ML13169A316

LTR-13-0530 Vinod Arora, Multiple E-mails San Onofre Nuclear Generating Station

ML13169A316
Person / Time
Site:	San Onofre
Issue date:	05/20/2013
From:	Arora V - No Known Affiliation
To:	Freedhoff M, Macfarlane A M, Walls W - No Known Affiliation, NRC/Chairman, NRC/OIG
References
LTR-13-0530, NRC-2013-00701
Download: ML13169A316 (20)

Text

Joosten, Sandy From: Sent: To:

Subject:

Attachments:

Vinod Arora [vinnie48in@gmail.com]

Monday, May 20, 2013 8:14PM Freedhoff, Michal; Walls, William San Onofre Sad Saga and Dangers of Unsafe Unit 2 Restart at 70% Power Operational Difference between SONGS Units Rev 1.docx This email is a proprietary and. confidential document of AVP Arora International Inc. and is for the benefit of the intended or authorized recipients only. If you receive this email in error, please destroy the contents and notify the Originator.

Disclosure of the contents of this email is strictly prohibited without the. written consent of the .Originator and his Attorneys.

Reference:

Nuclear Regulatory Commission

[Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2] Good Afternoon Mr. Walls and Dr. Freedhoff (For Submission to Senator Barbara Boxer Email and Attachment), Thanks and please write me an email, if you have any questions

.. God Bless You .. VA

Subject:

For The NRC Office Inspector General and Senator Barbara Boxer's Kind Consideration and Appropriate Action to protect the health and safety and environmental implications on 8.4 Million Southern Californians from a Potential Nuclear Accident due to restart of Unsafe San Onofre Unit 2 without adequate repairs, a Full 1 OCFR 50.90 License Amendment and Trial-Like Public Questioning of SCE and its Vendors and NRC Staff. Required Action: The NRC Chairman has publically stated that SCE is responsible for the work of MHI, Westinghouse, AREVA and lntertek.

From all the available past and present evidence, NRC Staff appears to be biased in favor of SCE restarting Unsafe Unit 2 without a thorough review of Unit 2 SCE Return to Service Reports. NRC staff has shown a lack of understanding of the full safety and environmental implications on 8.4 Million Southern Californians of a potential nuclear accident by restart of Unit 2 and appears to be opposing a full 1 OCFR 50.90 License Amendment and Public Hearings as expressed in ASLB Ruling. The safety issues presented in the attached documents are real, not trivial, as acknowledged in the ASLB panel decision of May 13, 2013. Under applicable law, the NRC Staff must withdraw the SCE 50.92 license amendment proposal and refer the proposed license amendment to the ASLB for an adjudicatory hearing, before a decision on the proposed amendment can be made. 1 NRC Staff is in direct violation of federal and its own rules and its public safety charter mission mandated under the law by the His Excellency, President of the United States and US Congress.

Therefore, The NRC Inspector General and Honorable, Senator Barbara Boxer in view of ASLB ruling are formally requested to retain an Independent Hydraulic Expert to examine the operational differences between Units 2 & 3 during Cycle 16 and determine its impact on NRC CAL Action 1 by examining the entire SONGS Cycle 16 operational data for Units 2& 3. Unit 2 Restart Permission at 70% power should be contingent on completion of the corrective actions required by NRC CAL Action 1, Results of ASLB Rulings, 1 OCFR 50.90 Review, Public Hearings and 1 OCFR 50 Appendix B. There are several important public safety questions regarding the full implications of FE I, tube-to-A VB contact forces, incubating cracks, tube fatigue life and adequacy of Unit 2 Tube Inspections, which have not been fully reviewed, understood, discussed and answered in Public Forums/Publically Disclosed Documents under the false shield of proprietary data and Licensee Protected Property by NRC Staff, SCE and its Consultants in a rush to Restart Unit 2. Summary: It is concluded that FEI and MHI Flowering effect redistributed the tube-to-A VB gaps in Unit 3 RSG's. FEI did not occur in Unit 2, because of the absence of high steam dryness and NOT the better supports and/or differences in fabrication, which resulted in substantially increased contact forces (reduced looseness) between tubes and AVBs for Unit 2 and prevented FEI from occurring.

