Media Alert January 28, 2013 from the DAB Safety Team

Media Alert 13-01-28

Allegation – NRC AIT Report Incomplete, Inconclusive, Inconsistent and Unacceptable

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

SONGS UNIT 3 RSG REAL ROOT CAUSE:  Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.

The DAB Safety Team has transmitted the following Allegation to “Office of the Chairman of the NRC” and “The United States Senate Committee on Environment and Public Work.”

It is the DAB Safety Team’s goal to help educate both the NRC and the Public by providing unbiased, logical and factual information in order to help assess the real dangers of any San Onofre Unit 2 restart.  According to Press Reports and San Onofre Insiders, Unit 2 permission for restart by the NRC is imminent yet the REAL Root Cause for the $1 Billion destruction of Units 2 and 3 RSGs (Including equipment cost and expenses) has not yet even been determined.  The Public does not know the status of SCE’s ongoing cause evaluations, SCE’s response to 32 NRR’s RAI’s and NRC’s Special San Onofre Inspections. We like to remind NRC San Onofre Special Panel, what NRC Chairman Macfarlane said during her recent Fukushima Trip, “Regulators may need to be ‘buffered’ from political winds, but they need to be fully subjected to the pressure of scientific and engineering truth and cannot be allowed to make decisions or order actions that are ‘independent’ of facts.” The NRC rush to a faulty judgment cannot be allowed to compromise Public Safety just to please SCE, as this conflicts with President Obama's Policy, the new NRC Chairman’s Standards and the advice of NRC retired Branch Chiefs who have also spoken out. 

A NRC Branch Chief gifted with MIT Intelligence, Intuition and a Sixth Sense, told an anonymous participant at an Industry Conference, “Sir, to resolve any complex technical problem and understand unclear regulations, you have to, ‘Read and reread in between the lines’, use, ‘Critical questioning and an investigative attitude’ and ‘Solid teamwork & alignment.”

SONGS UNIT 3 RSG ROOT CAUSE: It appears that Complacent SCE and Inexperienced MHI Engineers did not perform proper academic research and industry bench marking about the potential adverse consequences of the reduction of original CE steam generator pressures from 900 psi to say, 800 psi on fluid elastic instability and flow-induced vibrations.  These lower secondary steam operating pressure (800-833 psia) are the primary cause for shortening the life of SONGS Original Combustion Engineering Generators due to increased tube wear and plugging caused by flow-induced random vibrations and destruction of SONGS Unit 3 Replacement Steam Generators due to flow-induced random vibrations, Mitsubishi flowering effects and steam voids or steam dry-outs (AKA fluid elastic instability). In addition, SCE Engineers prepared defective 10 CFR 50.59 Evaluation and design specifications, which were not challenged by MHI, and adequately reviewed by NRC Region IV. MHI at the direction of SCE Engineers made numerous untested and unanalyzed design changes to the steam generators under the pretense of “like for like”, and Elmo Collins said, “The guts of the machinery look …. Different.”  

Therefore the SONGS UNIT 3 RSG REAL ROOT CAUSE:  Lack of “Critical Questioning & Investigative Attitude” by SCE, MHI and NRC Region IV and AIT Team.

·         Lessons Learnt: MHI, AIT, DAB Safety Team together with other World Nuclear and SG Experts / Manufacturers agree that, “Lower secondary steam operating pressures (800-833 psi) are severe and can easily cause SG flow-induced random vibrations and fluid elastic instability.”  At  lower secondary steam operating pressures (800-833 psi), the utility can generate more thermal megawatts out of the SG and hence add more power to the grid and thereby make more money for SCE/EIX Officers and shareholders. These lower secondary steam operating pressures (800-833 psi) were the primary cause for shortening the life of SONGS Original Combustion Engineering Steam Generators.  The real lesson learnt is that ALL parties must follow the NRC Chairman’s advice, pay attention to their work and always ensure that public safety is THE overriding obligation for the regulators, licensee, its vendors and contractors.  NRC “Reasonable Assurance” for the protection of adequate health and safety of the public from postulated radiological accidents cannot be compromised at any time during the design, operation, fabrication, testing, surveillance, maintenance and inspection of a nuclear power plant.  SCE and its offsite response organizations need to demonstrate; (1) Feasibility of an Operator Action during a postulated main steam line break with multiple Unit 2 steam generator tube ruptures, (2) They can effectively implement emergency plans and procedures with zero Drills/Exercise Performance indicator failures during a Fully Staffed NRC/FEMA evaluated exercise prior to any Unit 2 Restart.

·         Comments about the NRC Augmented Inspection Team San Onofre Report

NOTE: We highly recommend that NRC Augmented Inspection Team and NRC San Onofre Special Review Panel thoroughly review SONGS Unit 2 Return to Service MHI, AREVA, Westinghouse, DAB Safety Team and John Large Reports, then carefully examine the operational differences between Unit 2 and 3 and then update the NRC AIT report with a FACTUAL Root cause for FEI in Unit 3 and NO FEI in Unit 2. NRC San Onofre Special Review Panel also needs to review the SONGS Unit 2 Restart Reports (done by SCE, Westinghouse, AREVA and MHI), SCE Unit 3 Root Cause Evaluation, NRC AIT Report, ATHOS Modeling Results and Unit 2 Operational Data and then arrive: (1) At a unanimous, clear and concise conclusion whether FEI occurred in Unit 2 or not, and (2) Provide a GAP ANALYSIS (The scientific, technical and engineering reasons why all these reports are so different) prior the February 12, 2013 NRC Public Meeting.

