News the EPRI bulletin
Mike Howard said, "This is a tremendous accomplishment for the Nuclear Sector and the Materials Group. The savings to the nuclear industry is worth more than the total annual dues to EPRI’s Nuclear Sector."
August 31, 2007
NRC Approves EPRI Pressurizer Weld Crack Growth Analysis, Maintaining Spring Outage Schedule for Nine PWRs through Efforts of EPRI's Materials Reliability Program
By Christine King
Nine reactors will maintain their scheduled spring 2008 plant outages after receiving Nuclear Regulatory Commission’s notification last week that an accelerated schedule is not required to address potential pressurizer pipe weld cracking. The NRC based their decision on the results of the EPRI Materials Reliability Program’s (MRP) analysis of hypothetical circumferential cracking in welds located in the pressurizer spray, surge and safety/relief lines. The NRC considered moving these outages to some time this year for inspection or mitigation of these weld locations. NRC concerns arose from the discovery of indications of possible circumferential cracks at the Wolf Creek plant during its fall 2006 outage. The NRC had issued Confirmatory Action Letters to the nine utilities that confirmed their individual commitments to accelerate their outages if MRP’s analysis did not provide reasonable assurance that current inspection schedules were sufficient.
On the industry’s behalf, MRP performed finite element analyses to determine if such indications would show detectable leakage prior to structural failure given the potential concern for growing circumferential stress corrosion cracks. Leakage would alert operators to the condition, enabling them to take effective corrective action prior to structural failure of the piping. The investigation reconsidered the typical assumption of maintaining a semi-elliptical flaw shape as the flaw grows in the finite element model. In the MRP analysis, the crack was allowed to grow at each point along the crack front to the degree dictated by the local stress conditions producing an evolving crack shape as the crack penetrates through the pipe wall. This crack shape evolution is protypical to what happens in the field. The results demonstrated that the semi-elliptical flaw shape assumption is overly conservative and that a realistic crack will locally penetrate the pipe wall producing leakage without resulting in total pipe failure.
On August 23, the NRC confirmed that by their evaluation, reasonable assurance of safe operation until the next scheduled refueling outage has been demonstrated. This project has been a excellent opportunity for EPRI's MRP to show how we can work well with the NRC to support both NRC and industry when faced with difficult decisions associated with aging degradation in operating nuclear power plants.
Mike Howard said, "This is a tremendous accomplishment for the Nuclear Sector and the Materials Group. The savings to the nuclear industry is worth more than the total annual dues to EPRI’s Nuclear Sector."
Friday, August 31, 2007
Saturday, August 25, 2007
Petition for Rulemaking filed by Robert H. Leyse
Secretary
U.S. Nuclear Regulatory Commission
Washington, DC, 20555-0001
Attn: Rulemakings and Adjudications Staff
E-mail: secy@nrc.gov
PETITION FOR RULEMAKING
Mandatory Sharing of Information that is of Potential Regulatory Interest
This petition for rulemaking is submitted by Robert H. Leyse. Petitioner requests that the United States Nuclear Regulatory Commission require all of its resident inspectors at each nuclear power plant site to promptly share information with the agency that is of potential regulatory interest to the agency. This petition is at least partially inspired by the NRC press release of August 15, 2007, portions of which are copied below:
No. 07-102 August 15, 2007
NRC ISSUES ORDER TO FIRST ENERGY REGARDINGINFORMATION RELEVANT TO REGULATORY ACTIVITIES
The Nuclear Regulatory Commission has issued a Confirmatory Order to FirstEnergy Nuclear Operating Company (FENOC), regarding the company’s actions to ensure the prompt sharing of information of potential regulatory interest to the agency.
The Order is based on commitments FENOC made following the NRC’s Demand for Information earlier this year. The demand dealt with documents and analyses the company commissioned for non-regulatory purposes, along with delays between FENOC’s receipt of the information and its subsequent submittal to the agency. The analyses include the “Exponent Report,” which discussed corrosion that severely weakened the Davis-Besse reactor vessel’s head in 2002. The report provided conclusions that could have affected NRC-required activities at U.S. commercial reactors.
The Order requires FENOC to take several actions, including:
* Train selected employees by Nov. 30 to recognize and communicate information that could have a regulatory impact;
* Use an outside consultant to review FENOC staff’s sensitivity to potentially important information in January 2008 and 2009, reporting the results and any follow-on actions to the NRC, and;
* Develop a formal review procedure for technical reports created for non-regulatory uses, to ensure recognition of the reports’ possible regulatory impacts.
Following are examples of cases in which at least some resident inspectors either were aware of, or should have been aware of, the need to insure prompt sharing of information of potential regulatory interest to the agency. Apparently the resident inspectors did not promptly share that information with the agency.
1. A proprietary EPRI report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes was reviewed by NRC inspectors. I have been told, “This was an analysis of the fuel crud performed by the Electric Power Research Institute (EPRI). The EPRI analysis evaluated a sample of crud taken directly from the River Bend Station failed fuel. This document was marked proprietary. As such, the inspectors were restricted from disclosing sensitive information contained in the analysis to the general public. Instead, the inspectors sought other available information to provide a description of fuel crud cooling characteristics. The referenced document, Two-Phase Flow and Heat Transfer, D. Butterworth and G. F. Hewitt, Oxford University Press, 1977, suited this purpose.”
Now, it is evident that the proprietary EPRI report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes, that was (apparently temporarily) available to NRC inspectors, has information that is of potential regulatory interest to the agency. It is also evident that the resident inspectors did not share this information with the agency. Therefore, the resident inspectors or other involved NRC personnel must become subject to requirements that are comparable to FENOC’s as is illustrated in the above press release No. 07-102.
2. Ultrasonic Fuel Cleaning has been deployed at several sites and the NRC resident inspectors should have been aware of the equipment and procedures and the relevance of that information as of potential regulatory interest to the agency. There is no evidence that the resident inspectors were aware of the practice, and if they were aware of the practice, there is no evidence that they recognized ultrasonic fuel cleaning and the reasons therefore as being of potential regulatory interest to the agency.
The following disclosures by industry groups NEI and EPRI to NRC were well after the implementation of ultrasonic fuel cleaning at several licensed nuclear power reactors. To date there has been no documented NRC review and approval of the equipment and procedures, although the equipment and procedures have been applied at several nuclear power reactors in addition to those cited below.
Transcript of ACRS Reactor Fuels Subcommittee - Open Session, September 30, 2003, pages 1-152/229-281.
ML032940295
2003-09-30
Ultrasonic Fuel Cleaning was discussed at the above meeting.
10/01/2003 Summary of Meeting With Nuclear Energy Institute Regarding High Burnup Fuel Issues.
ML032930358
2003-10-16
Following is from the above document:
Mr. J. Deshon made a presentation on “Ultrasonic Fuel Cleaning Technology.” He described the current “single channel” and more advanced “dual channel” fuel cleaners. He indicated that these ultrasonic fuel cleaning technologies have been successfully tested at Callaway and South Texas Project, respectively. The use of these fuel cleaners could reduce fuel crud, avoid axial offset anomaly related operational problems, and minimize personnel radiation exposure during refueling outages.
Robust Fuel Program (Slides) Presented at 10/01/2003 Meeting with Nuclear Energy Institute regarding High Burnup Fuel Issues.
ML032810635
2003-10-01
Here are portions of slides from the above set.
