Friday, February 1, 2013

Press Release 2013-01-31 Four NRC Augmented Team Conflicts of Interest


Press Release 13-01-31 Four NRC Augmented Team Conflicts of Interest

 

The DAB Safety Team: January 31, 2013

 

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

 



Four More Statements From NRC Region IV Augmented Inspection Team (AIT) That Require A Nuclear Reactor Regulation (NRR) Investigation And Resolution.


The DAB Safety Team Has Transmitted The Following Request To The Offices Of Chairman Of The NRC, The California Attorney General and Senator Barbara Boxer’s Committee on Environment and Public Works (EPW).

 

1. NRC AIT in its report dated November 09, 2012 (Re: NRC ADAMS Library Accession Number ML 2012010 - Unresolved Item 05000362/2012007-03, “Evaluation of Unit 3 Vibration and Loose Parts Monitoring System Alarms (V&LPM)”) closed the referenced item by stating that, “The inspectors determined that the licensee properly responded to and evaluated the alarms and followed the applicable station alarm procedures and vendor recommendations. Subsequently, the licensee requested from the vendor an in-depth evaluation of the available acoustical data, which was documented in Nuclear Notification NN 201818719. This evaluation established the likely source of the alarms. The results were inconclusive because of limitations with the monitoring system. Specifically, because of sensor locations (lower portion of the steam generator below the tube sheet in the support structure) and sensitivity, it was not possible to determine the exact source of the Unit 3 alarms. Westinghouse engineering personnel performed an evaluation (Evaluation 201818719-SPT-2) of acoustical data and determined from the shape and intensity of the particular
responses that the acoustic source was not likely from the upper bundle of the
replacement steam generator or related to the tube-to-tube wear. The licensee (SCE) is considering additional sensor locations which are not required, but may help with monitoring the upper bundle region of the steam generator during power operation. The results of this additional monitoring and increased sensor sensitivity may provide the licensee with a potential means to monitor for tube-to-tube degradation.” (See Page 14 for the limitations of this improved version of V&LPM system related to NO detection capability of tube-tube wear as claimed by AIT Team and SCE and questioned by NRR below).  According to the December 18, 2012 SCE NRC Public meeting Press and Webcast Reports, Edison officials came under sharp questioning about the Vibration and Loose Parts Monitoring System monitors at a U.S. Nuclear Regulatory Commission panel meeting in Maryland. Richard Stattel of the NRC’s Nuclear Reactor Regulation (NRR) Instrumentation Branch told the Edison Officials in a roaring and loud voice on an international live web cast, “The equipment could not do the job described by the company or provide additional safety if the plant is restarted. The instrumentation that you're proposing ... does not appear to be capable of detecting the conditions that would lead to actual tube wear." Edison depicted the equipment in its restart plan as an important safety measure "but it doesn't appear to do that.”  See the DAB Safety Team’s Press Release + 12-12-28 Thirty Alarms Demonstrates SONGS Unsafe for details on this subject.

DAB Safety Team Comments: The NRR is saying loud and clear that both NRC AIT and SCE Engineers need to understand the basic functions of “Safety-Grade” Instrumentation and the concept of “tube-to-tube” wear (Fluid Elastic Instability). 
Since there are no means of monitoring tube wall thinning while the plant is in service, the risk of tube burst is wholly dependent upon the accuracy and reliability of SCE’s “Safety-Grade” Instrumentation. The DAB Safety Team has stated earlier that NRC AIT Report is just a replication of SCE Root Cause Evaluation and not a true assessment by an Independent Regulator tasked with ensuring Public Safety.  On December 21, 2012, the US Nuclear Regulatory Commission (NRC) blog posted a letter from Chairman Macfarlane titled, “A Visit to Japan: Reflections from the Chairman.”  She said, “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.” According to the March 16, 2012 Press reports, Senators Barbara Boxer (D-CA), Chairman of the Senate Environment and Public Works Committee (EPW), and Dianne Feinstein (D-CA) sent a letter to the Chairman of the Nuclear Regulatory Commission (NRC), Dr. Gregory Jaczko, calling on the NRC to perform a thorough inspection at the San Onofre plant, located in San Clemente.  The collusion and casual relationship between NRC AIT Team and SCE requires an Investigation by the Offices of NRC Chairman and Senator Barbara Boxer to determine the impact on both future US reactor operations and emergency preparedness planning.   This investigation by the AIT does not meet the NRC Chairman’s Standards.

