ML13095A236
| ML13095A236 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/03/2013 |
| From: | Arora V - No Known Affiliation |
| To: | Borchardt W, Collins E, Leeds E NRC/Chairman, NRC/EDO, Office of Nuclear Reactor Regulation, NRC Region 4 |
| References | |
| LTR-13-0292 | |
| Download: ML13095A236 (22) | |
Text
OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET Date Printed:Apr 04, 2013 14:29 PAPER NUMBER: LTR- 13-0292 LOGGING DATE: 04/04/2013 ACTION OFFICE:
AUTHOR: Vinod Arora AFFILIATION:
ADDRESSEE: Chairman Resource AA0fa&e) OtVW
SUBJECT:
Concerns disgraceful scheme re: San Onofre Nuclear Generating Station ACTION: Appropriate DISTRIBUTION: RF, SECY to Ack.
LETTER DATE: 04/03/2013 ACKNOWLEDGED Yes SPECIAL HANDLING: Lead office to publicly release 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SECY's assignment, via SECY/EDO/DPC.
NOTES:
FILE LOCATION: ADAMS DATE DUE: DATE SIGNED:
Joosten, Sandy From: Vinod Arora [vinnie48in@gmail.com]
Sent: Wednesday, April 03, 2013 9:07 PM To: CHAIRMAN Resource; Borchardt, Bill; Leeds, Eric; Collins, Elmo; Howell, Art; Dorman, Dan; Benney, Brian; Hall, Randy
Subject:
Disgraceful' scheme for risky nuclear experiment could begin June 1 - Revision 0 - to be updated periodically for the benefit of Concerned Public and Brilliant NRC Questioning Engineers Attachments: Attachment 1.docx; Attachment 2.docx Disclaimer & Acknowledgements I rather would make conservative and honest safety errors (unless proven wrong otherwise by the NRC) for making consistent efforts in pursuit of protecting the health and safety of 8.4 Million Southern Californians. My position is consistent with the papers published by Dr. Michel Pettigrew, Dr. Joram Hopenfeld, Arnie Gundersen, John Large, Professor Daniel Hirsch, Dale Bridgenbaugh, Friends of the Earth, DAB Safety Team, Anonymous and concerned San Onofre Insiders Experts, other steam generator experts, nuclear scientists, MIT, UCLA, European, Korean and Canadian researchers and numerous public safety organizations regarding the dangers of restarting San Onofre's defectively designed and degraded Unit 2 RSGs even at reduced 70% power without complete repairs/replacement.
Disgraceful' scheme for risky nuclear experiment could begin June 1 SCE and its Independent Experts either do not understand the concept of fluid elastic instability, or really not forthcoming with the true facts about the danger of restarting "defectively-designed and degraded" San Onofre Unit 2 replacement steam generators. A tube leak, a near meltdown and destruction of SONGS Unit 3 replacement steam generators, Mitsubishi Root Cause Analysis, Dr.
Joram Hopenfeld, John Large, Arnie Gundersen, Professor Daniel Hirsch and conflicting reports by SCE, Westinghouse, MHI, AREVA, Intertek, NRC Augmented Inspection Team report and SCE responses to NRC Request for Additional Information has already proven beyond the shadow of doubt that these replacement steam generators are not designed to operate at normal steady state 70% power level due to the adverse effects of fluid elastic instability and flow-induced vibrations caused by inadvertent equipment manipulations, operator errors, operational transients and main steam line breaks. In America, the most powerful and democratic country in the world, the appointed and elected government officials with regulatory authority responsible for the nuclear safety of the American Citizens are paid by the tax and rate payers. Under, no circumstances and for no reason, these officials can allow electric production and industry profits over public safety (See Note 1) by granting the New SCE License Amendment. These generators have to be repaired or replaced by a NEI qualified, "US Nuclear Plant Designer" other than MHI, before NRC can certify their safe operation.
From an observation of Draft SCE License Amendment Hearings in Rockville Maryland, it appears that Southern California Edison wants to restart unsafe Unit 2 nuclear reactor at 70% power under false pretenses, iust for profits, and as an unapproved risky experiment by subverting the NRC and Federal regulatory process. The true Root Cause (See Note 2) of the unprecedented tube-to-tube wear in Unit 3 has NOT been officially established, as required by NRC Confirmatory Letter Action 1 for restarting the defectively designed and degraded Unit 2. NRC has not even completed their review of Unit 2 Return to Service Reports, nor have they finished Special Unit 2 Tube Inspections, I
nor have they publicly reviewed SCE's Response to NRC's Requests for Additional Information (RAIs). Now, SCE wants the NRC to approve a new shady License Amendment, undermining public safety and do it without the involvement of Public Safety Experts/Attorneys and Citizens/Ratepayers.