PWRs that have now ultrasonically cleaned fuel:
Callaway
Purpose: AOA avoidance
Callaway experienced varying degrees of AOA from Cycle 4 through 11. In combination with reducing fuel duty, CY12 was free of AOA and CY13 has had no indications through 7500 MWD/MTU
South Texas Project Unit I
Purpose: AOA avoidance, allowed a reduction in fuel assembly purchases
* First cleaning performed in October 2002
South Texas Project Unit 2
Purpose: AOA avoidance following SG replacement & up-rate, allowed a reduction in fuel assembly purchases
* First cleaning performed in April 2003
Next PWR planning to ultrasonically clean fuel:
Vogtle Unit I
Planned Date: October 2003
Purpose: AOA avoidance (in anticipation of Injecting Zn) and dose rate reduction
First BWR planning to ultrasonically clean fuel:
Quad Cities Unit 2
Planned Date: Spring 2004
Purpose: Dose rate reduction
* Mock-up testing taking place at Vallecitos, 16 discharged assemblies will be cleaned on a pre-trial basis in Fall 2003. Anticipate cleaning first reload assemblies in Spring 2004.
3. Reports of main condenser performance and fuel element crud at Energy Northwest
This section illustrates information that is available to the public (on a delayed basis). The point is that reactor inspectors never shared the following with the agency as information of potential regulatory interest. Indeed, the reactor inspectors may not have been aware of these reports.
These reports, dated June 2006, are partially buried on the web site for Energy Northwest and are indexed under News and Information, Columbia Generating Station Information, and they may be found as follows:
http://www.energy-northwest.com/downloads/Main%20Condenser.pdf
http://www.energy-northwest.com/downloads/Main%20Condenser%20Addendum%201.pdf
Here is part of the first page of the interesting report on the main condenser that covers events that took place during a decade. The events include fouling (crud build-up) on the nuclear fuel elements.
Columbia Generating Station
Main Condenser
By W. Scott Oxenford, VP Technical Services
This document summarizes longstanding performance issues related to the design and operation of the Main Condenser at Columbia Generating Station and solutions to those challenges.
The following categories summarize the issues in introductory level detail:
1. System Components Overview
2. Condenser Leakage
3. Columbia’s Condenser
4. Columbia Historical Actions
5. Can Columbia Eliminate Condenser Leakage by Eliminating Debris?
6. Industry Data and Experience
7. Solutions
The following paragraph is copied from page 5 of the Main Condenser report:
Columbia’s management has thoroughly reviewed options for managing ongoing condenser challenges. On each occasion, continued operating risks were accepted instead of taking action, primarily to avoid costs and extended outage length.
The photograph that follows is from page 16 of the Main Condenser report. It shows the crud that has formed on the fuel rods. The crud layers are a heat transfer barrier that is not considered in the nuclear plant design or licensing process. Furthermore, with the crud layers, the fuel element has an increased friction factor that is also not considered in the design and licensing process. This is information of potential regulatory interest.
Note: Photograph has not been copied to this blog. It may be viewed at
http://www.energy-northwest.com/downloads/Main%20Condenser.pdf
There is a lot of depth to this reporting by Energy Northwest, but the reporting was never directed to the Nuclear Regulatory Commission.
Of course, there are undoubtedly additional situations that resident inspectors have encountered and have not reported to the agency even though those situations have been of potential regulatory interest to the agency. Clearly, the NRC must train all of its resident inspectors at each nuclear power plant site to recognize information that is of potential regulatory interest to the agency. The resident inspectors must be required to promptly report that information to the agency.
__________________________________
Robert H. Leyse
P.O. Box 2850
Sun Valley, ID 83353
Dated: August 24, 2007
U.S. Nuclear Regulatory Commission
Washington, DC, 20555-0001
Attn: Rulemakings and Adjudications Staff
E-mail: secy@nrc.gov
PETITION FOR RULEMAKING
Mandatory Sharing of Information that is of Potential Regulatory Interest
This petition for rulemaking is submitted by Robert H. Leyse. Petitioner requests that the United States Nuclear Regulatory Commission require all of its resident inspectors at each nuclear power plant site to promptly share information with the agency that is of potential regulatory interest to the agency. This petition is at least partially inspired by the NRC press release of August 15, 2007, portions of which are copied below:
No. 07-102 August 15, 2007
NRC ISSUES ORDER TO FIRST ENERGY REGARDINGINFORMATION RELEVANT TO REGULATORY ACTIVITIES
The Nuclear Regulatory Commission has issued a Confirmatory Order to FirstEnergy Nuclear Operating Company (FENOC), regarding the company’s actions to ensure the prompt sharing of information of potential regulatory interest to the agency.
The Order is based on commitments FENOC made following the NRC’s Demand for Information earlier this year. The demand dealt with documents and analyses the company commissioned for non-regulatory purposes, along with delays between FENOC’s receipt of the information and its subsequent submittal to the agency. The analyses include the “Exponent Report,” which discussed corrosion that severely weakened the Davis-Besse reactor vessel’s head in 2002. The report provided conclusions that could have affected NRC-required activities at U.S. commercial reactors.
The Order requires FENOC to take several actions, including:
* Train selected employees by Nov. 30 to recognize and communicate information that could have a regulatory impact;
* Use an outside consultant to review FENOC staff’s sensitivity to potentially important information in January 2008 and 2009, reporting the results and any follow-on actions to the NRC, and;
* Develop a formal review procedure for technical reports created for non-regulatory uses, to ensure recognition of the reports’ possible regulatory impacts.
Following are examples of cases in which at least some resident inspectors either were aware of, or should have been aware of, the need to insure prompt sharing of information of potential regulatory interest to the agency. Apparently the resident inspectors did not promptly share that information with the agency.
1. A proprietary EPRI report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes was reviewed by NRC inspectors. I have been told, “This was an analysis of the fuel crud performed by the Electric Power Research Institute (EPRI). The EPRI analysis evaluated a sample of crud taken directly from the River Bend Station failed fuel. This document was marked proprietary. As such, the inspectors were restricted from disclosing sensitive information contained in the analysis to the general public. Instead, the inspectors sought other available information to provide a description of fuel crud cooling characteristics. The referenced document, Two-Phase Flow and Heat Transfer, D. Butterworth and G. F. Hewitt, Oxford University Press, 1977, suited this purpose.”
Now, it is evident that the proprietary EPRI report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes, that was (apparently temporarily) available to NRC inspectors, has information that is of potential regulatory interest to the agency. It is also evident that the resident inspectors did not share this information with the agency. Therefore, the resident inspectors or other involved NRC personnel must become subject to requirements that are comparable to FENOC’s as is illustrated in the above press release No. 07-102.
2. Ultrasonic Fuel Cleaning has been deployed at several sites and the NRC resident inspectors should have been aware of the equipment and procedures and the relevance of that information as of potential regulatory interest to the agency. There is no evidence that the resident inspectors were aware of the practice, and if they were aware of the practice, there is no evidence that they recognized ultrasonic fuel cleaning and the reasons therefore as being of potential regulatory interest to the agency.
The following disclosures by industry groups NEI and EPRI to NRC were well after the implementation of ultrasonic fuel cleaning at several licensed nuclear power reactors. To date there has been no documented NRC review and approval of the equipment and procedures, although the equipment and procedures have been applied at several nuclear power reactors in addition to those cited below.
Transcript of ACRS Reactor Fuels Subcommittee - Open Session, September 30, 2003, pages 1-152/229-281.
ML032940295
2003-09-30
Ultrasonic Fuel Cleaning was discussed at the above meeting.
10/01/2003 Summary of Meeting With Nuclear Energy Institute Regarding High Burnup Fuel Issues.
ML032930358
2003-10-16
Following is from the above document:
Mr. J. Deshon made a presentation on “Ultrasonic Fuel Cleaning Technology.” He described the current “single channel” and more advanced “dual channel” fuel cleaners. He indicated that these ultrasonic fuel cleaning technologies have been successfully tested at Callaway and South Texas Project, respectively. The use of these fuel cleaners could reduce fuel crud, avoid axial offset anomaly related operational problems, and minimize personnel radiation exposure during refueling outages.