These findings on Unit 2 FEI are consistent with the findings of AREVA, Westinghouse, John Large, SONGS RCE Anonymous Root Cause Team Member and latest research performed by Eminent Professor Michel Pettigrew and others researchers between 2003 through 2011. In-plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is just conjecture on part of SCE, MHI, AREVA and NRC AIT Team in Unit 2 to justify the restart of an Unsafe Unit 2. Even at 70% power, under Anticipated Operational Transients and Main Steam Line Breaks, entire Unit 2 RSG tube bundle will experience void fractions of 100%, fluid velocities

>50 feet per second and high energy jet impingement from flashing feed water. According to Mitsubishi recent testing data, additional thicker tubes with contact forces in excess of 30N are required in Unit 2 are required to prevent adverse effects of FE I. Based on the best available evidence, existing Unit 2 AVBs provide a significant smaller contact force than of 30N required to mitigate the adverse effects of multiple tube ruptures caused by the potential collapse of Anti-vibration Bar Structure and incubating cracks. Therefore, of particular concern with SONGS Unit 2 restart at reduced power are undetermined and unexamined amount of incubating circumferential cracks located in tubes next to each other caused by fluid-induced random vibrations, high cycle thermal fatigue and in-plane fluid elastic instability.

When one circumferentially cracked tube ruptures, the additional stresses can cause multiple or cascading tube ruptures, which can result in a nuclear meltdown.

In addition, though the Unit 3 steam generators failed more catastrophically, it appears thatthere is a much larger pool of tubes is out of alignment and is in direct contact with support plates in Unit 2. SCE, MHI, AREVA, lntertek, Westinghouse and NRC Staff are ignoring these cracks in their analyses.

The difference in management of Steam Generator Tube Rupture between Finland and USA is, that no primary coolant (liquid and steam) release to the environment is allowed in Finland, while 2 in USA, primary steam releases are not forbidden for profits to conduct risky experiments with people's lives. This situation is unique to San Onofre Steam Generator and the Potential Extent of Condition does not affect any other MHI and US Steam Generators.

Having reviewed the analyses of Mr. John Large, Dr. Hopenfeld, Dr. Gilinsky, Mr. Gundersen, and consistent with LBP-07-13, the May 13, 2013 opinion of the Atomic Safety & Licensing Board (ASLB), the undersigned fully endorses Friends of the Earth (FoE) and Natural Resources Defense Council (NRDC) request that the proposed SCE "No significant hazards" consideration determination should be withdrawn because (1) The NRC Staff's proposal exceeds the authority granted to it by the Sholly amendment; (2) the licensee's application of the criteria under 10 C.F.R. § 50.92, as adopted by the NRC Staff, does not justify a finding of no significant hazards consideration; and (3) the Staff have not performed an environmental review of the proposed finding and license amendment as required by the National Environmental Policy Act (NEPA), and the proposed actions do not satisfy criteria for a categorical exemption from NEPA review, provided at 10 C.F.R. § 51.22(c)(9)(i).

3

Reference:

Nuclear Regulatory Commission

[Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2] Required Action: The NRC Chairman has publically stated that SCE is responsible for the work of MHI, Westinghouse, AREVA and lntertek.

From all the available past and present evidence, NRC Staff appears to be biased in favor of SCE restarting Unsafe Unit 2 without a thorough review of Unit 2 SCE Return to Service Reports. NRC staff has shown a lack of understanding of the full safety and environmental implications on 8.4 Million Southern Californians of a potential nuclear accident by restart of Unit 2 and is opposing a full 10CFR 50.90 License Amendment.

The safety issues presented in the attached documents are real, not trivial, as acknowledged in the ASLB panel decision of May 13, 2013. Under applicable law, the NRC Staff must withdraw the SCE 50.92 license amendment proposal and refer the proposed license amendment to the ASLB for an adjudicatory hearing, before a decision on the proposed amendment can be made. NRC Staff is in direct violation of federal and its own rules and its public safety charter mission mandated under the law by the His Excellency, President of the United States and US Congress.