After the review of the Mitsubishi Root Cause Evaluation and the Draft SCE License Amendment, it is crystal clear that the NRC needs to follow the example of their own enforcement at David Besse together with the lessons learned from Fukushima, when it comes to approving this new License Amendment for restarting San Onofre Unit 2's defectively designed and degraded replacement steam generators. In light of the unprecedented tube leakage at SONGS 3, the health and safety, along with the economic concerns and objections of 8.4 million Southern Californians' MUST OVERRIDE and PREVENT the restarting of Unit 2 at 70% or ANY power level until adequate repairs/replacement of the Unit 2 defectively designed and degraded replacement steam generators are complete. In addition, EIX Chairman Ted Craver, is strongly advised to clean house of the SONGS Senior Leadership Directors, Managers and Engineers, who are in violation of Federal Regulations for the flawed design of replacement steam generators, have been involved in systematic retaliation of workers reporting nuclear safety concerns and are responsible for the consistent low performance of SONGS for the last 5 years. In a Democratic Society, truth must prevail over profit motivations, misleading propaganda of electricity service disruption and temporary inconveniences to the public due to electric shortage based on phony data.
Notes:
1: Regulatory capture occurs when a regulatory agency, created to act in the public interest, instead advances the commercial or special concerns of interest groups that dominate the industry or sector it is charged with regulating. Regulatory capture is a form of ,government failure, as it can act as an encouragement for firms to produce negative externalities. The agencies are called "captured agencies".
- 2. Human performance errors resulting from the negative safety culture of production (profits) goals overriding public safety obligations.
WASHINGTON, April 3 - In an extraordinary admission, Southern California Edison said today that as part of the experimental plan to restart one of the crippled San Onofre nuclear reactors, the utility expects to have to shut it down and restart it four or five times in the next two years. Edison also confirmed that it will ask the Nuclear Regulatory Commission to approve an amendment to their operating license, which would mean a public hearing would be only held after the decision is made.
Friends of the Earth said the proposal is disgraceful -- another example of Edison putting profits before safety and treating legitimate public concerns, and those of independent nuclear engineers, about safety with disregard and contempt.
In a meeting at NRC headquarters, Edison announced that it wants the NRC to approve the request in time to restart the damaged unit 2 reactor by June 1 and wants the license to cover two years of operation. If those requests are granted, the NRC could approve the license before any public hearings.
NRC officials indicated their decision will rely on Edison's technical evidence as submitted. What it shows is that if the reactor is restarted, the already damaged steam tubes will vibrate, suffer further wear and potentially burst in 6 to 13 months -- well before the two year time frame Edison has proposed.
2
"Yet again Edison is putting profits before safety," said Kendra Ulrich, nuclear campaigner for Friends of the Earth. "To propose an experiment in which the damaged reactor is repeatedly turned on and off shows a disgraceful contempt for public safety. With more than 8 million people living near this reactor, the NRC must act to protect the public, reject this reckless request, and commit to a full public hearing process before they make any decision."
Friends of the Earth commissioned an in-depth technical analysis from a world-renowned nuclear engineer, John Large of Large & Associates in London. The analysis, released Tuesday and to be filed with the NRC, shows that Edison has yet to provide convincing evidence that it knows the full reasons or root cause of the severe wear damage in its steam generators. The problems remain unresolved and un-repaired, and the damage will continue if the NRC allows the reactor to be restarted. In fact, Edison's own experts hired by the utility to examine the problems disagree with one another as to the cause of the damage and the time left before a tube burst accident.
Letters from a coalition of grassroots organizations in Southern California, as well as national organizations, have been sent to Sen. Barbara Boxer (D-Calif.) and Rep. Henry Waxman (D-Calif.),
calling on them to use their influential positions on committees overseeing the NRC to demand a comprehensive license amendment process that includes all safety issues and the opportunity for full public hearing Analysis of Draft SCE License Amendment , Figures 1 through 14, and Tables 1 and 2, shows the damage in SONGS Unit 3 due to tube-to tube wear due to faulty steam generator design by SCE and MHI and other relevant information. , Figures 1 shows that small tube bundles and void fractions less than 98.5% in 200 Mitsubishi designed Steam Generators for the last 20 years are responsible for no tube-to-tube wear.