Robust Fuel Program (Slides) Presented at 10/01/2003 Meeting with Nuclear Energy Institute regarding High Burnup Fuel Issues.
ML032810635
2003-10-01
Here are portions of slides from the above set.
PWRs that have now ultrasonically cleaned fuel:
Callaway
Purpose: AOA avoidance
Callaway experienced varying degrees of AOA from Cycle 4 through 11. In combination with reducing fuel duty, CY12 was free of AOA and CY13 has had no indications through 7500 MWD/MTU
South Texas Project Unit I
Purpose: AOA avoidance, allowed a reduction in fuel assembly purchases
* First cleaning performed in October 2002
South Texas Project Unit 2
Purpose: AOA avoidance following SG replacement & up-rate, allowed a reduction in fuel assembly purchases
* First cleaning performed in April 2003
Next PWR planning to ultrasonically clean fuel:
Vogtle Unit I
Planned Date: October 2003
Purpose: AOA avoidance (in anticipation of Injecting Zn) and dose rate reduction
First BWR planning to ultrasonically clean fuel:
Quad Cities Unit 2
Planned Date: Spring 2004
Purpose: Dose rate reduction
* Mock-up testing taking place at Vallecitos, 16 discharged assemblies will be cleaned on a pre-trial basis in Fall 2003. Anticipate cleaning first reload assemblies in Spring 2004.
3. Reports of main condenser performance and fuel element crud at Energy Northwest
This section illustrates information that is available to the public (on a delayed basis). The point is that reactor inspectors never shared the following with the agency as information of potential regulatory interest. Indeed, the reactor inspectors may not have been aware of these reports.
These reports, dated June 2006, are partially buried on the web site for Energy Northwest and are indexed under News and Information, Columbia Generating Station Information, and they may be found as follows:
http://www.energy-northwest.com/downloads/Main%20Condenser.pdf
http://www.energy-northwest.com/downloads/Main%20Condenser%20Addendum%201.pdf
Here is part of the first page of the interesting report on the main condenser that covers events that took place during a decade. The events include fouling (crud build-up) on the nuclear fuel elements.
Columbia Generating Station
Main Condenser
By W. Scott Oxenford, VP Technical Services
This document summarizes longstanding performance issues related to the design and operation of the Main Condenser at Columbia Generating Station and solutions to those challenges.
The following categories summarize the issues in introductory level detail:
1. System Components Overview
2. Condenser Leakage
3. Columbia’s Condenser
4. Columbia Historical Actions
5. Can Columbia Eliminate Condenser Leakage by Eliminating Debris?
6. Industry Data and Experience
7. Solutions
The following paragraph is copied from page 5 of the Main Condenser report:
Columbia’s management has thoroughly reviewed options for managing ongoing condenser challenges. On each occasion, continued operating risks were accepted instead of taking action, primarily to avoid costs and extended outage length.
The photograph that follows is from page 16 of the Main Condenser report. It shows the crud that has formed on the fuel rods. The crud layers are a heat transfer barrier that is not considered in the nuclear plant design or licensing process. Furthermore, with the crud layers, the fuel element has an increased friction factor that is also not considered in the design and licensing process. This is information of potential regulatory interest.
Note: Photograph has not been copied to this blog. It may be viewed at
http://www.energy-northwest.com/downloads/Main%20Condenser.pdf
There is a lot of depth to this reporting by Energy Northwest, but the reporting was never directed to the Nuclear Regulatory Commission.
Of course, there are undoubtedly additional situations that resident inspectors have encountered and have not reported to the agency even though those situations have been of potential regulatory interest to the agency. Clearly, the NRC must train all of its resident inspectors at each nuclear power plant site to recognize information that is of potential regulatory interest to the agency. The resident inspectors must be required to promptly report that information to the agency.
__________________________________
Robert H. Leyse
P.O. Box 2850
Sun Valley, ID 83353
Dated: August 24, 2007
Monday, August 20, 2007
INFORMATION RELEVANT TO REGULATORY ACTIVITIES
Now, here is a fantastic press release from our NRC!
No. 07-102
August 15, 2007
NRC ISSUES ORDER TO FIRSTENERGY REGARDING INFORMATION RELEVANT TO REGULATORY ACTIVITIES
The Nuclear Regulatory Commission has issued a Confirmatory Order to FirstEnergy Nuclear Operating Company (FENOC), regarding the company’s actions to ensure the prompt sharing of information of potential regulatory interest to the agency.
The Order is based on commitments FENOC made following the NRC’s Demand for Information earlier this year. The demand dealt with documents and analyses the company commissioned for non-regulatory purposes, along with delays between FENOC’s receipt of the information and its subsequent submittal to the agency. The analyses include the “Exponent Report,” which discussed corrosion that severely weakened the Davis-Besse reactor vessel’s head in 2002. The report provided conclusions that could have affected NRC-required activities at U.S. commercial reactors.
The Order requires FENOC to take several actions, including:
Train selected employees by Nov. 30 to recognize and communicate information that could have a regulatory impact;
Use an outside consultant to review FENOC staff’s sensitivity to potentially important information in January 2008 and 2009, reporting the results and any follow-on actions to the NRC, and;
Develop a formal review procedure for technical reports created for non-regulatory uses, to ensure recognition of the reports’ possible regulatory impacts.
The Order will be published shortly in the Federal Register and will be available on the NRC’s Web site at http://www.nrc.gov/reading-rm/doc-collections/enforcement/actions/reactors/. Anyone adversely affected by the Order, other than FENOC, may request a hearing within 20 days of publication. Any request for a hearing must be submitted to the Secretary, U.S. Nuclear Regulatory Commission, ATTN: Chief, Rulemakings and Adjudications Staff, Washington, DC, 20555. Copies of the hearing request must also be sent to the Director, Office of Enforcement and the Assistant General Counsel for Materials Litigation and Enforcement at the same address, to the Regional Administrator for NRC Region III, 801 Warrenville Road, Lisle, IL, 60532-4351, and to FENOC. Hearing requests may also be faxed to 301-415-1101 or e-mailed to hearingdocket@nrc.gov and to OGCMailCenter@nrc.gov.
Saturday, August 18, 2007
A Few $Billion is Trivial, however there is TERROR in TRACE!
The impacts of software terrorism are everywhere. The NRC game of converting rigged experiments such as LOFT, FLECHT and countless others, into fake software that is applied to the licensing of nuclear power plants is a very interesting game. It is a game that legalizes nuclear power accidents with the potential for damages in the $Trillions.
So, the recent $3billion bailout of the hedge fund, Global Equity Opportunities, that is reported in the Financial Times on August 14, 2007, is trivial. The Times reports, "Goldman Sachs is to inject $2bn of its own money to bail out its Global Equities Opportunities hedge fund in an embarrassing admission that its highly computerized funds malfunctioned last week. The investment bank has raised a further $1bn from outside investors to support the $3.6bn GEO fund, which lost about $1.5bn when computer models failed to predict market turbulence."
Of course, all of the nuclear licensing "experts" and their lobbyists will be quick to assert that the turbulence of severe nuclear power accidents is far more predictable than market turbulence. Indeed, the fake computer models that are derived from rigged experiments provide much of the basis for hundreds of billions of dollars in financing of nuclear power plants. And this is more easily traced than the bases of the recent plague associated with complex collateralized debt obligations. However, the financiers in the big leagues have not yet become aware of the TERROR in TRACE.
So, the recent $3billion bailout of the hedge fund, Global Equity Opportunities, that is reported in the Financial Times on August 14, 2007, is trivial. The Times reports, "Goldman Sachs is to inject $2bn of its own money to bail out its Global Equities Opportunities hedge fund in an embarrassing admission that its highly computerized funds malfunctioned last week. The investment bank has raised a further $1bn from outside investors to support the $3.6bn GEO fund, which lost about $1.5bn when computer models failed to predict market turbulence."
