San Onofre: Too Close To Having Two Nuclear Accidents

Decommission San Onofre

SONGS Insider Secret: Southern California came close to having a level 3 nuclear incident (or worse, a nuclear disaster), at both SONGS Unit 2 and Unit 3.

A radioactive disaster was narrowly averted because of the off chance discovery that one of Unit 2 SG tubes had experienced a loss of 90 % percent of its wall thickness!  This was discovered before the tube had a chance to fail (like the eight tubes that failed in-situ testing in Unit 3) because Unit 2 happened to be shut down for a scheduled refueling outage, so its SG tubing were also inspected after the radioactive SG tubing leakage problems were discovered in Unit 3. Note: These SG tubes have a wall thickness (wear) plugging limit of 35%, so this single tube remained in use long after its safety limitation had been exceeded by almost double (55%)!   A single tube failure does not sound important, unless you realize it contains highly radioactive reactor core coolant that is under both very high pressure and temperature; then any leak is very bad (think massive radioactive atmospheric releases). A single tube rupture can also lead to a cascade of tube ruptures.

A Perfect Example of highly radioactive reactor core coolant: in 1991, leakage of about 55 tons of primary coolant occurred due to the failure of one SG tube in a steam generator built by Mitsubishi in the No. 2 pressurized water reactor at the Mihama nuclear power station in Japan.  At the same time, water pressure in the core had dropped drastically and the ECCS kicked in, flooding the reactor and shutting it down.  If the core had been left exposed, a meltdown -- an overheating of the fuel that can, if uncontrolled, lead to a large release of radioactivity -- could have occurred. In the following week an estimated 7 million Bq was released into the sea and an estimated 5 billion Bq of radioactive gas was released into the atmosphere.  This tube rupture caused the first International Nuclear Events Scale (INES) level 3 nuclear incident in Japan and ignited social concerns all over Japan because it shattered the nuclear industry’s myth of 100% safe reactors! SONGS MSLB  Analysis + 12-11-19, page 14

To date, there have been too many guesses, too many probabilities built around gamed safety margins and too many already damaged tubes (thousands of which have not been visually inspected) that any one or combination of, could cause a major nuclear accident even without something like a Main Steam Line Break (MSLB) or a big quake!

Given the documented UNSAFE condition of Unit 2, there is no excuse for SCE to even consider seeking a restart of Unit 2 except greed.  The NRC needs to make that crystal clear to SCE, otherwise they, like SCE, have failed the public’s trust! – The DAB Safety Team

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Nuclear Safety Is No Accident

The DAB Safety Team along with the support of an ever-growing number of SONGS Concerned Insiders and Whistleblowers have prepared this analysis, which is consistent with the conclusions presented in the publicly available reports provided earlier on this subject by:

·       Internationally Known Nuclear Consultant Arnie Gundersen and his team of Anonymous Industry insiders, who had lengthy careers in the design, fabrication, and operation of nuclear steam generators.

·       Professor Daniel Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.

·       Publicly available posted documentation by Dr. Joram Hopenfeld, a retired engineer from the Office of Nuclear Regulatory Research and NRC's Advisory Committee, and

·       David A. Lochbaum, Director of the Nuclear Safety Project for the Union of Concerned Scientists (UCS).

Background

Quote No 1: SCE is ultimately responsible for the work done by their vendors and contractors.  NRC Chairman Allison Macfarlane

Quote No 2: Both San Onofre units will remain shut down until repairs are made and we and the Nuclear Regulatory Commission are satisfied it is safe to operate. Pete Dietrich

Quote No 3:  Running the reactor at a 30 percent reduction in power may not fix the problems but rather make them worse or shift the damage to another part of the generators. It’s a real gamble to restart either unit without undertaking repairs or replacing the damaged equipment. Arnie Gundersen

Quote No 4: AREVA States, "The nominal distance between extrados and intrados locations of neighboring U-bends in the same plane ranges from 0.25 inches to 0.325 inches due to the tube indexing. There are 36 U-bends in Unit 2 SG E-088 and 34 in SG E-089 with a separation less than or equal to 0.050 inches. The U-bends with the smaller separation distances are much better candidates for wear by rubbing yet do not exhibit TTW.”

