On April 7, 2007, in this blog, I reported that the following reference was not available to the public. I was wrong. It became available on April 3, 2007. Apparently it was unintentionally referenced in the ACRS letter of March 22, 2007, and that led to its release to the public. The document is dated January 31, 2007. Why the long delay in its release? Below is extracted from page 1 of the 5 page document. And it reveals that after decades of spending, TRACE is a mess!
Memo to F. Gillespie,
TRACE V5.0 Documentation and Support - ML070260005. January 31, 2007
MEMORANDUM TO: Frank Gillespie, Executive DirectorAdvisory Committee on Reactor Safeguards
FROM: Farouk Eltawila, Director /RA/Division of Risk Assessment and Special ProjectsOffice of Nuclear Regulatory Research
SUBJECT: TRACE V5.0 DOCUMENTATION AND SUPPORT
We have recently completed work on TRACE Version 5.0 and have made it available to the NRC staff to finalize assessment for this version and perform audit calculations. This is an important but not necessarily final step in the TRACE consolidation and development effort that has been ongoing for several years. The next steps in this development are to complete the documentation and provide the support necessary to make TRACE an efficient and effective tool for agency use.
The purpose of this memorandum is to
a) summarize the current status of TRACE development,
b) identify specific documentation and the expected completion dates,
c) communicate our plans for a peer review
d) list the activities to help integrate TRACE into the regulatory process
Monday, April 9, 2007
Sunday, April 8, 2007
Disgusting
So, I'm repeating myself.
Somebody should collect all of the letters about TRACE and its relatives that the ACRS has written over the decades.
And now a new Chairman, NRC gets yet another duplication of decades of recommendations. He is told that with all of the great video games on the street, TRACE must get up to speed, and we need yet another set of peer reviews.
He is not told about the endless and useless test programs that are purported to feed TRACE.
Somebody should collect all of the letters about TRACE and its relatives that the ACRS has written over the decades.
And now a new Chairman, NRC gets yet another duplication of decades of recommendations. He is told that with all of the great video games on the street, TRACE must get up to speed, and we need yet another set of peer reviews.
He is not told about the endless and useless test programs that are purported to feed TRACE.
Saturday, April 7, 2007
Hiding the mammouth SMOKESCREEN
On February 28, 2007, I blogged, "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."
And now, on April 7, 2007, I observe, indeed it is interesting to note that the ACRS handled the chaotic TRACE situation by hiding its evaluations of the mammoth smokescreen. Indeed, the ACRS produced its report, TRACE Thermal-Hydraulic Analysis Code. However, the report is not available to the public, and there is no transcript of the ACRS proceedings that produced that report. And that is what I mean by Hiding a mammoth SMOKESCREEN!
Following is the very sketchy ACRS letter report in italics (ML070810911) and the reference, the Eltawila memorandum, is not available to the public.
ACRSR-2242
March 22, 2007
The Honorable Dale E. KleinChairman
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
SUBJECT: DEVELOPMENT OF THE TRACE THERMAL-HYDRAULIC SYSTEM ANALYSISCODE
Dear Chairman Klein:
During the 540th meeting of the Advisory Committee on Reactor Safeguards (ACRS), March 8-9, 2007, we completed our report on the development of the TRACE thermal-hydraulic (T/H) system analysis code. We also discussed this matter during our 539th meeting, February 1-3, 2007. Our Thermal-Hydraulic Phenomena Subcommittee discussed this matter on December 5, 2006. During these reviews, we had the benefit of discussions with representatives of the NRC staff and its contractors. We also had the benefit of the document referenced.
RECOMMENDATIONS
1. The schedule for documenting, validating, and peer reviewing TRACE should beaccelerated and the work completed expeditiously.
2. The development of a representative set of TRACE plant models and user testing onapplications should also be accelerated to facilitate timely incorporation of TRACE intothe regulatory process.
BACKGROUND AND DISCUSSION
In the mid-1990s, the Office of Nuclear Regulatory Research, working with the Office of Nuclear Reactor Regulation, determined that the four primary reactor system T/H codes that were in use at that time should be consolidated into one code. These codes included RELAP5 (LOCA), TRAC-P(PWR-LOCA), TRAC-B(BWR LOCA), and RAMONA (BWR Stability).