A review of the Mitsubishi Root Cause indicates that the normal tube bundle in Mitsubishi Steam Generators was purposely made taller in San Onofre Steam Generators by SCE to achieve 11 %
additional heat transfer to generate more heat and more profits in the pockets of EIX/SCE officers and Shareholders. SCE intentionally subverted the regulatory process. The fluid elastic instability or void fractions of 99.6% in Unit 3 were caused due to higher reactor coolant flows, high steam flows, high fluid velocities, high dry steam, narrow tube pitch to tube diameter, low tube-to-tube clearances, low frequency in-plane tubes, absence of positive in-plane vibration restraints, inadequate out-of-the-plane restraints design, and operation with low steam generator pressures and poor circulation ratios. These adverse effects destroyed SONGs Unit 3. By making the tube bundle taller without a 10CFR 50.90 License Amendment Process, Public Hearings and CPUC's Blessings, SCE increased the average length of 9727 tubes by 7 inches each to gain additional 7% heat transfer area equivalent to 700 new tubes to generate 120 more thermal megawatts per Steam Generator. Everybody is under the impression that SCE added only 377 tubes, but in reality, SCE added a total of effective 1,077 tubes including 700 tubes by making the tube bundle taller than contemporary successful operating Mitsubishi steam generators. SCE response to NRC RAI #13, states, "The RSGs have more tubes (9,727 versus 9,350) than the OSGs and a smaller value for the maximum number of plugged tubes (779 versus 2,000). RSG tubes have a larger average heated length (729.56 in. versus 680.64 in.) than the OSG tubes. These features result in larger values for the RSG for heat transfer area, tube bundle flow area, and tube bundle water volume. This is beneficial in the short and long term for SBLOCAs, which rely upon the steam generators for RCS heat removal. "
Unit 2 better supports and double the contact forces unproven theory is just a conjecture on the part of SCE/MHI based on hideous data and is contested by me based on the available plant data 3
evidence and review of John Large, MHI and AREVA Reports. Because of the unique flawed SCE replacement steam generator design dictated by profits over safety, dominant San Onofre negative safety culture and Mitsubishi's negligence and complacence, this analysis only applies to Unit 2 and Unit 3 replacement steam generators. No other Mitsubishi Steam Generators and NRC rules for Steam Generator Tube Integrity and NEI Steam Generator Management programs are affected or challenged at this time. The adverse effect of this change is 100% opposite of the benefit what SCE is telling the NRC and Public in RAI # 13 and as we witnessed in SONGS Unit 3. Here is why:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
SCE Response: No Significant Hazards Consideration. There is no significant increase in the probability or consequences of an accident previously evaluated.
Rebuttal: This adverse, unanalyzed and unapproved change involved more than a significant increase in the probability or consequences of a steam generator tube rupture leak due to 100% tube-to-tube wear in one tube in SONGS Unit 3, failure of 8 tubes at MSLB testing pressures and loss of
>35% wall thickness in more than 300 tubes. This event was caused due to the unexplained SCE/MHI design changes in the Mitsubishi Root Causes Analysis and misrepresented by SCE in response to NRC RAI #13. What, SCE did not tell the public that besides the leaking tube, there were 2 other additional tubes with a loss of wall thickness > 99%. If these tubes would have also leaked and resulted in SG over-pressurization causing lifting of safety relief valves, operator would not have been able to diagnose, manipulate, control and shutdown the reactor in a timely fashion and SONGS so called Engineered Safety Systems would not have been able to keep with the SG tube rupture LOCA. This would have caused a potential reactor meltdown and Southern Californians in the 10-mile Plume Pathway Zone would have experienced a Fukushima, or Mihama Unit 1 in their Backyards depending upon the direction of the wind and freeway traffic conditions. San Onofre, Interjurisdictional Planning Committee, Offsite Dose Assessment Committee, FEMA, NRC, State of California and Offsite Agencies tested Emergency Plans are totally inadequate to shelter and evacuate the affected transients, families, children, sick and disabled residents in such an event.
Therefore, the probability or consequences of these changes are/were more significant than analyzed by SCE in 10CFR 50.59 and seen by the destruction of SONGS Unit 3 than previously evaluated in the NRC Approved FSAR?
- 2. Does the proposed change create the possibility of a new or different kind of an accident from an accident previously evaluated?
SCE Response: No Significant Hazards Consideration. There is no possibility of a new or different kind of accident introduced because of this amendment Rebuttal: The requested SCE proposed License Amendment changes based on the false pretense of running defectively designed and degraded Unit 2 at 70% power to meet Peak Summer Month Power Loads significantly increase the possibility of an accident at normal steady state 70% power operations, during anticipated operational transients and a concurrent steam line break and consequential cascading tube ruptures due to fluid elasticity or high dry steam and jet impingement forces created as a result of 100% void fractions in the generator. Based on benchmarking of SONGS Unit 3, multiple tube leakages and/or ruptures are postulated due to 100% FEI in faulted and un-isolated (Assumed failure of MSIV to close) SG from a MSLB. Potential Collapse of floating AVB structure due to failure of low frequency retainer bars and high energy jet impingement can change 4
multiple tube leakages and/or ruptures into cascading tube ruptures. Current NRC rules consider main steam line break and steam generator tube ruptures as independent events. In addition, steam generator tube ruptures are considered to be a slow occurring event with plant operator able to detect the leak and take timely action to safely shut down reactor. This accident scenario applies and is unique to SONGS RSGs because of design flaws in the degraded AVB structure and is considered beyond design basis event. Consistent with Three Mile Island, Chernobyl, Mihama Unit 1, Fukushima and David-Besse Lessons Learnt, Operators cannot be relied on to control the plant and emphasis for accident controls and risk mitigations should be on defense-in-depth plant safety features.