The AIT inspection concluded that: (1) SCE was adequately pursuing the causes of the unexpected steam generator tube-to-tube degradation. In an effort to identify the causes, SCE retained a significant number of outside industry experts, consultants, and steam generator manufacturers, including Westinghouse and AREVA to perform thermal-hydraulic and flow induced vibration modeling and analysis; (2) The combination of unpredicted, adverse thermal hydraulic conditions and insufficient contact forces in the upper tube bundle caused a phenomenon called “fluid-elastic instability” which was a significant contributor to the tube to tube wear resulting in the tube leak. The team concluded that the differences in severity of the tube-to-tube wear between Unit 2 and Unit 3 may be related to changes to the manufacturing/fabrication of the tubes and other components which may have resulted in increased clearances between the anti-vibration bars and the tubes; (3) Due to modeling errors, the replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability. Unless changes are made to the operation or configuration of the steam generators, high fluid velocities and high void fractions in localized regions in the u-bend will continue to cause excessive and accelerated tube wear that could result in tube leakage and/or tube rupture; (4) The thermal hydraulic phenomena contributing to the fluid-elastic instability is present in both Unit 2 and 3 steam generators; (5) Based on the updated final safety analysis report description of the original steam generators, the steam generators’ major design changes were appropriately reviewed in accordance with the 10 CFR 50.59 requirements.

So based on a review of the AIT Report and World’s Experts, the three potential causes, which were significant contributors to the “fluid-elastic instability” in SONGS Unit 3 and the tube-to-tube wear resulting in the tube leak are as follows:

A. Insufficient contact tube-to-AVB forces and differences in manufacturing or fabrication of the tubes and other components between Units 2 & 3.

B. Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

C. Differences between Unit 2 and Unit 3’s Operational Factors.

A. Let us now examine that whether insufficient contact tube-to-AVB forces in the Unit 3 upper tube bundle caused “fluid-elastic instability” which was a significant contributor to the tube-to-tube wear resulting in the tube leak.

A.1- MHI states, “By design, U-bend support in the in-plane direction was not provided for the SONGS SG’s”. In the design stage, MHI considered that the tube U-bend support in the out-of-plane direction designed for “zero” tube-to-AVB gap in hot condition was sufficient to prevent the tube from becoming fluid-elastic unstable during operation based on the MHI experiences and contemporary practice. MHI postulated that a “zero” gap in the hot condition does not necessarily ensure that the support is active and that contact force between the tube and the AVB is required for the support to be considered active. The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation. This phenomenon is called by MHI, “tube bundle flowering” and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI. Observations Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were manufactured according to the design. It was confirmed that there were no significant gaps between the AVBs and tubes, which might have contributed to excessive tube vibration because the AVBs appear to be virtually in contact with tubes. MHI states, “The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. Because this high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors) the correlation between the void fraction (steam quality) and the number of

tubes with wear in a given void fraction region was investigated. From this investigation, a

strong correlation between the void fraction (steam quality) and the percentage of tubes with

the Type 1 (tube-to-tube) and Type 2 (tube-to-AVB) wear was identified. The correlation between flow velocity and the number of tubes with wear was also investigated. The results show that when the flow velocity is high, the percentage of tubes with wear increases, even though this correlation is not as strong as that between the void fraction (steam quality) and the percentage of tubes with wear.”

A.2 – AREVA states, “At 100% power, the thermal-hydraulic conditions in the U-bend region of the SONGS replacement steam generators exceeded the past successful operational envelope for U-bend nuclear steam generators based on presently available data. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location.”

A.3 – Westinghouse states, “Test data shows that the onset of in-plane (IP) vibration requires much higher velocities than the onset of out-of-plane (OP) fluid-elastic excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: There are several potential manufacturing considerations associated with review of the design drawings based on Westinghouse experience. The first two are related to increased proximity potential that is likely associated with the ECT evidence for proximity. Two others are associated with the AVB configuration and the additional orthogonal support structure that can interact with the first two during manufacturing. Another relates to AVB fabrication tolerances. These potential issues include: (1) The smaller nominal in-plane spacing between large radius U-bend tubes than comparable Westinghouse experience, (2) The much larger relative shrinkage of two sides (cold leg and hot leg) of each tube that can occur within the tubesheet drilling tolerances. Differences in axial shrinkage of tube legs can change the shape of the U-bends and reduce in-plane clearances between tubes from what was installed prior to hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs (designated as side narrow and side wide on the Design Anti-Vibration Bar Assembly Drawing that are attached to the AVB support structure on the sides of the tube bundle to become displaced from their intended positions during lower shell assembly rotation, (4) The potential for the 13 orthogonal bridge structure segments that are welded to the ends of AVB end cap extensions to produce reactions inside the bundle due to weld shrinkage and added weight during bundle rotation, and (5) Control of AVB fabrication tolerances sufficient to avoid undesirable interactions within the bundle. If AVBs are not flat with no twist in the unrestrained state they can tend to spread tube columns and introduce unexpected gaps greater than nominal inside the bundle away from the fixed weld spacing. The weight of the additional support structure after installation could accentuate any of the above potential issues. There is insufficient evidence to conclude that any of the listed potential issues are directly responsible for the unexpected tube wear, but these issues could all lead to unexpected tube/AVB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism. None were extensively treated in the SCE root cause evaluation.”