2. NRC AIT in its report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-03, "Evaluation of Retainer Bars Vibration during the Original Design of the Replacement Steam Generators”) closed the referenced item by stating that, “The inspectors determined that the licensee’s failure to verify the adequacy of the retainer bar design as required by SONGS Procedure SO123-XXIV-37.8.26 was of very low safety significance (Green) based on NRC Inspection Manual Chapter 0609.04, “Phase 1 – Initial Screening and Characterization of Findings,” and Inspection Manual Chapter 0609, Appendix A, “The Significance Determination Process (SDP) for Findings At-Power,” because the finding did not involve a degraded steam generator tube condition where one tube could not sustain 3 times the differential pressure across a tube during normal full power, steady state operation and none of the replacement steam generators violated the “accident leakage” performance criterion in plant Technical Specifications as a result of the retainer bar vibrations. The licensee also implemented actions to inspect all affected tubes in Unit 2 and 3 and remove from service all those tubes surrounding the smaller retainer bars that could wear due to vibration of the retainer bar. Because this violation has been determined to be of very low safety significance (Green) and has been entered in the licensee’s corrective action program as SONGS Nuclear Notification (NN) 201843216, it will be dispositioned as a non-cited violation in accordance with Section 2.3.2 of the NRC’s Enforcement Policy.”

John Large, internationally known Consulting Engineer, Chartered Engineer, Fellow of the Institution of Mechanical Engineers, Graduate Member of the Institution Civil Engineers, Learned Member of the Nuclear Institute and a Fellow of the Royal Society of Arts states concerning SONGS Restart Unit 2 in his testimony to the Atomic Safety Licensing Board, “In October 2012 MHI reported directly to the NRC safety concerns about the retainer bars:  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 a large amplitude. 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 [SCE] 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.  According to MHI, it is the lower resonance frequency of the smaller diameter retainer bars that is susceptible to turbulent two-phase flow exciting the bar into its prime resonance or some harmonic frequency thereof [p10, item 3].14 Whatever, a number of the tubes capturing the retainer bar had sustained abraded wear from interaction with it. These tubes comprised six tubes in U2 and four tubes in U3, with seven tubes in total showing wear greater than the 35% limit of the tube wall thickness for which isolation from service is required by plugging with, as previously noted, an incidence site in one of U2 RSGs having worn through 90% of its wall thickness. I agree with the findings of MHI that the tube wear at the retainer bar localities arises because of random flow induced (not FEI) vibration of the retainer bar itself, it being entirely independent of any tube motion excited from other sources. However, MHI’s advice to either plug the local tubes and/or remove the retainer bars at risk raises two issues unique to the retainer bar and its sub-assembly: (i) 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; (ii) MHI’s recommendation that those retainer bars at risk of large-amplitude fluid flow excited vibration should be removed is, of course, dependent upon reliable analysis to identify the at-risk assemblies; and, importantly, and (iii) this restraint system probably also serves to contain the tube bundle geometry during a main line steam break (MSLB) design basis event, so any change or removal of the retaining bar assemblage would require a full safety justification.”

Westinghouse states, “For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. The recirculating fluid flow rate is relatively constant at all power levels. However, in the U-bend region, the gap velocities are a strong function of power level. The steam flow in the bundle is cumulative and increases as a function of the power level and the bundle height which causes high fluid quality, void fraction, and secondary fluid velocities in the upper bundle.”