Therefore, The NRC Inspector General and Honorable, Senator Barbara Boxer in view of ASLB ruling are formally requested to retain an Independent Thermal-Hydraulic Expert to examine the operational differences between Units 2 & 3 during Cycle 16 and determine its impact on NRC CAL Action 1 by examining the entire SONGS Cycle 16 operational data for Units 2 & 3. Unit 2 Restart Permission at 70% power should be contingent on completion of the corrective actions required by NRC CAL Action 1, Results of ASLB Rulings, 1 OCFR 50.90 Review, Public Hearings and 1 OCFR 50 Appendix B. Summary: It is concluded that FEI and MHI Flowering effect redistributed the AVB gaps in Unit 3 RSG's. FEI did not occur in Unit 2, because of the absence of high steam dryness and NOT the better supports and/or differences in fabrication, which resulted in substantially increased contact forces (reduced looseness) between tubes and AVBs for Unit 2 and prevented FEI from occurring.

These findings on Unit 2 FEI are consistent with the findings of AREVA, Westinghouse, John Large, SONGS RCE Anonymous Root Cause Team Member and latest research performed by Eminent Professor Michel Pettigrew and others researchers between 2003 through 2011. plane fluid elastic instability did not happen in Unit 2 because of operational differences, so therefore double contact forces and better supports is just conjecture on part of SCE, MHI, AREVA and NRC AIT Team in Unit 2 to justify the restart of an Unsafe Unit 2. Even at 70% power, under Anticipated Operational Transients and Main Steam Line Breaks, entire Unit 2 RSG tube bundle will experience void fractions of 100%, fluid velocities

>50 feet per second and high energy jet impingement from flashing feed water. According to Mitsubishi recent testing data, additional thicker tubes with contact forces in excess of 30N are required in Unit 2 are required to prevent adverse effects of FEI. Best on the best available evidence, existing Unit AVBs have a significant smaller 1 contact force than 30N. Therefore, of particular concern with SONGS Unit 2 restart at reduced power are undetermined and unexamined amount of incubating circumferential cracks located in tubes next to each other caused by fluid-induced random vibrations, high cycle thermal fatigue and in-plane fluid elastic instability.

When one circumferentially cracked tube ruptures, the additional stresses can cause multiple or cascading tube ruptures, which can result in a nuclear meltdown.

In addition, though the Unit 3 steam generators failed more catastrophically, it appears that there is a much larger pool of tubes is out of alignment and is in direct contact with support plates in Unit 2. SCE, MHI, AREVA, lntertek, Westinghouse and NRC Staff are ignoring these cracks in their analyses.

The difference in management of Steam Generator Tube Rupture between Finland and USA is, that no primary coolant (liquid and steam) release to the environment is allowed in Finland, while in USA, primary steam releases are not forbidden for profits to conduct risky experiments with people's lives. This situation is unique to San Onofre Steam Generator and the Potential Extent of Condition does not affect any other MHI and US Steam Generators.

Having reviewed the submissions of SCE in support of the proposal to allow operation of Unit 2 at 70% of power and the analyses of Mr. John Large, Dr. Hopenfeld, Dr. Gilinsky, Mr. Gundersen, and consistent with LBP-07-13, the May 13, 2013 opinion of the Atomic Safety & Licensing Board (ASLB), the undersigned fully endorses Friends of the Earth (FoE) and Natural Resources Defense Council (NRDC) request that the proposed SCE "No significant hazards" consideration determination should be withdrawn because (1) The NRC Staff's proposal exceeds the authority granted to it by the Sholly amendment; (2) the licensee's application of the criteria under 10 C.F.R. § 50.92, as adopted by the NRC Staff, does not justify a finding of no significant hazards consideration; and (3) the Staff have not performed an environmental review of the proposed finding and license amendment as required by the National Environmental Policy Act (NEPA), and the proposed actions do not satisfy criteria for a categorical exemption from NEPA review, provided at 10 C.F.R. § 51.22(c)(9)(i).

Defects or Deviations:

The design of San Onofre replacement steam generators (RSGs) are identical.