A.4 - John Large, Internationally Known Scientist and Chartered Nuclear Engineer from London says about the SONGS Unit 2 Replacement Steam Generators (RSGs) AVB Structure, “It impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state. The combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed. Phasing of AVB-TSP Wear -v- TTW: I reason that, overall, the tube wear process comprises two distinct phases: First, the AVB (and TSP) -to-tube contact points wear with the result that whatever level of effectiveness is in play declines. Then, with the U-bend free-span sections increased by loss of intermediate AVB restraint(s), the individual tubes in the U-bend region are rendered very susceptible to FEI induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that the wear mechanics comprises two phases, there are strong differences over the cause of the first phase comprising in-plane AVB wear: AREVA claim this is caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse) favor random perturbations in the fluid flow regime to be the tube motion excitation cause. Put simply: i) if AREVA is correct then reducing the reactor power to 70% will eliminate FEI, AVB effectiveness will cease to decline further and TTW will be arrested; however, to the contrary, ii) if Mitsubishi is right then, even at the 70% power level, the AVB restraint effectiveness will continue to decline thereby freeing up longer free-span tube sections that are more susceptible to TTW; or that iii) the assertion of neither party is wholly or partly correct. I find that the AVB assembly, which features strongly in the onset of TTW, is clearly designed to cope only with out-of-plane tube motion since there is little designed-in resistance to movement in the in-plane direction - because of this, it is just chance (a virtually random combination of manufacturing variations, expansion andpressurization, etc) that determines the in-plane effectiveness of the AVBs.”

A.5 - SCE claims, “The facts identified in this analysis indicate that even though the Unit 3 tube bundle components (tubes and AVBs) might have been fabricated and assembled better, the tube "to" AVB gaps built gaps might have been in fact larger in the Unit 3 RSGs as suggested by the ECT results. Based on this, it cannot be ruled out that the tube-to-AVB gaps are larger and more uniform in the Unit 3 RSGs than the Unit 2 RSG’s. This might have resulted in reduction of the tube-to-AVB contact force and consequently in multiple consecutive AVB supports being inactive. Inactive tube support might have resulted in "tube-to-tube" wear. 

A.6 – DAB Safety Team Conclusions: SONGS Unit 3 RSG’s were operating outside SONGS Technical Specification Limits for Reactor Thermal Power and Current Licensing Basis for Design Basis Accident Conditions. Let us summarize what these experts are saying: (1) AREVA is saying, “After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large. Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location”, (2) Westinghouse is saying, “A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion”, (3) MHI is saying, “High steam flows and cross-flow velocities combined with narrow tube pitch-to-diameter ratio caused elastic deformation of the U-tube bundle from the beginning of the Unit 3 cycle, which initiated the process of tube-to-AVB wear and insufficient contact forces between tubes and AVBs. Tube bundle distortion is considered a major contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the Unit 3 SG’s. After 11 months of wear, contact forces were virtually eliminated between the tube and AVBs in the areas of highest wear of Unit 3 tubes as confirmed by ECT and visual inspections. According to MHI Technical Document, the RSG Anti-bar Vibration Structures were only designed for out-of plane vibrations and not in-plane vibrations”, (4) John Large is saying, “Its impossible to reliably predict the effectiveness of the many thousands of AVB contact points for when the tube bundle is in a hot, pressurized operational state.” In essence, all the three NEI qualified, “U.S. Nuclear Plant Designers” are saying: (a) Firstly, these AVB structures are not designed to prevent in-plane vibrations, (b) Secondly, once fluid elastic instability develops, the AVB is not strong enough to prevent large in-plane amplitude of tubes,  (c) Thirdly, excessive fluid-induced random vibrations and fluid hydrodynamic pressures (Mitsubishi Flowering Effect) cause the loss of tube-AVB contact forces and increase the tube-to-AVB gaps and the onset of fluid elastic instability develops, and (d) Lastly, tube-to-AVB gaps and contact forces are irrelevant to prevent fluid elastic instability from progressing and causing tube-to-tube and tube-to-AVB wear. Therefore, based on a review of MHI, AREVA and Westinghouse excerpts shown above, it is concluded, that FEI, flow-induced random vibrations and MHI Flowering effect redistributed the tube-to-AVB gaps in Unit 3 RSG’s as measured by SCE. Westinghouse even goes this far to state, “None of the MHI manufacturing issues were extensively treated in the SCE root cause evaluation.” Hence, It is concluded that the NRC and SCE claims that insufficient contact forces in Unit 3 Tube-to-AVB Gaps ALONE caused tube “to” tube wear are misleading, erroneous and designed to put the blame on MHI for purposes of making SCE look good in the public’s eyes and for collecting insurance money from MHI’s so-called “manufacturing defects”.

B. Let us now examine the effects of modeling errors, that the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.

B.1 – NRC AIT Report states, “The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result, it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Therefore this analysis cannot account for these mechanical factors and differences which could very likely also be contributing to the tube degradation.”

B.2 – Ivan Cotton states, “Fluid elastic instability is one of the most damaging types of instabilities encountered in heat exchangers and steam generators and can impose a severe economic penalty on the power and chemical industries. At present our understanding of the mechanisms leading to fluid-elastic instability is very limited and more experiments are needed to more fully delineate the conditions for the onset of fluid-elastic instability.” Such experimentation should only be done in a sealed lab, NOT our environment with the lives of eight million local residents at stake in the outcome!

B.3 – Ishihara, Kunihiko and Kitayama state, “Tube vibrations become large as tube thickness/diameter ratio (T/D) increases and tube length/diameter ratio (L/D) decreases, and the tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.” In the case of SanO’s replacement SGs, the tubes, especially in the U-bend region, were too close together, poorly restrained, poorly damped, along with too much heat flux and an inappropriate pressure-to-flow ratio, along with other causes which resulted in FEI and FIV.