The models, correlations, and solution methodologies in these codes did not reflect the state-of-the- art and required in-depth review and modification. It was also recognized that they had been designed at a time when computer capabilities were limited and included many structural aspects, such as memory management, that were no longer needed and increased the cost of code maintenance and development. The availability of graphical user interfaces and their wide acceptance also suggested the desirability of incorporating similar capability in NRC codes. All these considerations led to extensive code consolidation, model improvements, and implementation efforts culminating in the development of TRACE .
TRACE is intended to serve as the main tool for confirmatory analyses of a broad range of thermal-hydraulic problems for current and future reactor designs. It has the potential to offer significantly enhanced capabilities for state-of-the-art analyses of thermal-hydraulic issues. Applications include certification of new reactor designs and the regulatory review of power uprates for currently operating reactors. Therefore, the schedule for documenting, validating, and peer reviewing TRACE, as well as the development of plant input decks, should be accelerated. The work should be completed expeditiously to enable the incorporation of the code into the regulatory process.
Sincerely,
/RA/
William J. Shack Chairman
Reference:
1. Memorandum from Farouk Eltawila, Director, Division of Risk Assessment and SpecialProjects, Office of Nuclear Regulatory Research, to Frank Gillespie, Executive Director,ACRS, “TRACE V5.0 Documentation and Support”, January 31, 2007
Note: The following paragraphs are extracted from my other blog, http://nuclearenergyblog.blogspot.com/
It is not a TRACE; it is a mammoth SMOKESCREEN
Wednesday, February 28, 2007
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 are 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. He discussed four test series; FLECHT, CCTF, SCTF, THTF, but these 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 pagesApril 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.
And now, on April 7, 2007, I observe, indeed it is interesting to note that the ACRS handled the chaotic TRACE situation by hiding its evaluations of the mammoth smokescreen. Indeed, the ACRS produced its report, TRACE Thermal-Hydraulic Analysis Code. However, the report is not available to the public, and there is no transcript of the ACRS proceedings that produced that report. And that is what I mean by Hiding a mammoth SMOKESCREEN!
Following is the very sketchy ACRS letter report in italics (ML070810911) and the reference, the Eltawila memorandum, is not available to the public.
ACRSR-2242
March 22, 2007
The Honorable Dale E. KleinChairman
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
SUBJECT: DEVELOPMENT OF THE TRACE THERMAL-HYDRAULIC SYSTEM ANALYSISCODE
Dear Chairman Klein:
During the 540th meeting of the Advisory Committee on Reactor Safeguards (ACRS), March 8-9, 2007, we completed our report on the development of the TRACE thermal-hydraulic (T/H) system analysis code. We also discussed this matter during our 539th meeting, February 1-3, 2007. Our Thermal-Hydraulic Phenomena Subcommittee discussed this matter on December 5, 2006. During these reviews, we had the benefit of discussions with representatives of the NRC staff and its contractors. We also had the benefit of the document referenced.
RECOMMENDATIONS
1. The schedule for documenting, validating, and peer reviewing TRACE should beaccelerated and the work completed expeditiously.
2. The development of a representative set of TRACE plant models and user testing onapplications should also be accelerated to facilitate timely incorporation of TRACE intothe regulatory process.
BACKGROUND AND DISCUSSION
In the mid-1990s, the Office of Nuclear Regulatory Research, working with the Office of Nuclear Reactor Regulation, determined that the four primary reactor system T/H codes that were in use at that time should be consolidated into one code. These codes included RELAP5 (LOCA), TRAC-P(PWR-LOCA), TRAC-B(BWR LOCA), and RAMONA (BWR Stability).
The models, correlations, and solution methodologies in these codes did not reflect the state-of-the- art and required in-depth review and modification. It was also recognized that they had been designed at a time when computer capabilities were limited and included many structural aspects, such as memory management, that were no longer needed and increased the cost of code maintenance and development. The availability of graphical user interfaces and their wide acceptance also suggested the desirability of incorporating similar capability in NRC codes. All these considerations led to extensive code consolidation, model improvements, and implementation efforts culminating in the development of TRACE .
TRACE is intended to serve as the main tool for confirmatory analyses of a broad range of thermal-hydraulic problems for current and future reactor designs. It has the potential to offer significantly enhanced capabilities for state-of-the-art analyses of thermal-hydraulic issues. Applications include certification of new reactor designs and the regulatory review of power uprates for currently operating reactors. Therefore, the schedule for documenting, validating, and peer reviewing TRACE, as well as the development of plant input decks, should be accelerated. The work should be completed expeditiously to enable the incorporation of the code into the regulatory process.