Based on SONGS Unit 3 experience, this behavior can change because of plugged/staked tubes due to shifting of localized steam-dry out regions in the hot leg U-Tube Bundle region of these U-bends with these extremely low clearances.  These extremely low clearances can cause very high steam velocities, which can then result in fluid elastic instability, EVEN during the proposed 150-day monitoring period EVEN at reduced power levels and EVEN  at higher steam pressures.

The DAB Safety Team Solve the Big Mystery Behind The Destruction of SONGS Unit 3 Replacement Steam Generators (RSG’s) And The Limited Damage to Unit 2 RSG’s:

Start With Some Basic Thermodynamics: At lower steam pressures (~833 psi), steam-water mixture has a saturation temperature of 523⁰F and more internal energy aka HEAT (1198.2 Btu/lb.) as compared with higher steam pressure (~942 psi), steam-water mixture has saturation temperature of 537.5⁰F and less internal energy aka HEAT (1194.7 Btu/lb.).  Therefore, you can generate more Megawatts out of the steam generator running it at lower steam pressure, if you supply more reactor thermal power (HEAT) to the steam generator tubes from the reactor.  Lower steam pressures combined with high reactor thermal power (HEAT), high steam velocities and narrow tube clearances also promotes bad things, like localized steam dry-out regions (vapor fraction >99%) in the hot leg side of the U-Tube bundle, which can cause fluid elastic instability, flow-induced random and severe vibrations, and excessive hydrodynamic pressures (aka Mitsubishi Flowering Effect).  These adverse effects can, in turn cause excessive tube-to-tube fretting wear leading to cascading tube leakages and or ruptures, increased tube support clearances and/or tube-to-AVB gaps, and even deformation of the anti-vibration bar structure.  Tube-tube/AVB wear and deformation (aka Mitsubishi Flowering Effect) of the floating anti-vibration bar structure without structural beams and lateral/mid-span supports causes redistribution of tube-to-tube/AVB gaps and clearances, when the steam generator reverts back from hot operating to cold conditions.  This redistribution of gaps can lead to differences in measurements during steam generator refueling inspections compared with the original manufacturing/design gaps in the cold condition and can lead to wrong projections and misleading conclusions regarding past and future cycles operating conditions.

Add Some Operational History: The original SONGS CE Model 3410 steam generators (OSG’s) were rated for 1705 MWt reactor thermal power.  Over the years of operation of the SONGS OSG’s, it became evident that the steam generator tubes, made predominantly of Alloy 600, were susceptible to primary water stress corrosion cracking (PWSSC). This corrosion mechanism resulted in tube degradation necessitating plugging large numbers of tubes after each tube inspection. In addition, the SONGS OSG’s design had shown to be susceptible to tube through-wall wear and severe corrosion of the tube supports. Continuing to operate with these highly degraded steam generators involved substantial economic risks from forced outages, extended refueling outages, as well as the direct costs of inspections and repairs.  SCE’s bid specification required that the stay cylinder feature of the original steam generators be eliminated to maximize the number of tubes that could be installed in the replacement steam generators and to mitigate past problems with tube wear at tube supports caused by relatively cool water and high flow velocities in the central part of the tube bundle. Mitsubishi employed broached trefoil tube support plates instead of the egg crate supports in the original design. In addition to providing for better control of tube to support plate gaps and easier assembly, the broached tube support plates were intended to address past problems with the egg crate supports by providing less line of contact and faster flow between the tubes and support plates, reducing the potential for deposit buildup and corrosion.  Mitsubishi selected a u-bend configuration for the upper part of the tube bundle instead of the square bend design of the original steam generators based on its experience that u-bends were easier to fabricate and support and were easier to inspect.