Of course, all of the nuclear licensing "experts" and their lobbyists will be quick to assert that the turbulence of severe nuclear power accidents is far more predictable than market turbulence. Indeed, the fake computer models that are derived from rigged experiments provide much of the basis for hundreds of billions of dollars in financing of nuclear power plants. And this is more easily traced than the bases of the recent plague associated with complex collateralized debt obligations. However, the financiers in the big leagues have not yet become aware of the TERROR in TRACE.
Saturday, August 11, 2007
I could not sleep becuase I kept laughing
Of course, I have not been to bed yet. But I am in a sweat to get this into my blog because a few hours ago a sharp lady promised to go to my blog and even enter a comment. But that is not what I am laughing about. Anyway, back in the early 1950s, Andrei Vishinsky had the following response to one of Harry Truman's favorite proposals related to disarmament:
"I could hardly sleep at all last night. . . . I could not sleep because I kept laughing."
Now, so far I have not been a Red, but I never forgot Vishinsky's remark.
What is making me laugh so hard? OK, let's start out with some kind of analogy, and really, analogies have shortcomings. Suppose somebody proposed a law that required federal inspections of ice cream stands. And then suppose that the ice cream stand lobbyists told the feds that such a proposal was absurd and that it should not even be reviewed. So, go to the next step, the guy who proposed the law for inspections finds out that a few years earlier, the ice cream lobby had given its top award to an ice cream stand operator who devised an ingenious system for inspecting ice cream stands. At this point the somebody who proposed the law for inspections begins laughing, and remarks, "I could hardly sleep at all last night. . . . I could not sleep because I kept laughing."
The ice cream stand is a nuclear power plant that operates with a lot of dirt in its system and the proposal is to require attention to that problem via federal regulations. The lobbyist, the Nuclear Energy Institute, declares on August 6, 2007, "The industry position is that the petition for rulemaking submitted by Mr. Leyse is not needed and should not be considered for action by the Nuclear Regulatory Commission." However, back in 2005, the Nuclear Energy Institute issued the following press release:
Nuclear Energy Institute FOR IMMEDIATE RELEASE
Contact:202-739-8000
For Release:May 18, 2005
Employees at Entergy’s River Bend Station Earn Top Industry Practice 'Best of the Best’ Award
The River Bend Station entry was selected from 94 nominations for the team’s innovative technique to obtain pieces, or “flakes,” of residue from fuel rods while submerged in coolant water. This examination was necessary because workers had identified significant residue buildup as the primary cause for fuel failures.“Such examinations were previously only possible through use of shielded enclosures known as hot cells, an expensive and time-consuming technique that requires two to three years from start to finish and an investment of $2 million to $3 million. Entergy’s innovative method can be completed in six months, at a substantial cost saving. More importantly, worker radiation exposure is reduced three rem to 300 millirem,” said Skip Bowman, NEI’s president and chief executive officer.“All U.S. power reactors can use this technique. Indeed, River Bend’s efforts are now serving as a basis for additional incentives that will further the entire industry’s knowledge of residue formation and behavior,” Bowman said.
Of course, analogies are never precise, and this one has flaws. Nevertheless, I am laughing and that is too bad because I need all of the good sleep that I can get!
"I could hardly sleep at all last night. . . . I could not sleep because I kept laughing."
Now, so far I have not been a Red, but I never forgot Vishinsky's remark.
What is making me laugh so hard? OK, let's start out with some kind of analogy, and really, analogies have shortcomings. Suppose somebody proposed a law that required federal inspections of ice cream stands. And then suppose that the ice cream stand lobbyists told the feds that such a proposal was absurd and that it should not even be reviewed. So, go to the next step, the guy who proposed the law for inspections finds out that a few years earlier, the ice cream lobby had given its top award to an ice cream stand operator who devised an ingenious system for inspecting ice cream stands. At this point the somebody who proposed the law for inspections begins laughing, and remarks, "I could hardly sleep at all last night. . . . I could not sleep because I kept laughing."
The ice cream stand is a nuclear power plant that operates with a lot of dirt in its system and the proposal is to require attention to that problem via federal regulations. The lobbyist, the Nuclear Energy Institute, declares on August 6, 2007, "The industry position is that the petition for rulemaking submitted by Mr. Leyse is not needed and should not be considered for action by the Nuclear Regulatory Commission." However, back in 2005, the Nuclear Energy Institute issued the following press release:
Nuclear Energy Institute FOR IMMEDIATE RELEASE
Contact:202-739-8000
For Release:May 18, 2005
Employees at Entergy’s River Bend Station Earn Top Industry Practice 'Best of the Best’ Award
The River Bend Station entry was selected from 94 nominations for the team’s innovative technique to obtain pieces, or “flakes,” of residue from fuel rods while submerged in coolant water. This examination was necessary because workers had identified significant residue buildup as the primary cause for fuel failures.“Such examinations were previously only possible through use of shielded enclosures known as hot cells, an expensive and time-consuming technique that requires two to three years from start to finish and an investment of $2 million to $3 million. Entergy’s innovative method can be completed in six months, at a substantial cost saving. More importantly, worker radiation exposure is reduced three rem to 300 millirem,” said Skip Bowman, NEI’s president and chief executive officer.“All U.S. power reactors can use this technique. Indeed, River Bend’s efforts are now serving as a basis for additional incentives that will further the entire industry’s knowledge of residue formation and behavior,” Bowman said.
Of course, analogies are never precise, and this one has flaws. Nevertheless, I am laughing and that is too bad because I need all of the good sleep that I can get!
Thursday, August 9, 2007
CRUD STUFF
Leyse analyzed BWR crud during 1959 including thermal conductivity
Document
7 Mb View Document or Access Individual Pages - search, view and/or download individual pages
Title
WATER CHEMISTRY AND FUEL ELEMENT SCALE IN EBWR
Creator/Author
Breden, C.R. ; Charak, I. ; Leyse, R.H.
Publication Date
1960 Nov 01
OSTI Identifier
OSTI ID: 4081362
Report Number(s)
ANL-6136
DOE Contract Number
W-31-109-ENG-38
Resource Type
Technical Report
Resource Relation
Other Information: Orig. Receipt Date: 31-DEC-61
Research Org
Argonne National Lab., Ill.