SCE in its November 30, 2012, NRC Presentation stated, “Four tubes with retainer bars wear above 35% limit in Unit 2 were plugged.”  The NRC website states, “The severity of one of the wear indications at a Unit 2 retainer bar was significant enough (90 percent thru-wall) to warrant in-situ pressure testing. This pressure test confirmed the structural integrity of this tube (there was no leakage).” (Above Bolding Added)

DAB Safety Team Comments:  Let us summarize what John Large and Westinghouse are saying: (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 to vibrate 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/or in-service tube localities, resulting in the so-called ‘foreign object’ tube wear, (2) For most of the straight leg section of the tube, the gap velocities at lower power levels and at 100% power are similar. Therefore, even at 70% power, the tube-to-retainer bar wear will continue at the same rate as 100% power and plugging the tubes is not a satisfactory solution in terms of reducing the active tubes rupture safety risks. SCE is not stating the facts either in its Root Cause Evaluation nor in its NRC Presentation. Two better questions are, “How many tubes in Unit 2 have what amounts of fatigue cracks and why has SCE not used state-of-the-art technology to visually examine all RSG tubes at San Onofre?”  What this really means is that Southern Californians were lucky once again, that Unit 2 just happened to be shutdown for refueling!  Otherwise, one or more worn tubes could have leaked or failed due to a design bases accident and/or any unanticipated transients.  Almost 180 tubes had to be plugged and stabilized in Unit 2 Replacement Steam Generators due to retainer bar design mistakes. In addition, no reports are available to determine the extent of tube fatigue damage or damage to the small retainer bars caused by the worn tubes and whether the damaged retaining bars are strong enough to restrain the movement of the anti-vibration bar assembly during a main steam line break design basis event (Ref: NRR RAI #32).  The design of the retainer bars approved by SCE and manufactured by MHI clearly violated the Code of Federal Regulations, 10 CFR Part 50, GDC 14, “RCPB—shall have “an extremely low probability of abnormal leakage…and gross rupture” and Appendix B, Criterion III, “Design Control.”   The DAB Safety Team’s opinion is that NRC AIT is treating the retainer bar mistakes and its design approval by SCE just as a routine matter like “No big deal, nothing happened, so who cares” instead of performing the strict enforcement required of an Independent Regulator tasked with ensuring Public Safety. This investigation by the AIT does not meet the NRC Chairman’s Standards.

3. NRC AIT in its report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-08, “Non-Conservative Thermal-Hydraulic Model Results”) states that, “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 the original Report in July 2012, the NRC AIT concluded that, “Due to modeling errors, the SONGS replacement generators were not designed with adequate thermal hydraulic margin to preclude the onset of fluid-elastic instability.” NRC

John Large states, “I identify a number of issues with 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, … 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. A complete understanding of the causation of the in-plane FEI is essential to ensure that the SONGS Unit 2 plant is acceptably safe to restart and, once restarted, predictably safe to continue in operation over the proposed 150 day inspection interval. To the contrary, the understanding presented by SCE is neither comprehensive nor convincing. In my opinion, simply sweeping the FEI issue under the carpet on the basis of (in- or out-of-plane) FEI will not reoccur at 70% power is not only disingenuous but foolhardy.”

Arnie Gundersen states, “The AIT report indicated that the change to the FIT-III evaluation methodology was not discussed as part of Edison’s 50.59 screening because the details of thermal hydraulic models used for the design of the OSG were not discussed in the original FSAR. It should have been obvious to Edison that 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 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.”