The original steam generators installed throughout the domestic fleet of pressurized water  reactors, including SONGS, experienced widespread corrosion of the tubes and tube support plates, stress corrosion cracking of the tubes, and wear at tube supports.  These problems led to the replacement of nearly all of the original steam generators, in most cases well before the end of their design lifetime.  For SONGS, the design of the replacement steam generators included a number of design changes to correct life limiting problems with the original steam generators, based in part on consideration of SONGS-specific and industry-wide operating experience. This included use of more corrosion resistant materials for the tubing and tube support plates to mitigate corrosion. The tubes in the new Replacement Steam Generators (RSGs) were fabricated with thermally treated Alloy 690, which has superior corrosion resistance.  The only drawback the new Alloy 690 has, is that it has a 10% less thermal conductivity/heat transfer rate compared with old tube alloy 600. 

Therefore, to achieve the thermal output of 1729 MWt from the new RSGs as approved by the NRC in 2001 SONGS Power Uprate Application, SCE engineers needed to install 11% more tubes in the new RSGs.  By removal of the stay cylinder from the OSGs, there was enough space to add only 4% (377) more tubes. Since there was no room to add the additional 7% tubes, SCE increased the length of each tube in the U-Tube Bundle (Outside the Industry NORM and is partially responsible for Mitsubishi Flowering effect) by more than 7 inches to obtain 1729 MWt out of the RSGs.  The Unit 2 RSGs were built with bigger tube-to-AVB Gaps and no in-plane protection from vibrations caused by potential fluid elastic instability conditions, because the OSGs did not experience that phenomena.

Based on information from anonymous SONGS operations personal, the root cause team of concerned insiders, a cursory review of plant records (plant procedures, system descriptions and plant daily briefing sheets), engineering calculations performed using SONGS procedures and review of NRC AIT Report, SCE Operators were running Unit 2 RSGs at higher steam pressures (~863-942 psi) and lower reactor thermal power (HEAT 1715-1725 MWt) for almost 22 months with no reported and detected abnormality (6 tubes with 28 to 90 percent wear of the tube wall thickness due to retainer bar vibrations).  Because of the short measured lengths of these flaws, only the 90 percent indication was in-situ pressure tested as part of condition monitoring. The affected tube was successfully pressurized to 5300 psi with no leakage (Southern Californians were saved from another nuclear accident). 

The Unit 3 SG manufacturing process used more accurate and tighter tolerances, which improved tube-to-AVB alignment such that tubes had more contact forces with AVB's and provided an effective “zero” tube-to-AVB gap under operating (hot) conditions. According to trusted SONGS operation insiders, SCE Engineers were convinced that Unit 3 RSGs were built better than Unit 2 RSGs in terms of providing in-plane protection due to better control of tube-to-AVB gap. Therefore, SCE Engineers decide to operate Unit 3 RSGs at lower steam pressures (~833 psi) to get more Megawatts out of the RSGs by supplying more reactor thermal power (HEAT ~1729 -1739 MWt) to the RSG’s tubes from the Unit 3 reactor.  This was unfortunately their biggest miscalculation.  What happened next is well known, because of the low steam pressures (~833 psi), combined with low tube clearances (SONGS design:  0.250 inches; some found as low as 0.050 inches) and no effective in-plane tube support protection, the high reactor power (HEAT) resulted in fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and localized steam dry-out regions (vapor fraction >99.6%) which resulted in the destruction of SONGS Unit 3 RSGs (1 tube leak, 8 tubes failures at MSLB conditions, 1800 tubes with tube-to-tube /anti-vibrations bars/support plates wear – this amount of damage is unprecedented in the history of the US Operating Nuclear “fleet”).   Fluid elastic instability and Mitsubishi Flowering Effects increased the tube-to-AVB gaps, decreased tube-to-tube clearances and created lower and insufficient contact forces in Unit 3, which were described inaccurately by the NRC and SCE as a MHI manufacturing defect , presumably due to political and/or financial reasons. 

After 22 months of operation the severity of wear in Unit 2 was determined to be similar to that experienced by Unit 3 after 11 months of operation.  The steam generator in Unit 2 had about 2600 wear indications at AVB’s compared to about 3400 wear indications in Unit 3. Without an effective in-plane support system, these high fluid velocities and localized steam dry-out regions resulted in very large or uncontrolled vibrations (amplitudes) of tubes in the in-plane direction and caused fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and increased tube-to-AVB gaps and created insufficient contact forces in Unit 3.  