B.4 – Fairewind states, “Realistically, the 3-D steam analysis is not accurate enough to apply to such important safety related determinations. To make such mathematical risk 3-D analysis, a very large margin of error must be applied, and that has not been done. For example, if the 3-D steam analysis determines that plugging 100 tubes is a solution, then plugging ten times that number might be the appropriate solution due to the mathematical errors in the 3-D analysis being applied by Edison and Mitsubishi. Edison has taken many of its mathematical/computer models as “gospel” rather than accepting their wide margins of error, which directly affect safety, which is precisely what got them in trouble in the first place. It should have been obvious to Edison that MHI FIT-III has not been benchmarked, and had not been previously used in licensing procedures showing that the use of FIT-III might have an adverse effect on the FSAR safety analysis thus necessitating the entire license amendment review and public hearing process. As noted by the AIT, Edison approved the use of FIT-III code even though the code was not benchmarked, nor identified as acceptable in the FSAR. Consequently, Edison operated San Onofre without knowing the uncertainties in the Replacement Steam Generators’ performance characteristics. Predicted liquid levels, pressure drops, vibrations, and temperatures at both Units 2 and 3 were all subject to unknown uncertainties during both normal and abnormal operations. In my opinion, by approving the use of an un-benchmarked and untested design tool like FIT-III, Edison did not did not meet the requirements expected from a nuclear licensee.

Use of an un-benchmarked computer code that is not included in the FSAR protocol demands a formal FSAR license amendment process including the requisite public hearings.”

B.5 – Mitra, V.K. Dhir, I. Catton state, “Flow induced vibrations in heat exchanger tubes have led to numerous accidents and economic losses in the past. Efforts have been made to systematically study the cause of these vibrations and develop remedial design criteria for their avoidance. Instability was clearly seen in single phase and two-phase flow and the critical flow velocity was found to be proportional to tube mass. It is also found that nucleate boiling on the tube surface is also found to have a stabilizing effect on fluid-elastic instability.”

B.6 - John Large, says “Factual Issues v) & vi) – SONGS SG Comparison to Other Operating SGs: I identify a number of issues with the representation of Figures 4-3 and 5-1 of the AREVA Tube-to-Tube Report, including: i) it is not exactly clear which properties are being represented on the spider diagram for comparison with the other operational SGs; even so ii) since it is most unlikely that AREVA has undertaken a comprehensive (ATHOS) simulation of each of the five nominated SGs, the comparisons drawn are likely to be between aggregate or bulk flows within the entire tube bundle of each SG; iii) as acknowledged by AREVA, the SONGS RSGs are dominated by in-plane flow regimes whereas all other SGs are characterized by out-of-plane flow regimes; and iv) none of the comparative SGs has been identified.  In other words, unless the spider diagrams of Figure 4-3 and 5-1 somehow, and I cannot reason how, are making a direct comparison of the complex two-phase fluid cross-flow situation in the SONGS and other five comparative plant steam generators, then these figures only provide the bases of a somewhat meaningless comparisons.”

B.7 – SCE states that SONGS Unit 3 Damage (FEI) was caused due to outdated MHI Thermal-Hydraulic Computer Models. According to NRC AIT Report, SONGS did not specify the value of FEI in its Design and Performance Specifications SO23-617-1. Academic Researchers have discussed and warned about the adverse effects of fluid elastic instability (tube-to-tube wear) in nuclear steam generators since the 1970’s. Westinghouse and Combustion Engineering (CE) have designed CE replacement steam generators (RSGs) to prevent the adverse effects of fluid elastic instability since 2000’s (e.g., PVNGS).

B.8 – The NRC AIT Report dated November 9, 2012 states, “the FIT-III thermal-hydraulic model was still in-progress at the time of the inspection and no final conclusions were reached for the cause of the non-conservative flow velocities, which were used as inputs in the tube vibration analysis and resulted in non-conservative stability ratios. Since the licensee had not completed the cause evaluation for this unresolved item, the inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred. The inspectors were informed that Mitsubishi was performing an evaluation of the potential factors that contributed to the low flow velocities in FIT-III relative to the velocities calculated by the ATHOS model developed after the tube leak event in Unit 3. This evaluation was included in Document SO23-617-1-M1530, Revision 1, which also intended to demonstrate the validity of FIT-III results for the original tube vibration analysis. This evaluation was still being finalized and not yet approved by Edison. The licensee and Mitsubishi continued to evaluate this unresolved item and no final conclusions were reached at the time of the inspection. The NRC is continuing to perform independent reviews of existing information, and will conduct additional reviews as new information becomes available.” In another related finding, NRC inspectors stated, “SCE Engineers did not meet Procedure SO123-XXIV-37.8.26 requirements to ensure the design of the retainer bar was adequate with respect to the certified design specification. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur, as a result of contact with the retainer bars due to flow-induced vibration. The inspectors determined that the requirements for flow-induced vibration in the certified design specification, along with the expectations in Procedure SO123-XXIV-37.8.26, provided sufficient information to reasonably foresee the inadequate design of the retainer bars during the review and approval of design Calculations SO23-617-1-C749 and SO23-617-1-C157, including the associated design drawings provided by Mitsubishi.”

B.9 – Arnie Gundersen states, “Not only is Mitsubishi unfamiliar with the tightly packed CE design, but Edison’s engineers added so many untested variables to the new fabrication that this new design had a significantly increased risk of failure. As a result of the very tight pitch to diameter ratios used in the original CE steam generators, Mitsubishi fabricated a broached plate design that allows almost no water to reach the top of the steam generator.
The maximum quality of the water/steam mixture at the top of the steam generator in the U-Bend region should be approximately 40 to 50 percent, i.e. half water and half steam. With the Mitsubishi design the top of the U-tubes are almost dry in some regions. Without liquid in the mixture, there is no damping against vibration, and therefore a severe fluid-elastic instability developed.

Because of the Edison/Mitsubishi steam generator changes, the top of the new steam generator is starved for water therefore making tube vibration inevitable. Furthermore, the problem appears to be exacerbated by Mitsubishi’s three-dimensional thermal-hydraulic analysis determining how the steam and water mix at the top of the tubes that has been benchmarked against the Westinghouse design but not the original CE design.

