, largely on the basis that they were needed for code development and proof testing. There is no mention of extensive LOFT and SPERT projects that were conducted decades ago at the presently named Idaho National Laboratory. This 2005 transcript may be viewed at:

http://www.nrc.gov/reading-rm/doc-collections/acrs/tr/subcommittee/2005/th021505.pdf

There is no mention of the "... more than 50 tests ..." that are discussed in my entry of February 20, 2007. Returning to that matter, here are some extractions from the NRC Memo that predates the Staudenmaier slide show by almost one year.

Memo to Matthews/Black-Technical Safety Analysis of PRM-50-76, A Petition for Rulemaking to Amend Appendix K to 10 CFR Part 50 and Regulatory Guide 1.157 - ML041210109. 18 pages
April 29, 2004
Mr. Leyse states that:
“Petitioner is aware that more experiments with Zircaloy cladding have not been conducted on the scale necessary to . . . overcome the impression left from run 9573.”

In the above Memo, the NRC responded to its quote of Leyse as follows:

In the early 1980's, the NRC through Pacific Northwest Laboratories (PNL) contracted with National Research Universal (NRU) at Chalk River, Ontario, Canada to run a series of LOCA tests in the NRU reactor. More than 50 tests were conducted to evaluate the thermal-hydraulic and mechanical deformation behavior of a full length 32-rod nuclear bundle during the heatup, reflood and quench phases of a large break LOCA. Two tests were initially selected (References 17 and 18) for COBRA/TRAC (Reference 19) simulation to assess the applicability of that code. The NRC is reviewing the data from this program to determine the value of using it to assess the current generation of codes such as TRAC-M (Reference 20), now renamed TRACE.

The full ACRS meets March 8, 9 and 10, 2007 and will include the preparation of several ACRS reports in open sessions. One of the reports is "TRACE Thermal-Hydraulic Analysis Code." It will be interesting to find out what ACRS thinks about this chaotic situation.