Gundersen further states, “The AIT reported that FIT-III predictions differed considerably in comparison to an Electric Power Research Institute developed code named ATHOS. FIT-III predicted lower flow velocities and void fractions that were not conservative compared to ATHOS. The AIT Report neglected an analysis of the root cause of the critical differences between FIT-III and ATHOS, and the negative impact such lax calculational modeling had on the design, fabrication, and successful operation of the San Onofre RSGs. Had Edison sought the required FSAR license amendment, differences between FIT-III and ATHOS would have been identified six years ago. The AIT did not address the possibility that the lack of conservatism in FIT-III predictions, in addition to causing tube vibrations, could also result in non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR. The AIT noted that the non-conservatisms in FIT-III are a contributor to the failure by Edison to adequately calculate the San Onofre RSG tube vibrations. But equally important, the AIT failed to address that FIT-III could also create non-conservative predictions of the behavior of the steam generator pressure vessel and associated main steam piping during accident conditions that are required to be analyzed in the FSAR. Such a conclusion implies that damage to the steam generator pressure vessel itself, and not just the tubes, might have occurred at San Onofre and remains unanalyzed by either Edison or the NRC. The probability of an accident exceeding the plant’s Current Design Basis is increased by the radically different Edison Replacement Steam Generators. Hence, the risks involved in operating the San Onofre RSGs should have been addressed as part of an FSAR license amendment and hearing process. It is my professional opinion that Edison should have applied for the 50.59 process so that the FSAR license amendment evaluation and public hearings would have occurred six years ago, prior to creating an accident scenario and facing losses that by the end of this process will easily total more than $1 Billion. The seriousness of the licensing and safety impact of the damaged RSGs at San Onofre cannot be overstated or underestimated. Any Design Basis Accident (DBA) as defined in the FSAR needs to be accurately modeled in order to protect public health and safety. The FSAR’s DBA analysis including the extent of tube leakage in the event of a Main Steam Line Break significantly impacts the design and implementation of Emergency Evacuation Plans. In the event of a steam line break accident in the San Onofre Replacement Steam Generators with the degraded condition of the tubes, an accident would have occurred that is more severe than any design basis accident scenario previously analyzed by Edison in the FSAR. Such a DBA steam line break accident would render the San Onofre emergency plan totally inadequate and most likely cause  a permanent evacuation of a large portion of Southern California.”

DAB Safety Team Comments:  After the June 18, 2012 public Meeting, the NRC AIT Team Chief announced to the world, "The computer simulation used by Mitsubishi during the design of the steam generators had under-predicted velocities of steam and water inside the steam generators by factors of three to four times."  Now, six months later, the AIT Team is saying the matter is unresolved.  The AIT Team is just repeating what SCE says or is not sure what they said four months ago.  ATHOS Modeling results are not reliable, because the results by NRC AIT Team, Westinghouse, MHI, AREVA and Independent Experts show that fluid elastic instability occurred both in Units 3 and 2.  The investigations in the Root cause of SONGS Unit 3 FEI regarding computer modeling have not been completed by NRC AIT Team, SCE and MHI.  FEI did not occur in Unit 2 according to DAB Safety Team and Westinghouse.  As also shown in other DAB Safety Team reports, FEI was not caused in Unit 3 by tube-to AVB gaps as claimed by NRC AIT Team and SCE.  This is consistent with the findings of Westinghouse, AREVA, MHI, John Large and SONGS Anonymous Insiders. The AIT Team is hurting its own credibility by issuing contradicting and conflicting statements. This investigation by the AIT does not meet the NRC Chairman’s Standards.
performed by Mitsubishi in determining flow and heat transfer by Mitsubishi in d
4. NRC AIT report dated November 11, 2012 (Re: Unresolved Item 05000362/2012007-10, “Evaluation of Departure of Methods of Evaluation for 10 CFR 50.59 Processes”) closed the referenced item by stating:  (a) The change from ANSYS to ABAQUS did not require a license amendment prior to implementing the change, so with respect to section 2.10.D.6 of the NRC Enforcement Manual, there is no reasonable likelihood that the change from ANSYS to ABAQUS would ever require NRC approval. Therefore, in accordance with the NRC Enforcement Manual, the inspectors determined that the licensee’s change from ANSYS to ABAQUS was a minor violation of 10 CFR 50.59(d)(1), and (b)n Based on this, the inspectors determined that the licensee had changed from using ANSYS and STRUDL to analyze several events for the original steam generators, to using only ANSYS to analyze a single limiting event for the replacement steam generators. Therefore, because the licensee did not change the method described in the Updated Final Safety Analysis Report, the inspectors concluded that the licensee did not need to obtain a license amendment prior to implementing that change. In the original Report in July 2012, the NRR technical specialist reviewed SCE’s 10 CFR 50.59 evaluation and found two instances that failed to adequately address whether the change involved a departure of the method of evaluation described in the updated final safety analysis report: (a) Use of ABAQUS instead of ANSYS:  The SCE’s 50.59 evaluation incorrectly determined that using the ABAQUS instead of ANSYS was a change to an element of the method described in the updated final safety analysis report did not constitute changing from a method described in the updated final safety analysis report to another method, and as such, did not mention whether ABAQUS has been approved by the NRC for this application.
(b) Use of ANSYS instead of STRUDL and ANSYS:  While SCE’s 50.59 evaluation correctly considered this a change from a method described in the FSAR to another method, the 50.59 evaluation did not mention whether the method has been approved by NRC for this application.