SONGS plant does not have such beyond design basis accident defense-in-depth safety features.
Therefore, the proposed changes create: (1) The distinct possibility of a new or different kind of an accident than accidents previously evaluated, and (2) Involves more than significant reduction in the margin of safety previously evaluated in the FSAR and approved by NRC.
- 3. Does the proposed change involve a significant reduction in margin of safety?
SCE Response: No Significant Hazards Consideration. There is no change with this LAR that involves a significant reduction in margin of safety Rebuttal: The proposed License Amendment change operating Unit 2 at 70% power for 5 months involves more than a significant reduction in margin of safety. This is because of the high potential of beyond design basis steam generator tube ruptures caused by a main steam line break resulting in a potential nuclear meltdown beyond operator control, lack of SONGS Defense-in-Depth Features, single equipment and consequential equipment failures, communication problems, sonic booms, radiation/steam environment and access control in accordance with NRC Fukushima Task Force Lessons Learnt and offsite releases exceeding the limits than previously analyzed in the FSAR ? This condition is unique and applicable only to SCE/MHI defectively designed and degraded Unit 2 replacement steam generators. Please see Items 1 & 2 above for details.
5
FIGURE 8B' (U)STEAM PRESSURE STEADY STATE OPERATION CURVE 95-- - - m TubePlugging -__
Percentige - - - - -- - _ __ _ __
A76 00 *-1 AMR - - - - - - - - - - - - - - - - --
H8 a5a a i a i a i m am 0 10 20 30 40 50 60 70 80 90 100 REACTOR POWER (%)
Figure 1 - SONGS Unit 2 Steam Generators Steady Steady Ooeration Curve (June 2012)
(Received From Anonymous SONGS Insiders)
I
Figure 1A - SONGS Unit 3 Steam Generators Steady Stead-Y Operation Curve (June 2012)
(Received From Anonymous SONGS Insiders)
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Figure 2 - SONGS Steam Generator Design/Operations Features (Public Domain)
Reference Root Cause Evaluation: Unit 3 Steam Generator Tube Leak and Tube-to-Tube Wear Condition Report:
201836127, Revision 0, 5/7/2012, San Onofre Nuclear Generating Station (SONGS), Page 38 2
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I te a 0, 'ý M za,I Figure 3 - Westinghouse Operational Assessment. Attachment 6, Appendix B. Table 2-6 (www.songscommunity.com)
.. 7- 3 Table 2-7. Summary of ATHOS Results
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The color scheme on this figure isdirectly related to the %TW Depth (on the vertical axis), Iis lower %TW and I ishigher %TW.
Figure 5 - SONGS SG- TTW Depths Row And Column. Unit 3 E-088 (Public Domain)
Figure 4-10, AREVA Operational Assessment, Figure 4-10 (www.songscommunity.com) 4
000Lý_)0 0 00 Q0 00 0 0 'o 00 Q 0 Q 0 Q r 0 0 Q Q Q Q Q Q 00 0 0 00 0 0 0 0 0 0C 00 0 0 00000000 00 000 0000000000 Fi2ure 6 - SONGS Unit 3 SG E-088 Tube Damage - Figure Redrawn (Public Domain)
Solid Red Dots in circles show unprecedented tube-to-tube wear due to fluid elastic instability in the in-plane direction due to double the out-of plane velocities, high dry steam and high dynamic pressures caused by high reactor thermal power in the hot side heat flux, narrow tube to pitch diameter, excessive number of tall tubes, anti-FEI out-of-plane vibration bar structure, higher reactor coolant flows and operation at low steam pressures of 833 psi to generate more thermal heat out of the SG to generate more megawatts and more profits for EIX/SCE. This excessive heat in 4% region of the tube bundle with higher than normal wear(See Figures 10 through 18 below) was beyond the thermal design performance of the SG due to inadequate design and operational parameters discussed above. This adverse phenomena, studied only in experimental reactors, but not observed in operating steam generators prior to SONGS Unit 3, caused one tube Leak in SONGS Unit 3 SGS, failure of 8 tubes at main steam line break testing pressure (3X MSLB pressure), loss of wall thickness in hundreds of tubes > 35% NRC & SONGS Technical Plugging Limit, and additional 2 tubes lost wall thickness of 99%. If these 2 tubes would have leaked in conjunction with the one discovered leaking tube, SONGS Operators would not have been able to diagnose the accident in progress quickly enough, and could have resulted in a Meltdown of Unit 3. What you do not see in the above Red Dots, is that the hundreds of damaged tubes in Unit 2, in addition to losing strength due to wall thinning, have used up a significant fraction of their allowed fatigue life. Such damage cannot be detected by even the NRC Special Tube Inspections due to time, cost, unavailability of high technology probes and contactors, and/or impossible access within the tube bundle or radiation dose limitations. These tubes will be significantly susceptible to sudden ruptures without notice or early warnings during steady state normal operations at 70% power, Operational Transients (opening or closing of valves, scrams, loss of offsite power, moderate (anticipated) earthquakes, etc.) and under steam line break accidents at other reactors such as Crauss, Turkey Point, Robinson &
Mihama (1991 and 2004).