Sincerely,
/RA/
William J. Shack Chairman
Reference:
1. Memorandum from Farouk Eltawila, Director, Division of Risk Assessment and SpecialProjects, Office of Nuclear Regulatory Research, to Frank Gillespie, Executive Director,ACRS, “TRACE V5.0 Documentation and Support”, January 31, 2007
Note: The following paragraphs are extracted from my other blog, http://nuclearenergyblog.blogspot.com/
It is not a TRACE; it is a mammoth SMOKESCREEN
Wednesday, February 28, 2007
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 are 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. He discussed four test series; FLECHT, CCTF, SCTF, THTF, but these 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 pagesApril 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.
Friday, April 6, 2007
It is not a TRACE; it is a mamouth SMOKESCREEN
TRACE review at ACRS in not revealed to the public. The discussions at ACRS as it finalized its repot on TRACE as well as the report itself are not availaable.
This is copied from nuclearenergyblog.blogspot.com entry of
Wednesday, February 28, 2007
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:
Note: The slides have not been copied, the reader is referred to nulearenegyblog.blogspot.com, February 28, 2007.
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.
This is copied from nuclearenergyblog.blogspot.com entry of
Wednesday, February 28, 2007
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:
Note: The slides have not been copied, the reader is referred to nulearenegyblog.blogspot.com, February 28, 2007.
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.
Blind faith in single tube tests in the production of TRACE; This entry is copied from March 3, 2007, nuclearenergyblog.blogspot.com
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 significantly 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.
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 significantly 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.
Sunday, April 1, 2007
The Reality of "wick boiling"
In my earlier entry of today, a copy from my other blog, I quoted an NRC inspection report:
Now, "wick boiling" is credited with occurring "In most circumstances ...".
The following may be more informative: The wick boiling assertion is likely traceable to a proprietary report by EPRI that includes an evaluation of a sample of fuel element scale from River Bend Cycle 11. However, it is unlikely that NRC has reviewed the EPRI report as a topical report that may serve as a basis for plant licensing. And it is unlikely that the EPRI report has been incorporated in proprietary computer codes or versions of TRACE that reactor plant owners may use in licensing.
DeShon of EPRI did not use the term"wick boiling" in his presentation to ACRS during 2003. The proprietary EPRI report discussed above was issued later.
Transcript of ACRS Reactor Fuels Subcommittee - Open Session, September 30, 2003, pages 1-152/229-281. Following is from page 132:
6 You hear about crud. Crud is bad. Well,
7 not all crud is bad because having a little bit of
8 crud on your fuel surface actually enhances heat
9 transfer. It gets your subcooled boiling taking place
10 a little bit better. Subcooled nucleate boiling is a
11 much more effective heat transfer mechanism than
12 forced convection.
In the nuclear power business it has become convenient to refer to fouling and scale as crud. However, the nuclear power jargon does not change the impact of fouling and scale deposits. GOOGLE has several references that discuss "wick boiling." Here are four:
ScienceDirect - Journal of Nuclear Materials : A model of ...
Heat transfer takes place by wick boiling in which water flows through the porous deposit and evaporates into steam at the surface of chimneys. ...linkinghub.elsevier.com/retrieve/pii/S0022311506000729 - Similar pages
Received 4 January 2005; accepted 31 January 2005.
Abstract
A model is described for simulating thermal hydraulic and chemical conditions within fuel crud deposits. Heat transfer takes place by wick boiling in which water flows through the porous deposit and evaporates into steam at the surface of chimneys. The transport and chemistry of dissolved species within the deposit is also modelled. This chemistry includes the equilibrium chemistry of Li/boric acid species, the equilibrium chemistry of Fe/Ni species and the radiolysis chemistry of water. The unique feature of this model is that the chemistry is coupled to the thermal hydraulics via the increase in the saturation temperature with the concentration of dissolved species. This has a profound effect on evaporative heat transfer within thick deposits, leading to conditions that explain the precipitation of LiBO2 and the possible formation of bonaccordite. The model helps understand several crud scrape observations, including why AOA is observed to occur for a crud thickness in the region of 20–30 μm.
Now, "wick boiling" is credited with occurring "In most circumstances ...".
The following may be more informative: The wick boiling assertion is likely traceable to a proprietary report by EPRI that includes an evaluation of a sample of fuel element scale from River Bend Cycle 11. However, it is unlikely that NRC has reviewed the EPRI report as a topical report that may serve as a basis for plant licensing. And it is unlikely that the EPRI report has been incorporated in proprietary computer codes or versions of TRACE that reactor plant owners may use in licensing.