NRC AIT Report states, “For the Unit 2 and Unit 3 replacement steam generators, the licensee determined that the proposed activity did not adversely affect a design function, or the method of performing or controlling a design function described in the updated final safety analysis report. The licensee evaluated the following updated final safety analysis report design functions in the 50.59 screening: Steam Generator Design Functions. Let us examine the effect of these changes on Steam Generator Design Functions: The design functions of the steam generators tubes and tube supports are to:  (1.) Limit tube flow-induced vibration to acceptable levels during normal operating conditions, and (2) Prevent a tube rupture concurrent with other accidents.

Change Number 1: 105,000 square feet tube heat transfer area in OSGs; 116,100 square feet tube heat transfer area in RSGs; 11.1% increase in heat transfer area, which is more than a minimal change of 10% in the non-conservative direction.  Change accomplished by addition of 377 tubes in the central region by removal of stay cylinder and increasing the length of 9727 tubes by > 7 inches in each of the four RSGs.

Change Number 2: Operating Secondary Pressure – OSGs: 900 psi, RSG: 833 psi ~ 10% change – A catastrophic change for onset and ongoing exponential fluid elastic instability.


Change Number 3: Tube wall thickness was reduced from 0.048 inches to 0.043 to pump more reactor coolant through the tubes > 11.6% change.  The latest academic research indicates that the tube vibrations become large as T/D decreases and L/D increases, because the in-plane tube vibrations strongly depend on the dynamic characteristics of tubes such as the natural frequency and the damping ability.


Four other changes: Moisture content was reduced from 0.2% to 0.1% to improve SG performance, RCS Volume was increased from 1895 cubic feet to 2003 cubic feet, RCS Flow was increased from 198,000 gpm to 209,000 gpm, feedwater flow was increased from 7.4 million pound per hour to 7.6 million pound per hour and AVBs were not designed to prevent against adverse effects of fluid elastic instability (In-plane vibrations, Tube-to-Tube wear, steam dry-outs).  These unapproved and unanalyzed changes were claimed to be a conservative decision and improvements in the RSGs from OSGs were presented as a "like for Like" change.  No mixing baffles were added in the SONGS RSGs to improve the T/H Performance in the RSGs.  FEI and SR Values were not provided by SCE in the RSG Design Specifications.  SCE told MHI to avoid the NRC Approval…  MHI neither provided in-plane supports, nor provided the operational criteria to prevent FEI in one of the largest steam generators with such high steam flows.  MHI did not benchmark CE SG Computer codes or design details, neither did SCE, nor did SCE check the work of MHI.  And Dr. McFarlane says, “SCE is responsible for the work of its vendors and contractors.  Look at Palo Verde RSGs, a Success Story and SONGS RSGs, a $ Billion Blunder…. For complete list of changes, as identified by Arnie Gundersen and the DAB Safety Team see The Big #1 Attachment.

NRC AIT Report states, “The licensee’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 a 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.”

Arnie Gundersen states, “As the NRC confirmed in its AIT report, a large steam void has developed near where the additional tubes were added in the Replacement Steam Generators (called fluid elastic instability) that allows many types of excess vibrations to occur. Fairewinds review of Edison’s Condition Report clearly shows that the location within the steam generators where the steam “fluid elastic instability” has developed is precisely the region where the extra heat created by the 400 new tubes would create an excess of steam and various vibrational modes.”

NRC AIT report states, “Mitsubishi’s preliminary explanation of the failure mechanism started with the combination of two factors: (1) a relatively small tube pitch to tube diameter ratio (P/D), and (2) high void fraction in the tube bundle area where the tube-to-tube wear was identified. The small pitch to diameter ratio was a fixed parameter in the replacement steam generators established by the nominal center-to-center distance between adjacent tubes (P) and the nominal outside diameter of the tubes (D). The high void fraction was identified from the results of Mitsubishi’s thermal-hydraulic model for the secondary side of the replacement steam generators. Mitsubishi considered that the combination of these two factors may have resulted in favorable conditions for in-plane tube vibration based, in part, on the results of recent studies in fluid-elastic instability.” Mitsubishi also states, “Low secondary pressures are severe for vibration.”