Analysis Solves The Mystery: High steam pressures (~863-933 psi) and lower reactor thermal power (HEAT 1715-1725 MWt), coupled with low tube clearances and no in-plane tube support design accompanied with high steam velocities caused flow-induced random vibrations and excessive hydrodynamic pressures, which resulted in SONGS Unit 2 RSGs damage (high tube-to-anti-vibrations bars/support plates wear).  Since the steam pressures were higher, the void fractions were less than 98.5% and no fluid elastic instability occurred in Unit 2, which is consistent with the Westinghouse finding.  The NRC, AREVA, Westinghouse, MHI and SCE all missed this key operational observation in the NRC AIT Report, SCE Unit 3 Root Cause Evaluation and SCE Unit 2 restart Plan.  MHI indirectly alluded to this fact in their technical report but did not say it publicly for reasons unknown to the DAB Safety Team -- perhaps they were afraid of backlash from SCE and NRC.  At least one person working at SONGS discussed this fact about operational differences with other personnel between Units 2 and 3, but nobody listened to him.  His findings were intentionally or unintentionally ignored because everybody inside SCE was focused on blaming MHI in order to recover the insurance money and/or absolving themselves of all blame.  The DAB Safety Team is sure that MHI will pursue this fact during arbitration proceeding to absolve them of this blame and protect their reputation as a builder of high quality RSG’s.

Public Expectations: The public expects that the NRC complies with President Barack Obama, Senator Barbara Boxer and the NRC Chairman’s Open Government Initiative by using the Reactor Oversight Process, when it audits SCE’s Licensing Basis Documents, facility procedures/records, 10 CFR 50.59 Safety Evaluations, Unit 2 Restart Documents and issues/approves Safety Evaluation, License Amendment Applications and Inspection Reports, Responses to Confirmatory Action Letters and other enforcement violations, as appropriate.   The NRC must complete its mission of ensuring public safety with transparency and public involvement by issuing all documents, emails, telephone records along with holding open and trial-like public hearings without any time limitations from SCE, its vendors and/or contractors.

Questions which must be answered by NRC and SCE prior to any Unit 2 restart: The questions which NRC in its 95 page AIT Report the SCE and its vendors in their 1200 page San Onofre Nuclear Generating Station Unit 2 Return to Service Report have not answered completely, convincingly and unanimously are as follows:

(1)  Whether fluid elastic instability occurred, occurred for a limited time or did not occur at all in the SONGS Unit 2 Replacement Steam Generators (RSG’s)?

(2)  What was the contribution of fluid elastic instability and “Mitsubishi flowering effect” in increasing the Unit 3 Tube-to-AVB gaps, which were built better than Unit 2 and designed to provide an effective “zero” tube-to-AVB gap under operating (hot) conditions?

(3)  How were the SONGS RSG’s specified, designed and fabricated as comparable replacements on a like-for-like basis for the original steam generators in terms of fit, form and function with such numerous untested and unanalyzed design changes under the 10CFR50.59 rule?

(4)  Were these numerous untested and unanalyzed design changes responsible for damage to SONGS RSG’s?

(5)  Were SONGS RSG’s operating beyond their design basis or Industry NORMs and if so, how did that impact the degradation of RSGs?

 

(6)  Was this degradation the fault of SCE’s in-house design team, their Performance Specifications coupled with their numerous design changes and/or the MHI Fabrication/Testing Technology combined with Faulty Thermal-Hydraulic Computer Codes, which caused the unprecedented damage to SONGS RSG’s? 

(7)  Are the Unit 2 RSG’s qualified in the “As-designed and Degraded Condition” for a MSLB or other Anticipated Operational transients?

The DAB Safety Team and the Public expects that SCE and their three NEI Qualified, “US Nuclear Power Plant Designers”, Westinghouse, AREVA and MHI will revise their reports and arrive at concise and clear answers (meeting the NRC Quality Assurance requirements as stated by NRC Chairman Allison Macfarlane) to the puzzling public safety questions in the Unit 2 Return to Service Report.