DeShon of EPRI did not use the term"wick boiling" in his presentation to ACRS during 2003. The proprietary EPRI report discussed above was issued later.
Transcript of ACRS Reactor Fuels Subcommittee - Open Session, September 30, 2003, pages 1-152/229-281. Following is from page 132:
6 You hear about crud. Crud is bad. Well,
7 not all crud is bad because having a little bit of
8 crud on your fuel surface actually enhances heat
9 transfer. It gets your subcooled boiling taking place
10 a little bit better. Subcooled nucleate boiling is a
11 much more effective heat transfer mechanism than
12 forced convection.
In the nuclear power business it has become convenient to refer to fouling and scale as crud. However, the nuclear power jargon does not change the impact of fouling and scale deposits. GOOGLE has several references that discuss "wick boiling." Here are four:
ScienceDirect - Journal of Nuclear Materials : A model of ...
Heat transfer takes place by wick boiling in which water flows through the porous deposit and evaporates into steam at the surface of chimneys. ...linkinghub.elsevier.com/retrieve/pii/S0022311506000729 - Similar pages
Received 4 January 2005; accepted 31 January 2005.
Abstract
A model is described for simulating thermal hydraulic and chemical conditions within fuel crud deposits. Heat transfer takes place by wick boiling in which water flows through the porous deposit and evaporates into steam at the surface of chimneys. The transport and chemistry of dissolved species within the deposit is also modelled. This chemistry includes the equilibrium chemistry of Li/boric acid species, the equilibrium chemistry of Fe/Ni species and the radiolysis chemistry of water. The unique feature of this model is that the chemistry is coupled to the thermal hydraulics via the increase in the saturation temperature with the concentration of dissolved species. This has a profound effect on evaporative heat transfer within thick deposits, leading to conditions that explain the precipitation of LiBO2 and the possible formation of bonaccordite. The model helps understand several crud scrape observations, including why AOA is observed to occur for a crud thickness in the region of 20–30 μm.
Energy Citations Database (ECD) - Energy and Energy-Related ...
2) Without considering the concentration effect, wick boiling nullifies the thermal resistance due to the crud deposit and, therefore, ...www.osti.gov/energycitations/product.biblio.jsp?osti_id=6928828 - 14k - Cached - Similar pages
[PDF]
American Nuclear Society winter meeting; 15 Nov 1987; Los Angeles, CA, USA
Abstract
Boiling heat transfer occurs throughout most of the core of boiling water reactors, and at specific core locations in pressurized water reactors (PWRs).^For higher burnups, waterside corrosion of cladding may become a life-limiting feature of fuel rod design, especially when combined with daily load-following operation.^Corrosion of Zircaloy has been extensively investigated.^The major highlights gained from this study include the following: (1) in the absence of boiling, the presence of crud deposits increases the temperature at the Zircaloy/oxide interface and, therefore, accelerates Zircaloy corrosion dramatically, especially for thick crud layers.^(2) Without considering the concentration effect, wick boiling nullifies the thermal resistance due to the crud deposit and, therefore, mitigates significantly the acceleration of Zircaloy corrosion due to crud deposit.^(3) With the presence of LiOH, the concentration effect in the crud deposit due to wick boiling can be very detrimental to the Zircaloy corrosion resistance performance.^(4) The model prediction agrees in trend with limited plant data.
Modeling and Thermal Performance Evaluation of Porous Crud Layers ...
File Format: PDF/Adobe Acrobat - View as HTMLchimneys, an equivalent cylindrical region approximates the surrounding porous. region. Fig. 3. Schematic of the Wick Boiling in the Unit Cell ...www.osti.gov/bridge/servlets/purl/841238-rIfms9/native/841238.pdf - Similar pages
Note: The above reference, June 2005, may be viewed (in total) online. There is no claim that wick boiling enhances the heat transfer of fouling deposits.
A New Model for the Effect of Calcium Sulfate Scale Formation on ...