Sponsoring Org
USDOE
Subject
REACTOR TECHNOLOGY; CANNING; CHEMICAL REACTIONS; CONTAMINATION; CORROSION; DECOMPOSITION; DEPOSITS; DISTRIBUTION; EBWR; FAILURES; FISSION PRODUCTS; FUEL CANS; FUEL ELEMENTS; HEAT TRANSFER; HEATING; HIGH TEMPERATURE; LEAKS; MATHEMATICS; MEASURED VALUES; PLATES; QUALITATIVE ANALYSIS; RADIOACTIVITY; RADIOCHEMISTRY; REACTOR CORE; REACTORS; TEMPERATURE; THERMAL CONDUCTIVITY; VOLUME; WATER
Description/Abstract
The first two sections summarize investigations in EBWR concenced with some aspects of water chemistry. The results of many of these investigations have not been previously published in a form given wide distribution. Included are studies of water conditions, corrosion products (composition, activities, transportation, deposition, and distribution), water dissociation, water activity, fission-product release, and build-up of plant activity. The last two sections of the report give the results of studies of the heat tansfer characteristics of fuel-element scale and effects of high-temperature heating on scale removal and fuel element growth. The maximum scale thickness measured was about 0.008 in. Heat-transfer calculations based on scale thermal conductivity measurements indicate the possibility of maximum fuel temperatures as high at 1692/sup o/F at 100-Mw operation of the core. This temperature is in a range where fuel growth, with resulting fuel element distortion and damage, is expected. Observed trends that may alleviate damage are the tendency of scale to flake off in high-heat transfer areas and the restraining effect of cladding on growth of fuel. No satisfactory means has been found to descale the fuel plates. (auth)
Country of Publication
United States
Language
English
Format
Size: Pages: 105
Availability
NTIS
System Entry Date
2007 May 21
See the complete report at
http://www.osti.gov/energycitations/servlets/purl/4081362-fmbN3r/4081362.PDF
Nuclear Energy Institute discloses the following
http://www.nei.org/newsandevents/riverbendaward/
and the Nuclear Power Journal has this
www.npjonline.com/NPJMain.nsf/504ca249c786e20f85256284006da7ab/89609e291af0b7b286257194007576c1?OpenDocument
Following is a Bob Leyse presentation
www.inl.gov/relap5/rius/yellowstone/leyse.pdf
Following is Letter to NRC: PRM-50-84 and MELLLA+
Robert H. Leyse
P. O. Box 2850
Sun Valley, ID 83353
July 27, 2007
Ms. Annette L. Vietti Cook
Secretary
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555-0001
Attention: Rulemaking and Adjudications Staff
Public Comment on PRM-50-84:
PRM-50-84 and MELLLA+
Dear Ms. Vietti Cook
The need to implement PRM-50-84 is clearly illustrated by the NRC’s incomplete evaluations of GENERAL ELECTRIC (GE) LICENSING TOPICAL REPORTS ON MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (MELLLA+) ANDAPPLICABILITY OF GE METHODS TO EXPANDED OPERATING DOMAINS.
The NRC’s evaluators of MELLLA+ completely overlooked the impact of crud deposits on fuel elements. The reviewers of PRM-50-84 should study the ACRS letter of June 22, 2007, Shack to Reyes, ML071760346. Following are quotations from that letter.
A number of design developments enable operation in the MELLLA+ domain. The most important are: (1) fuel design features that accommodate operation at the higher power / lower flow conditions while maintaining acceptable fuel performance; and (2) a new detect and suppress system that provides protection against power and flow oscillations, which may arise more easily at higher powers and lower flows.The Safety Evaluations were very demanding tasks for which the staff should be commended.
The staff performed thorough evaluations and carried out convincing confirmatory analyses where tools were available, such as for the reactor physics and fuel related issues. Unfortunately, the staff did not have the thermal-hydraulic code capability that would have been needed to independently confirm some important parts of the evaluation such as ATWS instability. The TRACE thermal-hydraulic system analysis code has the capabilities needed to address such issues.
As recommended in our March 22, 2007 report, the TRACE code developmental work should be completed expeditiously to enable its incorporation into the regulatory process.
The reviewers of PRM-50-84 may note that among the fuel design features that accommodate operation at the higher power / lower flow conditions there is no allowance for the impact of crud deposits.
The reviewers may further note that the so-called thorough evaluations of MELLA+ by the NRC staff did not include any attention to the impact of crud deposits on fuel related issues. Regarding TRACE, even if it becomes operational, it has no specifications that incorporate the impact of crud deposits.
This letter cites the lack of allowances for crud deposits in the NRC’s evaluations of MELLLA+. Crud deposits are ubiquitous among the worldwide fleet of LWRs, and the issues are of very high safety significance.
Robert H. Leyse
Widespread coverage of tenacious crud: Another Leyse Letter to NRC
July 27, 2007
Ms. Annette L. Vietti Cook
Secretary
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555-0001
Attention: Rulemaking and Adjudications Staff
Public Comment on PRM-50-84
Dear Ms. Vietti Cook
The need to implement PRM-50-84 is clearly illustrated by analysis of the following reference that includes discussions of the wide spread coverage of tenacious crud and the consequent excessively high cladding temperatures and fuel damage:
NRC, "River Bend Station - NRC Problem Identification and Resolution Inspection Report 0500458/2005008," 02/28/06, Report Details, located at: www.nrc.gov, Electronic Reading Room, ADAMS Documents, Accession Number: ML060600503.
This reference discloses that for cycle 8 at River Bend, “General Electric (the fuel vendor) calculated that the cladding surface temperatures approached 1200 oF in localized areas.”
This reference also discloses that during cycle 11 at River Bend, “Rod Bowing: The licensee identified one problem that was unique to the Cycle 11 failures - significant bowing of the failed fuel pins. This was caused by high temperatures over a larger area of the fuel pins. The cladding temperature had been sufficiently high to anneal the metal, change the micro structure of the zircaloy material. The minimum temperature for annealing zircaloy is about 930 oF. The team determined that the wide spread coverage of the tenacious crud likely caused this phenomena.”
However, even with all of its awareness of the excessively high cladding temperatures and rod bowing caused by the wide spread coverage of tenacious crud, the NRC reported that the issue was of very low safety significance, “… during Operating Cycles 8 and 11, the licensee operated the core outside of the specified MCPR limits, as evidence by excessively high cladding temperatures and fuel damage. Because this issue is of very low safety significance …”
Of course, the NRC evaluators of PRM-50-84 should study ML060600503, and track the path to the determination that this issue is of very low safety significance. If the NRC evaluators do that, they will find the following interesting paragraph:
“The team reviewed one technical study that discussed the behavior of crud on the surface of boiler tubes (“Two-Phase Flow and Heat Transfer,” D. Butterworth and G.F. Hewitt, Oxford University Press, 1977). 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, it is a gross exaggeration to assert that in most circumstances, "wick boiling" occurs within the crud of LWRs. And it is clearly erroneous to assert that in the absence of wick boiling a static steam blanket forms on the cladding surface beneath the crud. I have studied the Butterworth report, and it is a compendium of reports; Chapter 15 is a report by R. V. Macbeth, Fouling in Boiling Water Systems. Macbeth is a recognized expert in the field. The conditions at River Bend and other LWRs are far from the relatively pure magnetite deposits that yield effective wick boiling. Indeed, Macbeth reports, “The risk of an excessive temperature occurs when porous magnetite becomes impregnated with compounds other than iron, and this is shown by the results for rod number 33 ICL in Fig. 15.4. In this rod the magnetite structure was extensively blocked or infilled with calcium and silicon compounds. Consequently, the wick boiling process must have been severely restricted and the transfer of heat was mainly by normal conduction.”
River Bend Station issued License Event Report, Thermally-Induced Accelerated Corrosion of BWR Fuel, (ML003692155) that addresses the unusually heavy deposition of crud on fuel bundles during cycle 8. The following is copied from page 4 of the LER:
“Cycle Differences
A synergy among various parameters related to plant chemistry and core operation is required, in conjunction with the iron deposits, to adequately explain the corrosion phenomenon. A review of parameters that changedin any significant way between Cycle 7 and Cycle 8 was performed.
The amount of iron input to the reactor vessel increased by 70% in Cycle 8, versus Cycle 7, due in part to the removal of low cross-linked resins from service in the condensate demineralizers (*SF*). This removal was done because of sulfate bleed-through associated with this particular resin type. An iron oxide crud layer on the fuel provides a means to concentrate soluble elements such as copper.
The amount of copper input to the reactor vessel increased by -30% in Cycle 8 versus Cycle 7, again due to the removal of low cross-linked resins from service in the condensate demineralizers. An additional source of increasing copper is the "blinding" effect of higher iron on the demineralizers copper removal efficiency. Copper has been previously implicated as an agent of local cladding corrosion in the BWR fleet. Analysis of the crud layers indicated that copper had concentrated in the crud layer adjacent to the cladding.
Zinc was injected into the feedwater system in significant quantities for the first time in Cycle 8. However, the amount of zinc injected and ultimately deposited on the fuel was unremarkable, as compared to the BWR fleet experience. There is no known corrosion or corrosive agent concentration mechanism associated with zinc injection. This is not believed to be a factor in the crud formation.