5
Deformled and Undeformed U-Bend, Row 106, Wjbe I I-d
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Figure 4-15, AREVA Operational Assessment (www.songscommunity.com) 6
Deformed and Undefomed U-Bend Row 106 Mode 2
-70 -50 -30 -10 10 30 50 XDistance Figure 8 - SONGS SG - Mode 2 Displacement Pattern (Public Domain)
Figure 4-18, AREVA Operational Assessment (www.songscommunity.com) 7
Deformed and Undeformed U-Bend, Row 106. Mode 3 Z0
-70 -50 -0 -10 0 50 7' X Dist ance Figure 9 - SONGS SG - Mode 3 Displacement Pattern (Public Domain)
Figure 4-19, AREVA Operational Assessment (www.songscommunity.com) 8
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Surfaco hoat flux q" as a function of oxcoss tomporaturo ATeo T - t Figure 10 - Heat Flux as a function of difference in temperature between tube surface and steam saturation http://en.wikipedia.org/wikilFilmnboiling Two-phase Steam water Fluid Flow: With reference to upflow in vertical channel (two-phase steam water flow in between the tubes in a nuclear steam generator), one can identify several flow regimes, or patterns, whose occurrence, for a given fluid, pressure and channel geometry, depends on the flow quality and flow rate. The situation of interest in nuclear reactors is nucleate boiling. Consider a vertical channel. At axial locations below the onset of nucleate boiling, the flow regime is single-phase liquid. As the fluid marches up the channel, more and more steam is generated because of the heat addition. As a result, the flow regime goes from bubbly flow (for relatively low values of the flow quality) to plug (intermediate quality) and annular (high quality). Eventually, the liquid film in contact with the wall dries out (this was the-cause of SONGS Unit 3 Fluid Elastic Instability). In the region beyond the point of dry-out, the flow regime is mist flow and finally, when all droplets have evaporated, single-phase vapor flow (as occurred in SONGS Unit 3 Destruction and as would occur in Unit 2 under MSLB conditions).
9
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void tracto (steam Whlen inc qualty) reases, ofthe U-flenu damping nLe Decreases 2'-
hý A 0 'ý I ýJ [I Figure 11 - Tube Dry Out Condition as a result of High Void fraction (Public Domain)
Mitsubishi Root Cause, Document No.L5-04GA588, Rev 0, Figure 3.3.3 MHI states in its Root Cause, "Causes of Type 1 Tube Wear (Tube-to-Tube wear): Most of the Type 1 wear (TTW) indications suggest that the wear is due to tube in-plane motion (vibration) with a displacement (amplitude) greater than the distance between the tubes in the adjacent rows, resulting in tube-to-tube contact. Tube in-plane motion can be caused by turbulence and fluid elastic instability (FEI). However, turbulence induced (random) vibration by itself is insufficient to produce displacements of this magnitude. Displacements as large as those associated with in-plane tube-to-tube contact can only be produced by fluid elastic vibration. Further, the contiguous grouping of the TTW tubes is another characteristic of fluid elastic instability, in order for large in-plane displacements to occur two conditions are necessary. First, the tube needs to be unrestrained in the in-plane direction and second the environment must be conducive to FEI (velocity, density, damping, etc.). MHI has analyzed whether random vibration was a precursor to the in-plane FEI that was observed in Unit 3. Two possible scenarios were considered. Scenario #1: In-plane FEI in Unit 3 had no precursor, Scenario #2: Wear from random vibration progresses to the point of loss of in-plane support, followed by the onset of in-plane FEI. The first scenario is more likely supported based on the investigation: (1) While the number of tubes with tube-to-AVB wear without in-plane TTW is greatest at the top of the tube bundle, the number of TTW tubes with tube-to-AVB wear is almost uniformly distributed along the different AVB intersections. If random vibration wear were a precursor for in-plane FEI TTW, then the pattern of AVB wear for TTW tubes should resemble the tube-to-AVB wear pattern (i.e. be concentrated at the top of the tube bundle). However, this is not observed for tubes with TTW, (2) While the tube-to-AVB wear depth for tubes without in-plane TTW is greatest at the top of the tube bundle, the tube-to-AVB wear depths for tubes with in-plane T-W is almost uniformly distributed along the AVB intersections. (See Fig. 3.5-2.) If random vibration wear were a precursor for in-plane FEI wear, then the AVB wear for the tubes with in-plane FEI would be greatest at the top of the U-bends. But for TTW tubes, the average wear depth is almost the same in all AVB support locations and there is no tendency to concentrate at the top of the tube bundle, and (3) The average 10% of AVB wear depth in Unit 2 and Unit 3 excluding TTW tubes is almost the same. (See Fig. 3.5-2.) Therefore, if random vibration were a precursor to in-plane FEI one would expect to see a similar number of tubes with tube-to-tube wear in the two RSG units. However, Unit 2 only has 2 tubes with TTW."