John Large states, “Referring to the short section of the FSAR provided to me by SCE, which I understand is not to be amended for the Unit 2 restart: (a) 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 (b) 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.”
05000362/2012007-10, “Evaluation of Departure of Method of
John Large states, “SCE’s assertion that reducing power to 70% will at the best alleviate, but not eliminate, the TTW and other modes of tube and component wear is little more than hypothesis - the supporting Operational Assessments and analyses have not proven it to be otherwise. I am of the opinion that trialling this hypothesis by putting the SONGS Unit 2 back into service will, because of the uncertainties and unresolved issues involved, embrace a great deal of change, test and experiment.  The terms of the Confirmatory Action Letter of March 11 2012, are versed such that to meet compliance the response of SCE via its Return to Service Report,11 must include considerable changes of conditions and procedures that are outside the reference bounds of the present FSAR – this is because the physical condition of the RSGs, and the means by which this is evaluated and projected into future in-service operation, have substantially and irrevocably changed since the current FSAR was approved. The fact that SCE fails to satisfy the requirements of the CAL is neither here nor there, although it illustrates the scope and complexity of the response required. At the time of preparing the CAL, the NRC being well-versed in the failures at the San Onofre nuclear plant, surely must have known that the only satisfactory response to the CAL would indeed require considerable changes, tests and experiments to be implemented.”

DAB Safety Team Comments:  Therefore, the DAB Safety Team concludes that the changes in design functions of the RSGs tubes and tube supports described above definitely: a) did not limit tube flow-induced vibration to acceptable levels during normal operating conditions and, b) involved a significant reduction in a margin of safety – Failure of 8 Unit 3 SG Tubes under MSLB test conditions and significant TTW  > 35% of ~381 tubes in Unit 3 RSGs.  A multiple tube failure event, if actually would have occurred during a MSLB would have resulted in a significant increase in the off-site radiological consequences over the single tube burst event, if currently considered in the SONGS approved FSAR by NRC Region IV. 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. The AIT has no business contradicting conclusions made earlier by the NRR technical specialist. This investigation by the AIT does not meet the NRC Chairman’s Standards.
NRC Region IV Response to DAB Safety Team Analysis of SONGS 10 CFR 50.59 Evaluation Comments:  The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. These reviews involved NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters. The results of these reviews are contained in NRC two inspection reports that are available at http://www.nrc.gov/info-finder/reactor/songs/tube-degradation.html. [see the Augmented Inspection Team Report dated July 18, 2012, and the Augmented Inspection Team Follow-Up Report dated November 9, 2012]. It is worthy of note that the NRC staff is currently reviewing 10 CFR 50.59 documents associated with the licensee’s proposed restart activities. The results of the ongoing review will be documented in a future inspection report. NRC
Comments from Mel Silberberg [NRC-RES, Retired [Chief, Severe Accident Research Branch; Waste Management Branch] to Region IV: I am disappointed in the composition of the special panel! Where is the representation from NRC-RES? The issues at SONGS involve thermal hydraulics and material science. The NRC-RES and its contractors are experts in these areas. The Office of Research was created by the Congress for such situations. Two RES staff covering these disciplines and one or two consultants, serving as peer-reviewers. Perhaps there needs to be a separate peer review. Public confidence can only be gained using logical, informed measures as I described above.  Inspection Reports are only one facet of the problem, no question. However, understanding the reasons for the fluid instability, possible cavitation corrosion effects, etc. are phenomena which require evaluation by T/H as well as materials experts, with appropriate oversight by the ACRS. The SCE, the nuclear industry, the NRC and the public need assurance, not educated guesses. I have not seen a bona fide attempt to understand resolve the issue such that all can be alert to potential problems. I still remain puzzled as to why the ACRS [at least one of the Subcommittees].  I am trying to reach the ACRS Exec. Director to discuss this point. Thank you.” reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS. These reviews involved NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor R
DAB Safety Team Comments:  NRC Region IV Inspectors need to be re-trained in interpretation of significance of 10 CFR 50.59 Evaluation rules and meaning of changes in design function on safety evaluations. Simply sweeping the 10 CFR 50.59 mistakes under the carpet on the basis of meaningless statements, “The NRC has already conducted several reviews of the 10 CFR 50.59 documents associated with the replacement of the steam generators at SONGS.  The present SONGS NRC approved for the total S/G tube leakage assumes a limit of 1 gpm for all S/Gs, which ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10CFR100 limits in the event of either a S/G tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 0.5 gpm (720 gpd) leakage limit per S/G ensures that S/G tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.” These reviews of 10 CFR 50.59 and SONGS FSAR S/G tube rupture limits from NRC inspectors from multiple offices including Region IV, Region II and the Office of Nuclear Reactor Regulation at NRC headquarters” are not only disingenuous but foolhardy.   A single tube leakage and/or rupture could result in a nuclear incident or accident with tube leakages assumed in the current SONGS FSAR as shown in the Table below. A multiple tube failure event for all phases of the reactor in-core fuel cycle, would result in a significant increase in the off-site radiological consequences (e.g., Fukushima, Chernobyl, etc.) over the single tube burst event currently considered in the FSAR. The rapid and extraordinarily severe wear that resulted in the 2012 failures of all of Edison’s San Onofre Replacement Steam Generators was the result of Edison’s 2005 decision to radically change the RSG design and to claim that the Part 50.59 licensing process did not apply. Arnie Gundersen and DAB Safety Team have stated consistently that 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. While the NRC Augmented Inspection Team (AIT) briefly described how Edison addressed its 50.59 requirements, the evidence shows that Edison did not comply with the NEI guidelines for implementing 50.59. Published reports indicate that the strategic decision made by Edison that the 50.59 process would not be applied to the RSGs was made by corporate officials before any engineering personnel had actually performed the 50.59 engineering analysis. Consequently, Edison made a management decision to claim that the 50.59 process did not apply and therefore San Onofre was not required to seek NRC approval for the proposed changes at San Onofre Units 2 and 3. These unlicensed unapproved design changes to the containment boundary violated Federal Regulations and therefore the FSAR must be amended prior Unit 2 Restart to reflect multiple steam generator tube ruptures with MSLB plus DBE due to Edison’s significant untested and unanalyzed modifications.
  