SONGS Poor RSG’s Design: Due to unsubstantiated claims, complacency, challenges and rewards of innovative steam generators, time and financial pressures, both inexperienced Southern California Edison (SCE) and Mitsubishi Heavy Industries (MHI) Engineers did a very poor job of Industry Benchmarking and keeping up with the Academic Research Papers on the basic lessons of steam generator design (how to prevent the adverse effects of fluid elastic instability, flow-induced random vibrations, excessive hydrodynamic pressures and preventing localized steam dry-out regions observed in SONGS Units 2 and 3 RSG’s with high steam flows to produce more megawatts than the original steam generators (1729 MWt vs. 1705 Mwt)).  In addition, SCE made numerous untested and unanalyzed design changes under the pretense of 10 CFR 50.59 process (like for like change) and intentionally (Public Perception) avoided the NRC 50.90 License Amendment Process.  These factors contributed to the catastrophic failure of a 570 million dollars piece of equipment vital to the safety of Southern California.

SCE Restart Report for SONGS Unit 2: To address the tube leak and its causes, SCE assembled a technical team from MPR Associates, AREVA, Babcock & Wilcox Canada, PVNGS, EPRI, INPO, Westinghouse, Intertek APTECH and MHI, as well as experienced consultants including former NRC executives and a research scientist.  On October 3, 2012, SCE submitted a 1200 page report prepared by these experts (Under the closed Scrutiny, Guidance and Leadership of SCE Managers) to NRC Region IV. The SCE Root Cause Evaluation Report, Operational Assessments reports prepared by SCE, AREVA and Westinghouse, and MHI Technical reports conflict and contradict with each other on the causes and extent of degradation pertaining to the fluid elastic instability in SONGS Unit 2 Replacement Steam Generators (RSGs) and Tube-to-AVB gaps in both Unit 3 and Unit 2 RSG’s.  MHI truly states that specific causes that resulted in tubes being susceptible to fluid-elastic excitation are not yet completely known.  

PRESS RELEASE 

The DAB Safety Team: November 28, 2012

Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261

Copyright November 28, 2012 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and/or the DAB Safety Team’s Attorneys.

San Onofre's Restart Reports Fail BOTH NRC Safety Definition & Quality Assurance Standards

The DAB Safety Team has serious concerns about SCE’s Unit 2’s Restart Reports, because these reports do not meet the NRC Chairman’s Safety Definition nor do they satisfy the NRC’s 10CFR50, Appendix B Quality Assurance Standards.

Ultimate Responsibility: The top U.S. nuclear safety official, the Chairman of the NRC said earlier this month the operator of the idled San Onofre nuclear plant is ultimately responsible for ensuring the quality of equipment and work provided by vendors or its contractors.  Addressing nuclear industry executives in Atlanta, Nuclear Regulatory Commission Chairman Allison Macfarlane touched upon challenges at the idled San Onofre nuclear station and highlighted the responsibilities of the plant license holder. "This obligation extends to the licensees’ use of vendors and contractors," McFarlane said. "The licensee is ultimately responsible for the work done by their vendors and contractors to ensure they meet our quality assurance requirements." [Source: U-T San Diego November 7-2012]

Question Number 1: Why the massive tube damage at San Onofre aka SONGS?

More than 100 Replacement Steam Generators (RSGs) in the USA with Alloy 690 TT Tubes have been designed, fabricated and tested by Westinghouse, BWI and other vendors, including Fort Calhoun by MHI.  These steam generators have only had very few plugged tubes according to NUREG-1841 and Professor Dan Hirsch’s September 12, 2012 Report.  MHI has built more than 100 Steam Generators since 1970. Only Mihama Unit 2 SG built by MHI had a single tube rupture due to a displaced Anti Vibration Bar.  The question is, why did the SONGS Replacement Steam Generators suffer so much severe degradation in such a short time?  Is this the fault of SCE’s in-house design team, their Performance Specifications coupled with their numerous design changes and or the MHI Fabrication/Testing Technology combined with Faulty Thermal-Hydraulic Computer Codes?   The DAB Safety Team and the Public expected that SCE and their three NEI Qualified, “US Nuclear Power Plant Designers”, Westinghouse, AREVA and MHI would arrive at a concise and clear answer (Meeting the NRC Quality Assurance requirements as stated by NRC Chairman Allison Macfarlane) to this puzzling question in the Unit 2 Return to Service Report.