10
-OWN Fort Calhoun-1 SG USA, San Onofre-23 SG USA EDF SG France Ooel-1 SG Belgium, 260tonsr 583 tons 1 31 etons f 2 tons /
(triangular pitch tubing) (triangular pitch tubing) (square pitch tubing) (triangular pitch tubingi Moisture separator Tube sup PoT' structure at U-bent Heat transfer tAibe Tube support p lawe tbbe piattC Ii Figure 12 - Comparison of Mitisubishi Steam Generators for Export (Public Domain)
A MHI 2006 Brochure states, "Designs differ between individual customers because the specifications of replacement components are determined for each individual power plant. There is no standard design for a replacement SG because the specifications and plant requirements vary among customers. Steam generators (SG) have been replaced in PWR plants worldwide for more than 20 years. Since power utilities replacing their SGs are apt to want to increase their electric output by improving efficiency and equipment reliability, an increase of the heat transfer area is often required for the replacement SGs. It is technically challenging to enhance the performance and improve reliability even using the latest technology due to the strict restrictions on the interface dimensions of the replacement equipment. Despite this tough situation, MHI has been increasing its exports steadily, and has already supplied four units since 2003 (two to Belgium and two to the USA), and is in the process of designing or manufacturing 12 more units (two to Belgium, four to the USA, and six to France). These achievements show that MHI's advanced technology, quality, and process control capability are acquiring the reputation for high reliability among European and American utilities."
Based on an Anonymous SONGS insider reports, "SCE installed brand new replacement engineers designed to last for 40 years with the purpose of RSGs thermal performance to be as large as possible within the dimensional and other limitations." Westinghouse states, "For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. The recirculating fluid flow rate is relatively constant at all power levels. However, in the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.' MHI states, "The SCE/MHI AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger down-comer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable consequences and the AVB Design Team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 C.F.R. §50.59. Thus, one cannot say that use of a different code than FIT-Ill would have prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any feasible design changes arising from the use of a different code would have reduced the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG's design and MHI's previously successful designs would not have resulted in a design change that directly addressed in-plane FEI." I conclude that due to high steam flows and a normal than tall tube bundle (Average length of heated tubes increased from 680 inches to 730 inches along with the addition of 377 tubes specified by SCE), narrow tube pitch to tube diameter ratio, low tube-to-tube clearances, and steam generator operation at 833 psi caused void fractions of> 99.6% in the high region of wear, which caused FEI in Unit 3. Had SCE/MHI not avoided the NRC 50.90 License amendment process, reduced the void fractions < 98.5 %, read Dr. Pettigrew's research 2006 papers and NUREG-1841, performed careful and thorough analysis of cumulative effects of the design changes including the departures from the OSG's design, FEI in Unit 3 could have been averted and Ratepayers would not have been subject to a potential nuclear meltdown and could have saved Billions of Dollars.
LESSONS LEARNT: Haste makes waste, inadequate use of human performance tools and improperly conducted and poorly designed research experiments can cause (Potential/Actual) nuclear meltdowns and accidents as we saw in Chernobyl, Fukushima, Three Mile Island, Fukushima, Mihama and SONGS 3. Southern Californians want to avert a Potential meltdown of"defectively designed and degraded Unit 2 by not allowing NRC/SCE/MHI to restart it at any power level as an unapproved experiment without adequate repairs/replacement 11
+ ,4 Sigr amn--1 Fig. 1 Comparison of steam generators for export Designs differ between individual customers because the specifications of replacement components are determined for each individual power plant In-plane directon Out-of-plane direction Steam Steam Vent guide n e tRetaining bar Deckplate U-bend heatinsfer e ie Water-steam mixed flow Fig. 2 Tube support structure at U-bend Fig. 3 High-performance snnall-size separator Included in improvement designs, this is a vibration prevention mechanism for (single structure) htheat tubes transfer improvement design, this is a h nother performance small-size separator Fizure 13 - Comparison of Mitisubishi Steam Generators for Export (Public Domain)
A MHI 2006 Brochure states, "Major items of improvements in Replacement Steam Generators: There is no standard design for a replacement SG because the specifications and plant requirements vary among customers. Figure 15 compares the dimensions of recently exported SGs, in which widely varied specifications were applied. However, by applying the following latest advanced technologies to all SGs, improvements were made which cope with all past problems such as tube corrosion, vibration and wear, fatigue, and water hammer, and products which satisfy customers' advanced demands for heat transfer capability and moisture content are being supplied. (1) Tube material of high nickel alloy TT690 with excellent corrosion resistance. (2) Outstanding tube support plate design, tube expansion technology in tube sheets. (3) Tube support structure at U-bends with high support function. (4) High-performance moisture separators. Of these advanced demands for recent replacement SGs, items (3) and (4) deserve special attention.