The plant operated in the Maximum Extended Load Line Limit Analysis (MELLLA) domain for the first time following RF-7. While this allowed plant operation at lower overall core flows, the locations of the fuel failures were not the locations of lowest flow. The failure locations show a strong correlation to peak nodal powers (as expected for a duty-related failure mechanism such as corrosion), but do not show such a correlation to low bundle flow. The lower flows due to MELLLA would only be a minor aggravating factor for crud deposition. Bundle inspections at other BWRs with high feedwater iron concentrations and MELLLA operation do not indicate any significant increases in crud levels due to MELLLA operation.”
Now, the LER was issued on May 30, 2000, and the severe crud deposits were discovered more than one year earlier on April 20, 1999. Clearly, the LER was not written without time for substantial reflection and analysis by River Bend staff and perhaps its consultants. It is apparent that with the large copper content, the crud at River Bend was deposited without any significant duration, if any duration, of wick boiling.
Returning to Macbeth, his chapter also discusses the effect of crud deposits on frictional pressure drop. He reported that the effect of crud deposits on frictional pressure drop in single phase adiabatic water flow is surprisingly large, with a friction factor for a crudded surface more than twice that of a clean surface. He added that with boiling, the impact of crud on the friction factor becomes less as the quality increases. However, at River Bend, the crud was so thick that the impact on the friction factor was must have been substantially greater than anything Macbeth considered. The reviewers of PRM-50-84 must recognize the implications of the change in flow distribution throughout the River Bend core as fouling progressed during cycle 8. Certainly, the flow in the very heavily fouled regions was substantially less than calculated for the MELLLA domain.
Returning next to the matter of a steam blanket on the cladding surface, Macbeth did not report that mechanism and I believe his statement is correct; that without wick boiling heat transfer is by normal conduction. So where did the NRC’s Inspection Team get that idea? It turns out that the team also had access to an EPRI proprietary report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes. I have been told, “This was an analysis of the fuel crud performed by the Electric Power Research Institute (EPRI). The EPRI analysis evaluated a sample of crud taken directly from the River Bend Station failed fuel. This document was marked proprietary. As such, the inspectors were restricted from disclosing sensitive information contained in the analysis to the general public. Instead, the inspectors sought other available information to provide a description of fuel crud cooling characteristics. The referenced document, Two-Phase Flow and Heat Transfer, D. Butterworth and G. F. Hewitt, Oxford University Press, 1977, suited this purpose.” Of course, the NRC evaluators of PRM-50-84 must study that EPRI report and include their findings in their evaluation of PRM-50-84. Perhaps the EPRI report does not address the matter of a steam blanket on the cladding surface, however, the idea came from somewhere and the reviewers must address this.
This letter cites only two of the references in PRM-50-84. Crud deposits are ubiquitous among the worldwide fleet of LWRs, and the issues are of very high safety significance.
Robert H. Leyse
Document
7 Mb View Document or Access Individual Pages - search, view and/or download individual pages
Title
WATER CHEMISTRY AND FUEL ELEMENT SCALE IN EBWR
Creator/Author
Breden, C.R. ; Charak, I. ; Leyse, R.H.
Publication Date
1960 Nov 01
OSTI Identifier
OSTI ID: 4081362
Report Number(s)
ANL-6136
DOE Contract Number
W-31-109-ENG-38
Resource Type
Technical Report
Resource Relation
Other Information: Orig. Receipt Date: 31-DEC-61
Research Org
Argonne National Lab., Ill.
Sponsoring Org
USDOE
Subject
REACTOR TECHNOLOGY; CANNING; CHEMICAL REACTIONS; CONTAMINATION; CORROSION; DECOMPOSITION; DEPOSITS; DISTRIBUTION; EBWR; FAILURES; FISSION PRODUCTS; FUEL CANS; FUEL ELEMENTS; HEAT TRANSFER; HEATING; HIGH TEMPERATURE; LEAKS; MATHEMATICS; MEASURED VALUES; PLATES; QUALITATIVE ANALYSIS; RADIOACTIVITY; RADIOCHEMISTRY; REACTOR CORE; REACTORS; TEMPERATURE; THERMAL CONDUCTIVITY; VOLUME; WATER
Description/Abstract
The first two sections summarize investigations in EBWR concenced with some aspects of water chemistry. The results of many of these investigations have not been previously published in a form given wide distribution. Included are studies of water conditions, corrosion products (composition, activities, transportation, deposition, and distribution), water dissociation, water activity, fission-product release, and build-up of plant activity. The last two sections of the report give the results of studies of the heat tansfer characteristics of fuel-element scale and effects of high-temperature heating on scale removal and fuel element growth. The maximum scale thickness measured was about 0.008 in. Heat-transfer calculations based on scale thermal conductivity measurements indicate the possibility of maximum fuel temperatures as high at 1692/sup o/F at 100-Mw operation of the core. This temperature is in a range where fuel growth, with resulting fuel element distortion and damage, is expected. Observed trends that may alleviate damage are the tendency of scale to flake off in high-heat transfer areas and the restraining effect of cladding on growth of fuel. No satisfactory means has been found to descale the fuel plates. (auth)
Country of Publication
United States
Language
English
Format
Size: Pages: 105
Availability
NTIS
System Entry Date
2007 May 21
See the complete report at
http://www.osti.gov/energycitations/servlets/purl/4081362-fmbN3r/4081362.PDF
Nuclear Energy Institute discloses the following
http://www.nei.org/newsandevents/riverbendaward/
and the Nuclear Power Journal has this
www.npjonline.com/NPJMain.nsf/504ca249c786e20f85256284006da7ab/89609e291af0b7b286257194007576c1?OpenDocument
Following is a Bob Leyse presentation
www.inl.gov/relap5/rius/yellowstone/leyse.pdf
Following is Letter to NRC: PRM-50-84 and MELLLA+
Robert H. Leyse
P. O. Box 2850
Sun Valley, ID 83353
July 27, 2007
Ms. Annette L. Vietti Cook
Secretary
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555-0001
Attention: Rulemaking and Adjudications Staff
Public Comment on PRM-50-84:
PRM-50-84 and MELLLA+
Dear Ms. Vietti Cook
The need to implement PRM-50-84 is clearly illustrated by the NRC’s incomplete evaluations of GENERAL ELECTRIC (GE) LICENSING TOPICAL REPORTS ON MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (MELLLA+) ANDAPPLICABILITY OF GE METHODS TO EXPANDED OPERATING DOMAINS.
The NRC’s evaluators of MELLLA+ completely overlooked the impact of crud deposits on fuel elements. The reviewers of PRM-50-84 should study the ACRS letter of June 22, 2007, Shack to Reyes, ML071760346. Following are quotations from that letter.
A number of design developments enable operation in the MELLLA+ domain. The most important are: (1) fuel design features that accommodate operation at the higher power / lower flow conditions while maintaining acceptable fuel performance; and (2) a new detect and suppress system that provides protection against power and flow oscillations, which may arise more easily at higher powers and lower flows.The Safety Evaluations were very demanding tasks for which the staff should be commended.
The staff performed thorough evaluations and carried out convincing confirmatory analyses where tools were available, such as for the reactor physics and fuel related issues. Unfortunately, the staff did not have the thermal-hydraulic code capability that would have been needed to independently confirm some important parts of the evaluation such as ATWS instability. The TRACE thermal-hydraulic system analysis code has the capabilities needed to address such issues.
As recommended in our March 22, 2007 report, the TRACE code developmental work should be completed expeditiously to enable its incorporation into the regulatory process.
The reviewers of PRM-50-84 may note that among the fuel design features that accommodate operation at the higher power / lower flow conditions there is no allowance for the impact of crud deposits.