The real problem in the replacement steam generators at San Onofre is that too much steam and too little water is causing the tubes to vibrate violently in the U-bend region. The tubes are quickly wearing themselves thin enough to completely fail pressure testing. Even if the new tubes are actively not leaking or have not ruptured, the tubes in the Mitsubishi fabrication are at risk of bursting in a main steam line accident scenario and spewing radiation into the air.”

B.10 – Comment on Limitations of ATHOS thermal-hydraulic Models: The ATHOS thermal-hydraulic model predicts bulk fluid behavior based on first principals and empirical correlations and as a result, it is not able to evaluate mechanical, fabrication, or structural material differences or other phenomena that may be unique to each steam generator. Furthermore, the combination of the omission of the in-plane AVB restraints, the unique in-plane activity levels of the SONGS RSGs, together with the very demanding interpretation of the remote probe data from the cold and depressurized tube inspection, render forecasting the wear of the tubes and many thousands of restraint components when in hot and pressurized service very challenging indeed. ATHOS thermal-hydraulic Models used for 70% power have not been benchmarked, and tested against SONGS Unit 2 RSG degraded tube bundles performance for several cycles of depressurized/pressurized operation.  Hence, ATHOS analyses cannot accurately predict the behavior of pressurized degraded SG tube bundles and their interaction with their anti-vibration bar support structure, which could very likely contribute to unknown amounts of tube-to-tube wear and/or AVB degradation in Unit 2, at 70% or 100% power during a main steam line break accident resulting in a cascade of multiple steam generator tube ruptures. 

B.11 – Conclusions on Modeling Errors: The NRC AIT Report dated November 9, 2012 states, “the FIT-III thermal-hydraulic model was still in-progress at the time of the inspection and no final conclusions were reached for the cause of the non-conservative flow velocities, which were used as inputs in the tube vibration analysis and resulted in non-conservative stability ratios. Since the licensee had not completed the cause evaluation for this unresolved item, the inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred. SCE and MHI are both negligent because they did a very poor job of Industry and Academic Research benchmarking regarding the applicability of thermal-hydraulic computer models during the redesign of San Onofre’s original CE SGs. SCE is negligent because they did not check the results of MHI’s outdated Thermal-Hydraulic Computer Models to meet their specification and procedure requirements. This does not meet the NRC Chairman’s Standards. Therefore, it is concluded that SCE claims as stated above are not factual. SCE engineers did not check the work of MHI with a critical and questioning attitude and did not meet the San Onofre Design Procedures, 10CFR50, Appendix B, Quality assurance Standards and or NRC Regulations.  NRC AIT Team jumped the boat by putting all the blame on MHI in July 2012 and now they are retreating in November 2012 by stating with a sunken face, “The inspectors were not able to make a final determination of whether a performance deficiency or violation of NRC requirements occurred.”

C. Let us now examine the other differences between Unit 2 and Unit 3’s Operational Factors, which were significant contributors to the “fluid-elastic instability” in San Onofre Unit 3 and the tube-to-tube wear resulting in the tube leak.

C.1 – Adverse Design/Operational Factors responsible for Fluid Elastic Instability: Low steam generator pressures (SONGS RSGs range 800-850 psi, the primary cause of the onset of severe vibrations) allow the onset of FEI, whereby U-tube bundle tubes start vibrating with very large amplitudes in the in-plane directions. Extremely hot and vibrating tubes need a little amount of water (aka damping, 1.5% water, steam-water mixture vapor Fraction 99.5%). Without the water, the extremely hot and vibrating tubes cannot dissipate their energy. In effect, one unstable tube drives its neighbor to instability through repeated violent impact events which causes tube leakage, tube failures at MSLB test conditions and/or unprecedented tube-tube wear, Tube-to-AVB/Tube Support Plates wear, as we saw in San Onofre Unit 3. So in review, due to narrow tube pitch to tube diameter, tube natural frequency, low tube clearances, in certain portions of the RSGs U-tubes bundle, fluid velocities exceed the critical velocities due to extremely high steam flows (100% power conditions). These high fluid velocities cause U-tubes to vibrate with very large amplitudes in the in-plane direction and literally hit other tubes with repeated and violent impacts. Due to lower secondary steam operating pressures (required to generate more heat, electricity and profits) and excessive pressure drops due to high flows and velocities, steam saturation temperature drops. This lowering of steam saturation temperature combined with high heat flux in the hot leg side of the U-tube bundle causes steam dry-outs to form (Vapor fraction >99%), known as “NO Effective Thin Tube Film Damping.” Thin film damping refers to the tendency of the steam inside the generators to create a thin film of water between the RSG tubes and the support structures and each other. That film is enough to help keep the tubes from vibrating with large amplitudes, hitting other tubes violently, and to protect the Anti-Vibration Bar support structures and maintain the tube-to-AVB gaps and contact forces. These adverse conditions in Unit 2 at 70% power operation (RTP) with the present defective design and degraded RSGs, known as fluid elastic instability (Tube-to-Tube Wear, or TTW) can lead to rapid U-tube failure from fatigue or tube-to-tube wear in Unit 2 due to a main steam line break as seen in Unit 3’s RSG’s. In summary, FEI is a phenomenon where due to San Onofre RSGs design intended for high steam flows causes the tubes to vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. The lack of Nucleate boiling on the tube surface or absence of water is found to have a destabilizing effect on fluid-elastic stability.