Observations On Number 1: The SCE Cause Evaluation Report, Operational Assessments reports prepared by SCE, AREVA and Westinghouse, and MHI Technical reports conflict and contradict with each other on the causes and extent of degradation pertaining to the fluid elastic instability in SONGS Unit 2 Steam Generator Replacement Generators (RSGs) and Tube-to-AVB gaps in both Unit 3 and Unit 2 RSG’s.  MHI further states that specific causes that resulted in tubes being susceptible to fluid-elastic excitation are not yet completely known.  Furthermore SCE has not plugged all the 2 tubes in one of the Unit 2 RSG’s nor have they removed the Retaining Bars (RB’s) as recommended by MHI in their latest NRC notification, issued after their preliminary report!

Operational Note On Number 1: Unit 2 was running at higher steam pressures than Unit 3 and lower thermal power than Unit 3.  That is why the void fractions were lower than 98.5% and no fluid elastic instability occurred in Unit 2.  AREVA, Westinghouse, MHI and SCE missed this key observation in the SCE Unit 2 restart Plan. At least one person working at SONGS spoke up about this fact but nobody listened to him and it was ignored because everybody in SCE was focused on blaming MHI to recover the insurance money and or absolving themselves of all blame.  The DAB Safety Team will explain the probable reasons other than the ECT results for Unit 3’s increased clearances between the anti-vibration bars and the tubes in their next Press Release.

Comments And Observations About Number 1: It is the opinion of the DAB Safety Team’s Expert Panel, former NRC Staff and SONGS Concerned Insiders that this Westinghouse Operational Assessment is full of holes based on incomplete inspection data, under-conservative computer modeling and is in effect, just “Smoke & Mirrors,” because:

(1)  SCE Engineers have either not provided, or they are withholding important information from Westinghouse because of  “The consequences of being Wrong, Terminated or Fired,”

(2)  Due to competing and proprietary interests between Westinghouse and MHI, Westinghouse Engineers do not have all the MHI Manufacturing Details and are just guessing in their Deterministic Operational Analysis of Unit 2, the second worst Degraded Replacement Steam Generators in the Operating US Nuclear Fleet,

(3)  Due to Time Pressure exerted by SCE, Westinghouse Staff did not have proper time for independent validation of all facts, documentation and data provided by SCE’s Engineers, in their original report.

(4)  Since nobody knows what really happened, all the Parties have a shared interest to “Operate Unit 2 at reduced power as a nuclear “RSG Tube Wear Test Lab”.

UCS Observations: The Union of Concerned Scientists (UCS) has serious concerns about Southern California Edison’s (SCE) restart plans for San Onofre Unit 2. In a 10/12/2012 letter submitted to the Nuclear Regulatory Commission (NRC), David Lochbaum, Director, Nuclear Safety Project, identified the following issues:

·      Unit 2 replacement steam generator 2SG89 has significantly more wear indications per number of supports than does [Unit 2] replacement steam generator 2SG88. Until the reason for this marked difference between the wear degradation for the Unit 2 replacement steam generators is understood, the operational assessment performed for future operation is suspect.

·      Since all four replacement steam generators came from the same manufacturer, were of the same design, made of the same materials, assembled using the same procedures, and operated under nearly identical conditions in twin reactors, the reason for this marked difference is unclear… [the] explanation is not well documented and therefore appears to be more convenient than factual.

·      The document states that the owner will “administratively limit Unit 2 to 70% reactor power prior to a mid-cycle” outage to inspect the replacement steam generators. What are the legal and safety consequences if the reactor power were to increase to 75%, 85% or 100% power, advertently or inadvertently?  The NRC has licensed San Onofre Unit 2 to operate at 100% power. What would legally prevent the owner from restarting Unit 2 and increasing its output to the NRC licensed limit? The NRC’s enforcement program includes sanctions when its regulations are violated, but nothing if promises are broken. If the NRC agrees that reactor operation at less than 100 percent power is warranted, it should enforce that reduction with an order or comparable legally enforceable document.  However even that will not necessarily prevent its occurrence.  Has NRC even considered that fact?