 History of Steam Generator Tube Ruptures
Year
Plant
Location
Flow Rate
Rupture or Leak
1975
Point Beach 1
Wisconsin
125 gal/min
Rupture
1976
Surry 2
Virginia
330 gal/min
Rupture
1979
Prairie Island 1
Minnesota
390 gal/min
Rupture
1982
Ginna
New York
630 gal/min
Rupture
1987
North Anna 1
Virginia
600 gal/min
Rupture
1989
McGuire 1
North Carolina
500 gal/min
Rupture
1991
Mihama 2
Japan
Undetermined
Rupture Level 3 INES Incident – Offsite Releases
1993
Palo Verde 2 
Arizona
240 gal/min
Rupture
2000
Indian Point 2 
New York
  90 gal/min
Rupture
2004
Crauss NP
France
Unknown
Leak
2005




2006




2012
SONGS  3


+ SONGS  2
had  a tube with 90% Wear
California
625-750 gal/min
(See Note 1 below)
Leak, almost 8+ Ruptures



8 + Ruptures ~=




5000 - 6000  gal/min1


NOTE 1: Arnie Gundersen states, “Edison dramatically increased the radiation risk to the public as a result of San Onofre with Replacement Steam Generators that were extremely flawed beginning with their original design. The fact that 8 tubes failed the pressure tests in Unit 3 indicates that those tubes would have failed during a main steam line break (MSLB). It is uncertain if a reactor operator would have been able to shut the plant down without melting the core. A simultaneous rupture of 8 tubes would have caused a primary to secondary leak of radioactive coolant of about 5000-6000 gallons per minute. This leakage would have begun to drain the nuclear core as well as releasing radioactive primary coolant to the atmosphere. The ability of a reactor operator to control the water level in the affected steam generator with this high leakage rate and keep the nuclear reactor core cooled has never been analyzed or tested. An accident of this magnitude is outside ANY reactor’s Current Design Basis (CDB).”