2.2.1 Tube support structure at U-bends (3) The tube support structure at a U-bend is shown in Fig. 2. This is a unique design with reduced flow resistance while assuring a high support function by increasing the number of support points. Together with excellent assembly technology during manufacturing, high reliability against vibration and wear of heat transfer tubes is achieved. 2.2.2 High-performance moisture separators (4) MHI has developed a small, high-performance moisture separator by optimizing the geometry of the parts based on extensive field pressure tests (Fig. 3). As a result, replacement SGs corresponding to power up-rating and/or advanced moisture requirements can be designed."
12
S.INS 1 ABIIAI I y el.
VORIl Ex IND1) 177(E)
V IBR/A I ION 1-0.5 -I lRES1101,I) 1NDIAU11E1) (* 0 X7!BIRA lION0 20 0 ... L- -
0 1.0 2.0 3.0 PITCH FLOW VELOCITY Im/s)
Figure 14 - Vibrations amplitude as a function of flow pitch velocity for a flexible cylinder in a rigid cluster Violette R., Dr. Pettigrew M.J. & Dr. Mureithi N. W. state in a paper published in 2006 "In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability (See Figure 3 below) because of the high velocity of the two-phase mixture flow in the U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are asymmetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability."
13
Tahlim I - Qcan nnrdfra Q-Q. Doeignn nd )nersationnal Dlatsa U2/3 OSGs Other UZ/3OSGs before 2001 Plants after 2001 Design and Operational Parameters PVNGS Power Power which caused FEI, FIRV and MFE U2 RSGs U3 RSGs ANO U2 uprate uprate (1) Reactor Thermal Power, MWt 1724 1729 1995 1729 1705 (1A) Unit Electrical Generation, MWe 1183 1186 N/A N/A N/A (2) Number of Tubes 9727 9727 12,580 9350 9350 (3) Average Length of Heated Tubes, 729.56 729.56 766.8 680.64 680.64 inches mr (4) Heat Transfer Area, ft2 116,100 116,100 105,000 105,000 (5) Tube Wall Thickness, inches 0.043 0.043 0.043 0.048 0.048 (5A) Tube Diameter, inches 0.75 0.75 0.75 0.75 0.75 (5B) Tube Pitch, inches 1.0 1.0 0.87 1.0 1.0 (SC) Tube Array Triangular Triangular Triangula Triangular Triangular r/Square (5D) Tube Index 1.33-1.43 1.33 - 1.43 1.52-1.67 1.33-1.43 1.33-1.43 (5E) Tube to tube clearance, inches 0.25 0.25 0.25 0.25 (5F) Nominal Gap between tube and 0.002 0.002 0.003 N/A N/A AVB", cold, inches (5G) Nominal Gap between tube and 0 0 N/A N/A N/A AVB", Hot, inches (SH) Nominal Gap, Manufacturing N/A N/A N/A N/A N/A Dispersion, inches (51) Tube Wall Thickness/Tube 0.057 0.057 N/A 0.064 0.064 Diameter Ratio (5) Average Heated Tube Length/Tube 973 973 N/A 908 908 Diameter Ratio (6) Reactor Coolant Flow (at cold leg 75.76 79.79 5.6* 198,000 198,000 temperature),, Million lbs./hour (6A) Reactor Coolant Operating 598 598 618.9 599 611 Temperature (Thot), OF (6B) Reactor Coolant Operating 541.3 545.3 560.9 542 553 Temperature (Tcold), OF (7) RSG Operating Pressure (@100% 892 -942 833 1039 816 900 power), psia (7A) Steam Saturation Temperature,OF 531 523 549 520.5 532 (8) Feed-water Inlet Temperature ,OF 442 442 N/A 445 445 (9) Feedwater/Steam Flow, Million 7.7/7.6 7.7/7.6 8.95 7.18-7.63 7.41 lbs./hour (10) Steam Moisture Content, % <0.10 <0.10 <0.20 <0.20 (11) Steam Quality, % ? 87.6- 89.7 73.4 N/A N/A (11A) Void Fraction,% < 98.5% 9 9 .6% 98.5 96.1 96.1 (11B) Interstitial or Gap Fluid Velocity, -v22.00 28.30 17.91 22.90 22 feet/second (11C)Maximum Dynamic Pressure < 4000 4140 4220 N/A N/A (N/m2)
(12) Reactor Coolant Volume, ft3 2003 2003 N/A 1895 1895 (13) Circulation Ratio 3.3 3.3 4.3-5.7 N/A N/A (14) Delta Te= (Ts, 6A) - (TSAT, 7A), OF 67 75 68 78.5 79 Overall Heat Transfer Coefficient 1235 1280 N/A N/A N/A (Estimated) Btu/hr-ft2-OF Fluid Elastic Instability NO YES NO NO NO Flow-induced Random Vibration YES YES ? YES YES Mitsubishi Flowering Effect YES YES N/A N/A N/A 14