C.2 – Unit 2 FEI Conflicting Operational Data

·         NRC AIT Report SG Secondary U2/3 Pressure Range 833 - 942 psi

·         SCE RCE SG Secondary U2/3 Pressure - 833 psi

·         RCE Team Anonymous Member - Unit 2 SG Secondary Pressure 863 psi

·         SONGS SG System Description Unit 2 SG Pressure Range 892 - 942 psi

·         Westinghouse OA SG Secondary U2/3 Pressure ~ 838  psi, Void Fraction 99.55%

·         SCE Enclosure 2, MHI ATHOS results - U2/3 Void Fraction 99.6%

·         SCE Enclosure 2, Independent Expert results - ATHOS U2/3 Void Fraction 99.4%

·         DAB Safety Team SG Secondary U2 Pressure 863 -942 psi, Void Fraction 96-98%

·         SONGS Plant Daily Briefing Unit 3 Electrical Generation – 1186 MWe

·         SONGS Plant Daily Briefing Unit 2 Electrical Generation – 1183 MWe

C.3 – Unit 2 FEI Conclusions

C.3.1 - NRC AIT Report - Operational Differences between U2/3 - The result of the independent NRC thermal-hydraulic analysis indicated that differences in the actual operation between units and/or individual steam generators had an insignificant impact on the results and in fact, the team did not identify any changes in steam velocities or void fractions that could attribute to the differences in tube wear between the units or steam generators.

C.3.2 - SCE Unit 2 Restart Report Enclosure 2 Conclusions - Because of the similarities in design between the Unit 2 and 3 RSGs, it was concluded that FEI in the in-plane direction was also the cause of the TTW in Unit 2.

C.3.3 - SCE U2 FEI SONGS RCE Team Anonymous Member Conclusions – FEI did not occur in Unit 2

C.3.4 - Westinghouse OA Conclusions: (a) An evaluation of the tube-to-tube wear reported in two tubes in SG 2E089 showed that, most likely, the wear did not result from in-plane vibration of the tubes since all available eddy current data clearly support the analytical results that in-plane vibration could not have occurred in these tubes, and (b) Operational data – Westinghouse ATHOS Model shows no operational differences in Units 2 & 3 (void fraction ~99.6%) and then Westinghouse says in (a) above that FEI did not occur in Unit 2.  Westinghouse is contradicting its own statement.

C.3.5 - AREVA OA Conclusions - Based on the extremely comprehensive evaluation of both Units, supplemented by thermal hydraulic and FIV analysis, assuming, a priori, that TTW via in-plane fluid-elastic instability cannot develop in Unit 2 would be inappropriate.

C.3.6  - John Large States, “I note here that there are three clear conflicts of findings between the OAs: From AREVA that AVB-to-tube and TTW result from in-plane FEI, contrasted to Westinghouse that there is no in-plane FEI but most probably it was out-of-plane FEI, and from MHI that certain AVB-to-tube wear results in the absence of in-plane FEI from just turbulent flow. My opinion is that such conflicting disagreement over the cause of TTW reflects poorly on the depth of understanding of the crucially important FEI issue by each of these SCE consultants and the designer/manufacturer of the RSGs.”

C.3.7 - DAB Safety Team Conclusions – Due to higher SG pressure (Range 863 – 942 psi) and lower thermal megawatts as compared to Unit 3, FEI did not occur in Unit 2.   This is consistent with the position of RCE Team Anonymous Member.  The NRC AIT Report, SCE, Westinghouse, MHI, SCE chosen “Independent” Experts and AREVA conclusions on Unit 2 FEI are inconsistent, confusing and inconclusive. The NRC AIT Report, SCE and AREVA conclusions on Unit 2 FEI are unacceptable.

C.3.8 - The NRC San Onofre Special Review Panel should direct other branches within the NRC (NRC-RES and/or the ACRS) to review the above data without any prior “turf” bias and present their findings to the public for review and comment prior to any restart decision being made by the NRC.

Dangers of SAN ONOFRE Unit 2 Restart

1 - Chairman Allison Macfarlane said Unit 2 would not be permitted to restart unless the NRC has reasonable assurance it can be operated safely. Let us examine that scenario below and determine whether NRC can have that reasonable assurance or not:

2 - Let us examine what John Large says, “There is no account of the changes that have been made in the evaluation of the tube structural and leakage integrity, that is from the stage of predicting those tubes at risk of TTW and other forms of wear, the tube thinning wear rates, through to the nature of the tube failure being unique to the type and extent of the wear pattern and tube thinning; and the methods of deducing, mainly by unproven inference, from the probe inspection results particularly to determine the in-plane AVB effectiveness, includes unacceptably large elements of test and experimentation that are inconsistent with the analyses and descriptions of the FSAR. I provide a number of explicit examples where I consider that the circumstances and risks accompanying the proposed restart of Unit 2 will result in unacceptable levels of test and experiment.” What he is saying is that these degraded tube bundles cannot prevent multiple tube ruptures from fluid elastic instability as we saw by the failure of 8 tubes in Unit 3 RSGs under Main Steam Line Break (MSLB) test conditions.

3 - Let us examine what Arnie Gundersen says, “Eight replacement steam generator tubes failed their pressure tests in 2012 and more than 1,000 others have been plugged. Therefore, a review of the evidence makes it clear that the San Onofre Replacement Steam Generator tube damage discovered in 2012 was so severe and extensive that both reactors have been operating in violation of their NRC FSAR license design basis as defined in their Technical Specifications.  The Main Steam Line Break with radiological leakage through the steam generator tubes is one of the bounding conditions in emergency plan evaluation and the extent of steam generator tube failures directly impacts the FSAR analysis. The Replacement Steam Generator (RSG) modifications at San Onofre increased both the likelihood of equipment failure and the radiological consequence of such failure and therefore directly affect the FSAR Current Design Basis.”