The reviewers may further note that the so-called thorough evaluations of MELLA+ by the NRC staff did not include any attention to the impact of crud deposits on fuel related issues. Regarding TRACE, even if it becomes operational, it has no specifications that incorporate the impact of crud deposits.
This letter cites the lack of allowances for crud deposits in the NRC’s evaluations of MELLLA+. Crud deposits are ubiquitous among the worldwide fleet of LWRs, and the issues are of very high safety significance.
Robert H. Leyse
Widespread coverage of tenacious crud: Another Leyse Letter to NRC
July 27, 2007
Ms. Annette L. Vietti Cook
Secretary
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555-0001
Attention: Rulemaking and Adjudications Staff
Public Comment on PRM-50-84
Dear Ms. Vietti Cook
The need to implement PRM-50-84 is clearly illustrated by analysis of the following reference that includes discussions of the wide spread coverage of tenacious crud and the consequent excessively high cladding temperatures and fuel damage:
NRC, "River Bend Station - NRC Problem Identification and Resolution Inspection Report 0500458/2005008," 02/28/06, Report Details, located at: www.nrc.gov, Electronic Reading Room, ADAMS Documents, Accession Number: ML060600503.
This reference discloses that for cycle 8 at River Bend, “General Electric (the fuel vendor) calculated that the cladding surface temperatures approached 1200 oF in localized areas.”
This reference also discloses that during cycle 11 at River Bend, “Rod Bowing: The licensee identified one problem that was unique to the Cycle 11 failures - significant bowing of the failed fuel pins. This was caused by high temperatures over a larger area of the fuel pins. The cladding temperature had been sufficiently high to anneal the metal, change the micro structure of the zircaloy material. The minimum temperature for annealing zircaloy is about 930 oF. The team determined that the wide spread coverage of the tenacious crud likely caused this phenomena.”
However, even with all of its awareness of the excessively high cladding temperatures and rod bowing caused by the wide spread coverage of tenacious crud, the NRC reported that the issue was of very low safety significance, “… during Operating Cycles 8 and 11, the licensee operated the core outside of the specified MCPR limits, as evidence by excessively high cladding temperatures and fuel damage. Because this issue is of very low safety significance …”
Of course, the NRC evaluators of PRM-50-84 should study ML060600503, and track the path to the determination that this issue is of very low safety significance. If the NRC evaluators do that, they will find the following interesting paragraph:
“The team reviewed one technical study that discussed the behavior of crud on the surface of boiler tubes (“Two-Phase Flow and Heat Transfer,” D. Butterworth and G.F. Hewitt, Oxford University Press, 1977). 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, it is a gross exaggeration to assert that in most circumstances, "wick boiling" occurs within the crud of LWRs. And it is clearly erroneous to assert that in the absence of wick boiling a static steam blanket forms on the cladding surface beneath the crud. I have studied the Butterworth report, and it is a compendium of reports; Chapter 15 is a report by R. V. Macbeth, Fouling in Boiling Water Systems. Macbeth is a recognized expert in the field. The conditions at River Bend and other LWRs are far from the relatively pure magnetite deposits that yield effective wick boiling. Indeed, Macbeth reports, “The risk of an excessive temperature occurs when porous magnetite becomes impregnated with compounds other than iron, and this is shown by the results for rod number 33 ICL in Fig. 15.4. In this rod the magnetite structure was extensively blocked or infilled with calcium and silicon compounds. Consequently, the wick boiling process must have been severely restricted and the transfer of heat was mainly by normal conduction.”
River Bend Station issued License Event Report, Thermally-Induced Accelerated Corrosion of BWR Fuel, (ML003692155) that addresses the unusually heavy deposition of crud on fuel bundles during cycle 8. The following is copied from page 4 of the LER:
“Cycle Differences
A synergy among various parameters related to plant chemistry and core operation is required, in conjunction with the iron deposits, to adequately explain the corrosion phenomenon. A review of parameters that changedin any significant way between Cycle 7 and Cycle 8 was performed.
The amount of iron input to the reactor vessel increased by 70% in Cycle 8, versus Cycle 7, due in part to the removal of low cross-linked resins from service in the condensate demineralizers (*SF*). This removal was done because of sulfate bleed-through associated with this particular resin type. An iron oxide crud layer on the fuel provides a means to concentrate soluble elements such as copper.
The amount of copper input to the reactor vessel increased by -30% in Cycle 8 versus Cycle 7, again due to the removal of low cross-linked resins from service in the condensate demineralizers. An additional source of increasing copper is the "blinding" effect of higher iron on the demineralizers copper removal efficiency. Copper has been previously implicated as an agent of local cladding corrosion in the BWR fleet. Analysis of the crud layers indicated that copper had concentrated in the crud layer adjacent to the cladding.
Zinc was injected into the feedwater system in significant quantities for the first time in Cycle 8. However, the amount of zinc injected and ultimately deposited on the fuel was unremarkable, as compared to the BWR fleet experience. There is no known corrosion or corrosive agent concentration mechanism associated with zinc injection. This is not believed to be a factor in the crud formation.
The plant operated in the Maximum Extended Load Line Limit Analysis (MELLLA) domain for the first time following RF-7. While this allowed plant operation at lower overall core flows, the locations of the fuel failures were not the locations of lowest flow. The failure locations show a strong correlation to peak nodal powers (as expected for a duty-related failure mechanism such as corrosion), but do not show such a correlation to low bundle flow. The lower flows due to MELLLA would only be a minor aggravating factor for crud deposition. Bundle inspections at other BWRs with high feedwater iron concentrations and MELLLA operation do not indicate any significant increases in crud levels due to MELLLA operation.”
Now, the LER was issued on May 30, 2000, and the severe crud deposits were discovered more than one year earlier on April 20, 1999. Clearly, the LER was not written without time for substantial reflection and analysis by River Bend staff and perhaps its consultants. It is apparent that with the large copper content, the crud at River Bend was deposited without any significant duration, if any duration, of wick boiling.
Returning to Macbeth, his chapter also discusses the effect of crud deposits on frictional pressure drop. He reported that the effect of crud deposits on frictional pressure drop in single phase adiabatic water flow is surprisingly large, with a friction factor for a crudded surface more than twice that of a clean surface. He added that with boiling, the impact of crud on the friction factor becomes less as the quality increases. However, at River Bend, the crud was so thick that the impact on the friction factor was must have been substantially greater than anything Macbeth considered. The reviewers of PRM-50-84 must recognize the implications of the change in flow distribution throughout the River Bend core as fouling progressed during cycle 8. Certainly, the flow in the very heavily fouled regions was substantially less than calculated for the MELLLA domain.
Returning next to the matter of a steam blanket on the cladding surface, Macbeth did not report that mechanism and I believe his statement is correct; that without wick boiling heat transfer is by normal conduction. So where did the NRC’s Inspection Team get that idea? It turns out that the team also had access to an EPRI proprietary report, BWR Fuel Deposit Sample Evaluation, River Bend Cycle 11 Crud Flakes. I have been told, “This was an analysis of the fuel crud performed by the Electric Power Research Institute (EPRI). The EPRI analysis evaluated a sample of crud taken directly from the River Bend Station failed fuel. This document was marked proprietary. As such, the inspectors were restricted from disclosing sensitive information contained in the analysis to the general public. Instead, the inspectors sought other available information to provide a description of fuel crud cooling characteristics. The referenced document, Two-Phase Flow and Heat Transfer, D. Butterworth and G. F. Hewitt, Oxford University Press, 1977, suited this purpose.” Of course, the NRC evaluators of PRM-50-84 must study that EPRI report and include their findings in their evaluation of PRM-50-84. Perhaps the EPRI report does not address the matter of a steam blanket on the cladding surface, however, the idea came from somewhere and the reviewers must address this.