Table 2 - San Onofre Unit 2 RSGS Design and Operational Data U2 RSGs Design and Operational Parameters which caused FEI, @100% U3 RSGs U2 RSGs FIRV and MFE Power @70% Power @MSLB (1) Reactor Thermal Power, MWt 1729 1210 0 (1A) Unit Electrical Generation, MWe 1183 ? N/A (2) Number of Tubes 9727 9727 12,580 (3) Average Length of Heated Tubes, inches m 729.56 729.56 729.56 (4) Heat Transfer Area, ft2 116,100 116,100 116,100 (5) Tube Wall Thickness, inches 0.043 0.043 0.043 (5A) Tube Diameter, inches 0.75 0.75 0.75 (5B) Tube Pitch, inches 1.0 1.0 0.87 (5C) Tube Array Triangular Triangular Triangular/Square (5D) Tube Index 1.33-1.43 1.33 - 1.43 1.52-1.67 (5E) Tube to tube clearance, inches 0.25 0.25 Not Determined (5F) Nominal Gap between tube and AVB", cold, 0.002 0.002 0.003 inches (5G) Nominal Gap between tube and AVB", Hot, inches 0 0 0 (SH) Nominal Gap, Manufacturing Dispersion, inches N/A N/A N/A (51) Tube Wall Thickness/Tube Diameter Ratio 0.057 0.057 0.057 (5J) Average Heated Tube Length/Tube Diameter 973 973 973 Ratio (6) Reactor Coolant Flow (at cold leg temperature),, 79.8 78.2 78.2 Million lbs./hour (6A) Reactor Coolant Operating Temperature (Thot), 598 591 591 OF (6B) Reactor Coolant Operating Temperature (Tcold), 541 551 551 OF (7) RSG Operating Pressure (@100% power), psia 892 -942 946 ATM (8) Steam Operating Temperature (@ 0% power), °F 531 538 212 (8A) Steam Flow, Million lbs./hour 7.6 5.1 549 (8B) Feed-water Inlet Temperature , OF 442 407 N/A (9) Feedwater Flow, Million lbs./hour 7.6 5.1 33.8* to Environment in 3-5 Minutes (10) Steam Quality, % 90% 36% > 90%
(11) Void Fraction, % 98.5% 92.6% 100%
(11A) Maximum Gap Fluid Velocity, feet/second 25.1 12.6 > 50 (11B )Secondary fluid density, Ibm/cubic feet 7 12 <7 (12) Reactor Coolant Volume, ft3 2003 2003 2003 to Environment in 3-5 Minutes (13) Circulation Ratio 3.3 4.9 0 (13A) Down-comer Feed-water Flow, MIbs./hour 24.8 24.8 0 (14) Delta Te = (Ts, 6A) - (TSAT, 7A), OF 67 53 -400 Fluid Elastic Instability NO NO YES (Film Boiling)
Significant Radiation Flow-induced Random Vibration YES YES YES Mitsubishi Flowering Effect YES YES YES Flashing Feedwater Jet Impingement Forces on Tubes NO NO YES 15
--ort Ualrun.-1 W;UZA, ban UnotreLI St$. UMA, LU- 5U. t-rance, .
Uo6e1. Wielgium, 26Otons/ 583 tons / 316tons/ 268tons/
(triangular pitch tubing) (triangular pitch tubing) (square pitch tubing) (triangular pitch tubing)
SAN Onofre Replacement Generator Extremely Taller Tube Bundle (Effective Addition of 700 tubes) compared with other Mitsubishi Steam Generators specially built for Maximium Heat Transfer along with Inadequate Anti-Vibration Bars and Risky Floating Tube Bundle with Catastrophic Low Frequency Retainer Bars for Both Units 2 &
3 - Significantly Adverse Design Changes - Very high steam flows and low steam generator pressures caused fluid elastic instability in Unit 3 - Lower steam flows and higher steam generator pressures did not caused fluid elastic instability in Unit 2. Because of the above design flaws, Unit 2 at 70% power can experience potential cascading tube ruptures1 due to fluid elastic instability (100% void fractions, flashing feedwater jet impingment cutting holes in pressurized and plugged tubes, differential pressures, loose objects, low tube-to-tube clearances, and other unknown conditions) and flow-induced random vibrations caused by normal steady state 70% power anticipated operational transients ( closing and opening valves, loss of offsite power, etc,) and main steam line breaks.
1 Rapid tube-to-tube wear due to repeated and violent tube impacts caused by double in-plane fluid and steam velocities due to very high void fractions compared with normal out-of-plane fluid and steam velocities assumed in outdated out-of-plane ATHOS Models results certified by SCE and its Independent Experts. Sudden Tube ruptures due to unidentified, uninspected and undetected tube cracks caused by high cycle thermal fatigue due to high in-plane and out-of-plane random vibrations in Unit 2 during 22 months of adverse operations.
Figure 1- Comparison of Mitisubishi Steam Generators Tube Bundles