·       Table 8-1 of Enclosure 2 and its accompanying text attempt to explain how operating Unit 2 at 70% power will prevent the tube-to-tube wear (TTW) experienced on Unit 3 by comparing it to an anonymous reactor (called Plant A). Reliance on one suspect data point (Plant A) is hardly solid justification for operation at 70% power being acceptable.

·      There is no justification in this 80-plus-page document for an operating duration of 150 days.

·      There are no legal means compelling the plant’s owner to shut down Unit 2 after 150 days of operation at or above 15% power.

·      A temporary nitrogen-16 radiation detection system will be installed prior to the Unit 2 startup. However, there is no commitment to use it after startup, or to keep it in service should it stop functioning. The detection system is proposed as a defense-in-depth measure, but there is no assurance it will be operated.  Furthermore, it will NOT provide the necessary warning that tube rupture is eminent.  It will only indicate that it is already occurring.

·      Attachment 6 to Enclosure 2 has proprietary information redacted. Section 1.4 of Enclosure 2 states that the owner used AREVA, Westinghouse Electric Company LLC, and Intertie/APTECH to review the operational assessment. At least one of these companies manufactures replacement steam generators and would therefore be a competitor to Mitsubishi Heavy Industries (MHI), which made the replacement steam generators for San Onofre. If the owner did not withhold the proprietary information from MHI’s competitors, why withhold it at all? If SCE did withhold the proprietary information from these reviewers, what is the value of their independent, but limited, review?

Conclusions And A Final Question: The DAB Safety Team Agrees with NRC Chairwoman Allison Macfarlane that SCE is ultimately responsible for the work done by their vendors and contractors to ensure they meet our quality assurance requirements.  Based upon the review of all Restart Documents and all the issues identified by David Lochbaum, The DAB Safety Team’s Expert Panel along with their SONGS Concerned Insiders opinion that these reports are full of holes and based on incomplete inspection and or operational data, under-conservative computer modeling and represents Smoke & Mirrors which does not meet the NRC Chairwoman’s Safety Definition nor the standards outlined in the 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.”

The Trillion Dollar Question is now, how can the NRC’s Region IV in good faith, even consider having a November 30 Public Hearing, except to possibly give SCE a Billion Dollar Christmas present, by allowing them to restart their damaged Unit 2 without a 50.90 License Amendment Process by completely ignoring the safety of all those living in Southern California due to the potential of having a Trillion Dollar Eco-Disaster at San Onofre because of their already well documented massively damaged RSG tubes?  

PRESS RELEASE 

The DAB Safety Team: November 26, 2012

Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261

Copyright November 26, 2012 by The DAB Safety Team.  All rights reserved.  This material may not be published, broadcast or redistributed without crediting The DAB Safety Team.  The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and or the DAB Safety Team’s Attorneys.

Plug All Of San Onofre's Unsafe Tubes, Not Just Some

The DAB Safety Team Agrees With Newly Released MHI Data:

Plug All Of SONGS Unsafe Tubes, Not Just Some

The DAB Safety Team along with the support of an ever-growing number of SONGS Concerned Insiders and Whistleblowers, prepared the following analysis, which is consistent with the conclusions presented in the publicly available reports provided earlier on this subject by:

1.   Fairewinds Associates Internationally Known Nuclear Consultant Arnie Gundersen and his team of Anonymous Industry insiders, who have had lengthy careers in the design, fabrication, and operation of nuclear steam generators.