This letter cites only two of the references in PRM-50-84. Crud deposits are ubiquitous among the worldwide fleet of LWRs, and the issues are of very high safety significance.
Robert H. Leyse
Nuclear Energy Institure Opposes PRM-50-84
James H. Riley
DIRECTOR
ENGINEERING
NUCLEAR GENERATION DIVISION
August 3, 2007
Ms. Annette L. Vietti-Cook, Secretary
Rulemaking and Adjudications Staff
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
DOCKETED USMRC
August 3,2007 (4:16pm)
OFFICE OF SECRETARY RULEMAKINGS AND
ADJUDICATIONS STAFF
Subject: Leyse Petition for Rulemaking: PRM-50-84
Project Number: 689
On May 23, 2007, the Federal Register published for comment a petition for rulemaking by Mr. Mark Edward Leyse who submitted this petition pursuant to Title 10 of the Code of Federal Regulations (10CFR) Section 2.802. It requests new regulations to effectively limit the thickness of crud and/or oxide layers on fuel rod cladding surfaces during normal operations, so that compliance with 10CFR50.46 is ensured. The petition also requests that the IVRC amend Appendix K to Part 50 Emergency Core Cooling System (ECCS) Evaluation Models to require that the steady-state temperature distribution and stored energy in the fuel at the onset of a postulated Loss of Coolant Accident (LOCA) be calculated by including the thermal resistance effects of crud and/or oxide layers on the cladding surface. The petition further specifes that these requirements should also apply to any NRC approved best-estimate ECCS evaluation model. Finally, the petition requests that the NRC amend 10CFR50.46 to include a regulation stipulating a maximum allowable percentage of hydrogen content in the cladding.
The Nuclear Energy Institute (NEI) offers the following comments on the petition for rulemaking.
1. It should be noted that the petitioner relies heavily on four abnormal operating experiences at River Bend (1998-99 and 2001-03), Three Mile Island Unit 1 (1995), Palo Verde Unit 2 (1997), and Seabrook Nuclear Operating Unit (1997). These units all experienced localized sections of thick crud formation during normal operation. The Industry has taken corrective actions to mitigate both general and localized crud formation during operation. These actions include developing revisions to existing water chemistry guidelines.
2. It is well recognized that the effects of corrosion on the cladding and grid spacer surfaces and other fuel system structural components need to be considered to ensure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analyses. Guidelines in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), Section 4.2, "Fuel System Design" do not specify an explicit limit on the maximum allowable corrosion thickness. The guidance contained in SRP Section 4.2 does require that the impact of corrosion on the thermal and mechanical performance be considered in the fuel design analysis, when comparing to the design stress and strain limits. For the fuel rod cladding, the effects include:
(I) The heat transfer resistance provided by the cladding oxide and crud layers, thereby increasing cladding and fuel pellet temperatures,
(11) The metal loss as a result of the corrosion reaction, thereby reducing the cladding load carrying ability.
These effects are already considered in the design analyses to ensure that the cladding does not exceed the mechanical design limits e.g. design stress and design strain.
3. Approved fuel performance models are used to determine fuel rod conditions at the start of a postulated LOCA. The impact of crud and oxidation on fuel temperatures and pressures may be determined explicitly or implicitly in the system of models used. 'the impact of crud and oxidation is addressed, since the system of approved models is benchmarked to temperature and fission gas release data which inherently include corrosion up to high burnup levels.
4. For those Pressurized Water Reactor (PWR) cases cited in the petition in which unusual crud patterns and deposits where observed, post-cycle fuel inspection has shown that there was no significant increase in over-all cladding corrosion compared to existing approved corrosion models. Thus, the cladding temperature was not significantly affected by the presence of the crud with the exception of a very limited number of localized damage sites. These localized damage sites are limited both axially and azimuthally such that their thermal resistance effect on the overall fuel temperature and stored energy is small. Furthermore, any damage on the limited surface area of the cladding affected by unusual crud patterns is no different than other types of cladding damage such as fretting wear or secondary
hydriding of leaking rods. Thus, assuming that cladding with localized crud damage has failed using existing fuel acceptance criteria (e.g. SRP 4.2), consequences associated with unusual crud patterns and deposits are no different than the other types of fuel rod failure modes already accounted for in the plant Technical Specification limits. The Reactor Coolant System (RCS) iodine levels in plant Technical Specification limits inherently restrict the number of damaged rods in a core. Crud-affected fuel rods are not expected to have any significant effects on initial core conditions that could affect LOCA consequences. Any impact on LOCA analyses would be negligible.
5. For the one Boiling Water Reactor (BWR) case cited, in River Bend Cycle 8 significant increases in cladding corrosion were observed only in conjunction with unusually heavy tenacious crud formation. Such crud formation occurred only at lower elevations and thus would have had an impact on the initial stored energy in the fuel only for these locations. Whereas it is true that flow through the affected bundles would be reduced leading to higher initial voiding in the upper part of these bundles, this effect is of secondary importance for a postulated LOCA and is within the envelope determined for core operations with reduced core flow. The calculated Peak Clad Temperature (PCT) in a BWR LOCA event is relatively insensitive to the initial stored energy because PCT values that can challenge the licensing limit occur later in the event and are dominated by the balance between the decay heat and the amount of steam cooling after the initial stored energy difference has been mitigated. It is true that a very early peak in the calculated PCT is sensitive to stored energy but this value is seldom the most limiting value and when it is, this peak is far from the licensing limit of 2200 OF. The similar crud anomaly that occurred for River Bend Cycle 11 was generally considered to be less severe than the Cycle 8 occurrence in that the heavier crud deposition was even more localized. Both events were operational experiences and would not have been prevented or mitigated by the imposition of specific licensing limits on crud thickness. After the second event, River Bend implemented specific hardware changes to prevent further high-crud events based on the root cause determination for these anomalies. These changes have been effective to date.
6. It is true that cladding hydrogen content can have an adverse effect on ductile/brittle behavior of zirconium alloy material heated into the beta phase and quenched (as would occur during a typical LOCA scenario). The hydrogen impact on post-quench cladding ductility is a complex function of the oxidation temperature and pre-quench cooling path. The potential impact of hydrogen on the fuel acceptance criteria specified in 10CFR50.46(b) has been recognized for several years and experimental programs are currently underway to assess this impact on current cladding alloys as well as on newer alloys developed to minimize hydrogen build-up during irradiation. Based on the data being generated from several experimental programs, NRC-RES is in the process of preparing the technical basisfor performance-based fuel acceptance criteria in 10CFR50.46 that include the effects of hydrogen.
In summary, the Industry opinion is that the requirement to consider the impact of crud and/or corrosion layers resident on the fuel rod cladding surface is adequately specified within the current regulations and Staff guidance documents used to prepare and review fuel design and plant safety analyses. The specific incidents referenced by Mr. Leyse in his petition were isolated operational events and would not have been prevented by imposition of specific limits on crud thickness. The Industry is actively pursuing root cause evaluations and has developed corrective actions, including specific hardware changes, to mitigate further cases of excessive crud formation. Any effects of cladding hydrogen content will be addressed in upcoming revisions to criteria under preparation by the NRC Staff.
The Industry position is that the petition for rulemaking submitted by Mr. Leyse is not needed and should not be considered for action by the Nuclear Regulatory Commission. If any further discussion is desired, please contact me at (202) 739-8137; jhr@nei.org or Gordon Clefton at (202) 739-8086; gac@nei.org .
Sincerely,
James H. Riley
c: Mr. Michael T. Lesar, Chief, Rulemaking, Directives and Editing Branch, NRC
Mr. Odelli Ozer, Manager, LWR Fuel Reliability and Storage, EPRI
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