4 – Unit 2 Main Steam Line Break (MSLB) Scenario: The most severe design basis accident to meet the San Onofre Unit 2 TS 5.5.2.11.b.1 steam generator structural integrity is a MSLB at the first weld outside containment. This assumption minimizes the flow resistance between the break and the affected SG and maximizes the mass & energy (M&E) release. The analyses focus on M&E releases at licensed Rated Thermal Power (RTP or 100% Power). The outside containment case includes the assumption that the main steam isolation valve (MSIV) in the steam line with the least flow resistance fails to close following the main steam isolation signal (MSIS). This assumption maximizes the M&E release during a MSLB outside of the containment. Super-heating within the SG initiates upon U-tube uncovery as specified in the NRC Information Notice 84-90. The turbine stop valves are assumed to close instantaneously at the time of the reactor trip. This assumption is conservative for a MSLB event because the entire steam inventory at the time of reactor trip is assumed to be forced out the break in 300 seconds (5 minutes). No Operator action outside Control Room can be assumed, if it takes less than 30 minutes. Westinghouse states, “it should be understood that there is more mass of secondary coolant in the steam generator at no load than at full power. Therefore, no load is the worst case for steam line break analyses.”

The depressurization of the non-isolable steam generator would result in 100% void fractions in the degraded Unit 2 U-Tube bundle due to instant flashing of the sub-cooled 440 degrees Fahrenheit feedwater into steam. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI), flow-induced random vibrations and excessive hydrodynamic pressures (Mitsubishi Flowering Effect). The force of the flashing steam would create high-energy jets, lifting loose parts and debris present in the steam generator, which would do additional damage by cutting holes into the already degraded tubes and creating additional loading (See Note 1 below) on the tube support plates (TSPs) due to heavy build-up of deposits on trefoil/quadrifoil-shaped holes from SG blowdown and crack the high cycle fatigued U-bend tubes not supported by Anti-Vibration Bars (AVB). These cumulative adverse conditions in all likelihood would result in a massive cascading of RSG’s tube failures (tubes would excessively rattle or vibrate, hitting other tubes with violent impacts) due to extremely low tube-to-tube clearances and no effective or non-existent in-plane anti-vibration bar support protection system. This jackhammering effect would involve hundreds of degraded active SG tubes along with all the inactive (plugged /unstabilized) tubes causing a catastrophic amount of simultaneous tube leaks/ruptures. Under this adverse scenario, 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 to fifteen minutes from a broken steam line would EXCEED the SONGS NRC approved offsite radiological release doses safety margins based on assumption of a single tube rupture in the SONGS FSAR. So, in essence, these RSG’s are like loaded guns, or a Fukushima-type nuclear accident, waiting to happen. Any failure under these conditions would allow significant amounts of radiation to escape to the atmosphere and a major Loss of Coolant Accident (LOCA) could easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor.

SCE states, “A MSLB alone does not generate sufficient differential pressure to cause tube rupture. The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators (See Note 2 below) prevent RCS re-pressurization in accordance with Emergency Operating Instructions.” SCE’s suggested DID Actions and proven unreliable operator actions to detect a leak and/or to re-pressurize the steam generators as claimed by Edison are not practical to stop a major nuclear accident from occurring in Unit 2 in the first 5-15 minutes of a MSLB during the proposed 5-month trial period.

NOTES:

1. Plugging of the at-risk tubes is not a satisfactory solution because it is the retainer bar that vibrates via random fluid flow processes at sub FEI critical velocity levels – these are likely to continue in play or, indeed, exacerbate at the proposed U2 restart at 70% power, leading to through-tube abrasion, the detachment of tube fragments, lodging at other unplugged and in-service tube localities, resulting in the so-called ‘foreign object’ tube wear. This additional loading would exceed: (1) the safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials, and (2) significantly affect burst or collapse pressures determined and assessed in combination with the loads due to a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. SCE’s suggested “defense-in-depth” actions are insufficient to stop multiple tube ruptures due to the short duration of a main steam line beak event. Human performance weaknesses, such as mis-diagnoses, substantial delays in isolating the faulted steam generator, communication errors and delayed initiation of the residual heat removal system, have been identified in past events at SONGS and other US Nuclear Power Plants. The events also involved unnecessary radiation releases, lack of RCS subcooled margin, excessive RCS cooldown rates, and overfilling the SG because of human or procedural problems.

CONCLUSIONS: Until the NRC can determine that San Onofre is 100% safe to operate at its approved rated power, granting any Unit 2 Restart testing is unacceptable, because if a nuclear accident occurred during testing who would be held liable, the Nuclear Utilities, the Insurance Carriers, the Federal Government, the State of California, the CPUC, the NRC Commissioners, NRC Region IV, EIX/SCE Shareholders & Employees or just the millions of affected southern Californians?  The DAB Safety Team believes that once the true amount of existing tube fatigue and all other associated damage is quantified, anything short of a total SG rebuild and/or SG replacement will be unacceptable prior to any restart being authorized by the NRC.

Special Comments about San Onofre

1. San Onofre is rated by the Institute of Nuclear Operations (INPO) as an INPO 4 Plant (The Worst Nuclear Plant Rating) and it should also should be rated in NRC Region IV Response Column V (Worst rating) and not in the NRC Response Column I (Best Nuclear Plant Rating).