The transient change in heat transfer is closely related to wick boiling, and the associated changes in bubble departure diameter and bubble site density. ...link.aip.org/link/?JHTRAO/126/507/1 - Similar pages
April 2004
Abstract
Scale deposition on the heat transfer surfaces from water containing dissolved salts considerably reduces fuel economy and performance of heat transfer equipment. This problem is more serious during nucleate boiling due to the mechanisms of bubble formation and detachment. Using a precision pool boiling test apparatus, the effects of heat flux and calcium sulfate concentration on heat transfer coefficient and formation and growth of deposits are investigated. The transient change in heat transfer is closely related to wick boiling, and the associated changes in bubble departure diameter and bubble site density. A physically sound prediction model was developed for the prediction of heat transfer coefficients as a function of time during deposition processes. Based on comparison with experimental data over a wide range of foulant concentrations and heat fluxes, the model is considered to be sufficiently accurate for practical application. ©2004 ASME
None of these four papers claim that wick boiling is a practical long-term enhancement of heat transfer with fouling deposits. The second reference (1987) may come close to this for a very limited case. The final reference may be the most credible with its application of precision pool boiling test apparatus. It observes that with nucleate boiling the problem with scale deposition is more serious.
Regulation by Myth (Copied from http://nucelarenergyblog.blogspot.com)
Monday, February 12, 2007
Regulation by myth
Here are quotes from NRC's inspection report for River Bend, ML060600503, February 28, 2006:
General Electric (the fuel vendor) calculated that the cladding surface temperatures exceeded 1200 F in localized areas.
and
The team noted that the thermal resistance of crud is not normally sufficient to cause cladding temperature increases consistent with those observed during Cycle 8. In most circumstances, "wick boiling" occurs within the crud. That is, capillary coolant channels within the crud deliver coolant to the cladding surface. Steam then escapes from the cladding surface in chimney type plumes. This is a fairly effective method of heat transfer. However, in some instances the capillary coolant channels can become clogged, creating a static steam blanket on the cladding surface. Steam is an exceptionally good thermal insulator. This is the process that caused the very high cladding surface temperatures and ultimately resulted in fuel cladding failure.
Now, "wick boiling" is erroneously credited as occurring "In most circumstances...". Fouling (crud) is ubiquitous among the worldwide fleet of LWRs, however, wick boiling has not been universally acclaimed as occurring in any of the instances.
Moreover, steam blanketing does not occur beneath the crud deposit. At River Bend the crud had a very high thermal resistance and that led to a very high cladding surface temperature, but it is fiction to assert that a steam blanket existed between the surface of the cladding and the crud layer.
So, I sent an e-mail to Chairman NRC advising him to get that paragraph eliminated. I e-mailed, "Indeed that entire paragraph should be deleted in a corrected report. The team should study report ANL 6136."
I should also have told the Chairman to have the inspection team study McAdams, Heat Transmission, 1942, Chapter X, HEAT TRANSFER TO BOILING LIQUIDS. On page 316, "The small amount of scale necessary to reduce a high coefficient by a substantial amount is not generally realized."
Maybe the NRC thinks that crud is not scale!
Regulation by myth
Here are quotes from NRC's inspection report for River Bend, ML060600503, February 28, 2006:
General Electric (the fuel vendor) calculated that the cladding surface temperatures exceeded 1200 F in localized areas.
and
The team noted that the thermal resistance of crud is not normally sufficient to cause cladding temperature increases consistent with those observed during Cycle 8. In most circumstances, "wick boiling" occurs within the crud. That is, capillary coolant channels within the crud deliver coolant to the cladding surface. Steam then escapes from the cladding surface in chimney type plumes. This is a fairly effective method of heat transfer. However, in some instances the capillary coolant channels can become clogged, creating a static steam blanket on the cladding surface. Steam is an exceptionally good thermal insulator. This is the process that caused the very high cladding surface temperatures and ultimately resulted in fuel cladding failure.
Now, "wick boiling" is erroneously credited as occurring "In most circumstances...". Fouling (crud) is ubiquitous among the worldwide fleet of LWRs, however, wick boiling has not been universally acclaimed as occurring in any of the instances.
Moreover, steam blanketing does not occur beneath the crud deposit. At River Bend the crud had a very high thermal resistance and that led to a very high cladding surface temperature, but it is fiction to assert that a steam blanket existed between the surface of the cladding and the crud layer.
So, I sent an e-mail to Chairman NRC advising him to get that paragraph eliminated. I e-mailed, "Indeed that entire paragraph should be deleted in a corrected report. The team should study report ANL 6136."
I should also have told the Chairman to have the inspection team study McAdams, Heat Transmission, 1942, Chapter X, HEAT TRANSFER TO BOILING LIQUIDS. On page 316, "The small amount of scale necessary to reduce a high coefficient by a substantial amount is not generally realized."
Maybe the NRC thinks that crud is not scale!
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