2.  Professor Daniel Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.

3.  Publicly available posted documentation by Dr. Joram Hopenfeld, a retired engineer from the Office of Nuclear Regulatory Research and NRC's Advisory Committee on Reactor Safeguards (ACRS) report issued in February 2001, which substantiated many of Dr. Hopenfeld's concerns,

4.  David A. Lochbaum, Director of the Nuclear Safety Project for the Union of Concerned Scientists (UCS).

  

MHI Part 21 (10/05/2012) - Steam Generator Tube Wear Adjacent To Retainer Bars:  The following information was received via email:  "Mitsubishi Heavy Industries, LTD (MHI) has identified steam generator tube wear for San Onofre Nuclear Generating Station.  "The Steam Generator tube wear adjacent to the retainer bars was identified as creating a potential safety hazard. The maximum wear depth is 90% of the tube thickness. The cause of the tube wear has been determined to be the retainer bars' random flow-induced vibration caused by the secondary fluid exiting the tube bundle. Since the retainer bar has a low natural frequency, the bar vibrates with large amplitudes. This type tube wear could have an adverse effect on the structural integrity of the tubes, which are part of the pressure boundary. The plugging of the tubes that are adjacent to the retainer bars was performed. MHI has recommended to the purchaser to remove the retainer bars that would have the possibility of vibration with large amplitude or to perform the plugging and stabilizing for the associated tubes."

SCE Unit 2 Restart Plan, Attachment 4, Page 9, Line 13, MHI States, "In order to ensure the structural integrity of the tubes after restarting the plant, all tubes which have a potential for losing their integrity during the next operating period should be plugged and thermal power output of the plant should be decreased.  Plugging for the Type 1 wear should include not only the tubes with the Type 1 (tube-to-tube) wear but also tubes which are susceptible to the Type 1 wear, for preventative reasons." Attachment 4, Page 82, Section 8.1.3, MHI states, “ Tubes with wear indications adjacent to the retainer bars should be plugged regardless of the wear depth. Furthermore, all tubes that have a possibility to come in contact with the retainer bars should be preventatively plugged.” SONGS Technical Specification states, “Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.”  General design criteria (GDC) 14, “Reactor Coolant Pressure Boundary (RCPB)” of Appendix A to United States Code of Federal Regulations 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” states “The RCPB shall have “an extremely low probability of abnormal leakage…and gross rupture.” 

Even at 70% power operations, if a steam line break outside containment were to occur in Unit 2, the depressurization of the steam generators with the failure of a main steam isolation valve to close would result in 100% void fraction in the entire U-Tube bundle.  This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI) and flow-induced random vibrations.  This adverse condition, in turn would result in hundreds of SG tube failures/ruptures due to tubes hitting each other because of extremely low tube clearances, NO in-plane support protection, and movement of retainer bars with large amplitudes due to low natural frequencies. With an undetermined amount of tube leaks/ruptures, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system.  The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five minutes from a broken steam line would EXCEED the SONGS NRC approved safety margins and result in a nuclear meltdown like Fukushima in Southern California.

Many steam generator tube ruptures and steam line break events have occurred in the last 30 years at nuclear power plants throughout the world (See DAB Safety Team’s SONGS MSLB Analysis).  In light of the Unit 3 Replacement Steam Generators (RSGs) unprecedented eight tube failures due to 99.6% steam voiding, narrow tube pitch to tube diameter ratio, low tube clearances and NO Designed "In-plane Fluid Elastic Instability support protection" and other tube ruptures/steam line break events, the DAB Safety Team agrees with MHI that all the Unit 2 Tubes would be susceptible to the Type 1 (tube-to-tube) failures/ruptures due to 100% steam voiding of the entire U-Tube Bundle in case of a Main Steam Line Break (MSLB).   Therefore, to meet the SONGS Technical Specifications and GDC 14 of Appendix A to 10 CFR Part 50 for a MSLB and prevent a nuclear accident and reactor meltdown in California from cascading tube ruptures, all Unit 2 RSG’s Tubes should be preventatively plugged before Unit 2 Restarts.  In other words, the Unit 2 RSG’s in the “As Designed and Degraded Configuration” cannot be OPERATED at any “Power Levels” due to the substantial risk of nuclear meltdown described above.

PRESS RELEASE 

The DAB Safety Team: November 21, 2012

Media Contact: Don Leichtling (619) 296-9928 or Ace Hoffman (760) 720-7261

Copyright November 21, 2012 by The DAB Safety Team. All rights reserved. This material may not be published, broadcast or redistributed without crediting the DAB Safety Team. The contents cannot be altered without the Written Permission of the DAB Safety Team Leader and or the DAB Safety Team’s Attorneys.