2.    San Onofre is the worst nuclear plant in the country with the worst safety record, worst retaliation record, an INPO 4 rating and it is a mockery to place it in NRC Response Column I. NRC Region IV by listing San Onofre in NRC Response Column I, is putting its credibility on line and is displaying clear trends of collusion with SCE. It would be informative to learn who made the decision on San Onofre’s current ranking and why…

Definitions of NRC Response Columns [Source: NRC Inspection Manual Chapter 0305]

Column I – All Assessment Inputs (Performance Indicators (PIs) and Inspection Findings) Green; Cornerstone Objectives Fully Met
Column II – One or Two White Inputs (in different cornerstones) in a Strategic Performance Area; Cornerstone Objectives Fully Met.
Column III – One Degraded Cornerstone (2 White Inputs or 1 Yellow Input) or any 3 White Inputs in a Strategic Performance Area; Cornerstone Objectives Met with Moderate Degradation in Safety Performance

Column IV – Repetitive Degraded Cornerstone, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or 1 Red Input;  Cornerstone Objectives Met with Longstanding Issues or Significant Degradation in Safety Performance
Column V. Overall Unacceptable Performance; Plants Not Permitted to Operate Within this Band, Unacceptable Margin to Safety Licensed Activities

###

Definitions, Abbreviations and Acronyms

·         10 CFR 50.59 Safety Evaluation - Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR 50.59) defines the conditions under which reactor licensees may make changes to their facilities or make changes to procedures or to conduct tests or experiments without prior NRC approval. In general, such changes, tests, or experiments may be carried out unless they would involve a change to the technical specifications or an unreviewed safety question (as defined in  50.59(a)(2)). [Source: World Wide Web]

·         AIT: NRC’s Augmented Inspection Team

·         AREVA: Nuclear engineering firm owned by French Atomic Energy Commission

·         Arnold "Arnie" Gundersen is chief engineer of energy consulting company Fairewinds Associates and a former nuclear power industry executive, and who has questioned the safety of the Westinghouse AP1000, a proposed third-generation nuclear reactor. Gundersen has also expressed concerns about the operation of the Vermont Yankee Nuclear Power Plant. He served as an expert witness in the investigation of the Three Mile Island accident. Source: World Wide Web]

·         ATHOS is a three-dimensional computational fluid dynamics (CFD) code for analyzing steam generator (SG) thermal-hydraulic (TH) performance characteristics [Source: Westinghouse OA]

·         AVB: Anti Vibration Bar

·         CFR: Code of Federal Regulations

·         CPUC: California Public Utilities Commission

·         DBA: Design Basis Accident

·         ECT: Eddy Current Testing

·         EIX – Edison International

·         EPRI: Electric Power Research Institute

·         FEI: Fluid Elastic Instability is a phenomenon where the tubes vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports, because of lack of squeeze film damping, nucleate boiling or presence of localized regions in U-Tube bundles with high vapor fractions (e.g., >99.6%, steam voids, steam dry-outs, etc.) [Source: MHI/DAB]

·         FIRV: Flow-Induced Random Vibrations is a phenomenon where the tubes vibrate due to forces created by turbulent flow as a result of fluid velocity and density fluctuations. Vibration amplitudes due to random vibration are generally small (smaller than those due to tube fluid-elastic instability). [Source: MHI]

·         FSAR: Final Safety Analysis Report

·         FSM: Fluid elastic Stability Margin

·         FWLB: Feed-Water Line Break

·         John Large is a Chartered Engineer, a Consulting Engineer, a Fellow of the Institution of Mechanical Engineers, Graduate Member of the Institution of Civil Engineers and Fellow of the Royal Society of Arts. [World Wide Web]

·         LOCA: Loss Of Coolant Accident

·         MHI: Mitsubishi Heavy Industry

·         MSIV: Main Steam (line) Isolation Valve

·         MSLB: Main Steam Line Break

·         MWt: Mega-Watts Thermal

·         NOPD: Normal Operating Pressure Differential

·         NRC: Nuclear Regulatory Commission

·         NRC Reasonable Assurance is the recognition that “adequate protective measures can and will be taken in the event of a radiological emergency.” Reasonable assurance is based on licensees complying with NRC regulations and guidance, as well as licensees and offsite response organizations demonstrating that they can effectively implement emergency plans and procedures during periodic evaluated exercises. [Source: www.nrc.gov]

·         NRC San Onofre Special Review Panel - The NRC has established a special panel to coordinate the agency’s evaluation of Southern California Edison Co.’s proposed plan for restarting its Unit 2 reactor and ensuring that the root causes of problems with the plant’s steam generators are identified, and addressed. [Source: US NRC Blog]

·         NRR: NRC’s Office of Nuclear Reactor Regulations

·         OD: Outer Diameter

·         P/D: Pitch to Diameter ratio

·         OSG: Original Steam Generator

·         RCE: Root Cause Evaluation

·         RCPB: Reactor Coolant Pressure Boundary

·         RCS: Reactor Coolant System

·         RSG: Replacement Steam Generator

·         RWST: Refueling Water Storage Tank

·         SCE: Southern California Edison

·         SG: Steam Generator

·         SGTR: Steam Generator Tube Rupture

·         SM: Stability Ratio

·         SONGS: San Onofre Nuclear (Waste) Generating Station (alternate abbreviation: SONWGS)

·         TSP: Tube Support Plate

·         TTS: Top-of-Tube Sheet

·         TTW: Tube-to-Tube Wear

·         TW: Tube Wear

·         Westinghouse - Westinghouse Electric Company provides fuel, services, technology, plant design, and equipment for the commercial nuclear electric power industry.

The DAB Safety Team: Don, Ace and a BATTERY of safety-conscious San Onofre  insiders plus industry experts from around the world who wish to remain anonymous.  These volunteers assist the DAB Safety Team by sharing knowledge, opinions and insight but are not responsible for the contents of the DAB Safety Team's reports.  We continue to work together as a Safety Team to prepare additional: DAB Safety Team Documents, which explain in detail why a SONGS restart is unsafe at any power level without a Full/Thorough/Transparent NRC 50.90 License Amendment and Evidentiary Public Hearings.  For more information from The DAB Safety Team, please visit the link above.

Our Mission: To prevent a Trillion Dollar Eco-Disaster like Fukushima, from happening in the USA.

Copyright January 28, 2013 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 Attorney