ABBREVIATIONS AND ACRONYMS
·         ACRS: NRC’s Advisory Committee on Reactor Safeguards
·         ADAMS: NRC’s Agencywide Documents Access and Management System
·         AIT: NRC’s Augmented Inspection Team
·         AREVA: Nuclear engineering firm owned by French Atomic Energy Commission
·         AVB: Anti Vibration Bar
·         CAL: Confirmatory Action Letter
·         CFR: Code of Federal Regulations
·         CPUC: California Public Utilities Commission
·         DBA: Design Basis Accident
·         ECT: Eddy Current Testing
·         ECCS: Emergency Core Cooling System
·         EDF: French nuclear parts manufacturing company, also owns transmission lines in France, etc.
·         EPRI: Electric Power Research Institute
·         FEI: Fluid Elastic Instability
·         FIRV: Flow-Induced Random Vibrations
·         FSAR: Final Safety Analysis Report
·         FSM: Fluid elastic Stability Margin
·         FWLB: Feed-Water Line Break
·         GDC: General Design Criteria
·         GSI: Generic Safety Issue
·         ID: Inner Diameter
·         INES: International Nuclear Events Scale
·         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
·         NPP: Nuclear Power Plant
·         NRC: Nuclear Regulatory Commission
·         NRR: NRC’s Office of Nuclear Reactor Regulations
·         OD: Outer Diameter
·         P/D: Pitch to Diameter ratio
·         OSG: Original Steam Generator
·         PRA: Probabilistic Risk Assessment
·         PVNGS: Palo Verde Nuclear Generating Station
·         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 Generating Station
·         SR: The specific mechanism for the flow-induced vibration has been determined to be a fluid-elastic instability.  The fluid-elastic mechanism has a significant effect on tube response in cases where the fluid-elastic stability ratio equals or exceeds 1.0.  The stability ratio, SR, is defined as:  SR = V eff / V c, where V eff is the effective cross-flow velocity and V c is the critical velocity beyond which the displacement response to the tube increases rapidly.
·         TS: Technical Specifications (for operation of a NPP)
·         TSP: Tube Support Plate
·         TTS: Top-of-Tube Sheet
·         TTW: Tube-to-Tube Wear
·         TW: Tube Wear
·         UT: Ultrasonic Testing
·         V&LPM: The Vibration and Loose Parts Monitor - consists of Vibration and Loose Parts Channels. Each system consists of piezoelectric sensors, preamplifiers, a signal processor unit and other peripheral equipment. The vibration and loose parts monitoring system is designed to provide continuous monitoring and conditioning of loose parts accelerometer signals. Two separate accelerometers are installed on each of the steam generators. The location of these instruments are on the steam generators’ lower supporting structures and provide acoustic information about loose parts impacts specifically on the reactor coolant or primary side of the steam generators. The vibration and loose parts monitoring system real time functions consist mainly of impact alarm validation of suspected loose part events and recording acoustic data.  The improved Westinghouse DMIMS-DX™ systems are installed at Millstone 3, Krsko, Diablo Canyon 1 & 2, Wolf Creek, and Beaver Valley 1, which are all Westinghouse NSSS plants, and at Crystal River 3, which is a B&W plant. DMIMS-DX™ provides fast, reliable detection of loose part impacts within the Reactor Coolant System (RCS), while minimizing the generation of false alarms. This monitoring system is a greatly enhanced version of the previous Westinghouse DMIMS system, employing the latest digital technology and offering significant operational advantages to our customers.  Loose parts monitoring is based on listening for the impact of loose parts against fixed components within the primary system as they are propelled by the coolant flow. This application appears simple on the surface, but its effective implementation is not an easy task. The noises typical of an operating plant can generate false alarms that reduce operator confidence, interfere with normal operations, and cause unnecessary expense. The Westinghouse DMIMS System uses a patented algorithm to determine the metallic characteristics typical of loose parts. This algorithm and the associated alarm algorithms, together, minimize the generation of false alarms and have established a reputation for reliability within the industry.
This press release will be posted on the web at this linkDAB Safety Team Documents.

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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 31, 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