As shown below, SONGS Unit 2 potentially did not suffer in-plane fluid elastic instability because of lower void fractions (98-98.8%

range) due to operation at higher steam pressures and lower RCS flows compared with Unit 3. SONGS Unit 3 suffered in-plane fluid elastic instability due to operation at lower steam pressures and higher RCS flows. If the operating and local thermal-hydraulic conditions were the same in both Units, then Unit 2 should have suffered tube-to-tube wear like Unit 3. This is because the double the tube-to-A VB contact force (2N) and better supports in Unit 2 are not enough to prevent FEI or tube-to-tube wear for the following reasons: (1) AREVA states, "A contact force of 1 N did not resist in-plane motion but a force of 1 ON was completely effective", and (2) MHI states, "Tube-to-A VB contact forces in excess of 30N will prevent in-plane tube-2 displacement and tube-to-tube contact in high region of wear." The number of Unit 2 AVB wear indications and their wear rates are less than that of Unit 3, because the lower void fractions (98-98.8%

range) in Unit 2 produced lower fluid velocities (25 feet/second), lower hydrodynamic pressure and hence lower intensity flow-induced random vibrations.

It is therefore concluded that lower intensity flow-induced random vibrations produced lower Unit 2 AVB wear indications with less wear rates than that of Unit 3. NRC AIT Report, SCE, MHI and AREVA conclusions on Unit 3 and Unit 2 operating and thermal-hydraulic conditions causing FEI and double the tube-to-A VB contact forces and better supports for prevention of FEI in Unit 2 are incomplete, inconsistent, confusing and inconclusive and based on faulty computer simulations and hideous testing data (Shielded under the false pretense of Proprietary information).

The analysis in these reports does not meet the intent of NRC CAL ACTION 1, which states "Southern California Edison Company (SCE) will determine the causes of the tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections." Repeated requests to NRC AIT Leader, NRC SONGS Special Panel and NRC Region IV Allegation Coordinator to examine carefully the operational difference between Units 2 & 3 and determine its impact on the tube-to-tube interactions and contact forces that resulted in steam generator tube wear in Unit 3, and actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes have not been addressed to date. NRR has not asked SCE in its RAI(s) the impact of operational differences between Units 2 and 3 on Unit 2 and Unit 3 tube-to-tube wear. Honorable NRC Commissioner Mr. Apostolakis was totally confused on Unit 2 FEI inconsistent statements by SCE, Westinghouse and AREV A.

Background:

Based on the 2011 research papers and 2012 Industry research, SONGS Unit 3 RSGs suffered localized in-plane fluid elastic instability in 4% of the u-tube bundle area of the hot-leg due to the following factors: 1. Severe vibrations resulted in excessive tube vibrations by operation at lower steam pressures of 833 psi to produce more thermal output from the RSGs 2. The heat transfer capability of the RSGs was exceeded due to higher heat flux of the hot-leg caused by RCS flows of -76 Million Lbs./hour.

This resulted in very high in-plane localized fluid velocities(>

35-50 feet/sec.), very high steam flows and high dry steam (vapor fractions

(>99.6%) in 4% of the u-tube bundle. This adverse change resulted in change of the flow regime in the affected area of the steam generator from nucleate boiling (1.5% water on tube surface) to film boiling (0% water on tube surface) 3. Narrow tube pitch to tube diameter (ranging from 0.05-0.25 inches) and extremely low tube clearances

4. Addition of 377 tubes in space vacated by removal of stay cylinder 5. Extremely tall tubes, average length of the heated tube increased by 50 inches (Addition of 650 tubes) 6. 11% increase in tube heat transfer area 3

7. Tubes-A VB zero gap during bundle in the hot and pressurized condition and AVBs with a contact force of > 1 N (Based on AREVA operational Assessment)

Mitsubishi states, "In general, structures in a two-phase flow field have lower resistance to vibration when a void fraction (percentage of vapor volume in a saturated mixture) or steam quality (percentage of vapor mass in a saturated mixture) is high. The high void fraction (steam quality) results in the two-phase fluid having a low density, which in turn results in an increase of the flow velocity of the two-phase fluid, and in a low damping factor. The increase of the flow velocity (v) causes the increase of the hydrodynamic pressure (pv2) which causes structures to vibrate in the flow field. The hydrodynamic pressure is a measure of energy imparted on the structure by the flow field, and damping is a measure of how easily the structure can dissipate this energy. If the amount of energy imparted on the structure is higher than the amount of energy dissipated, the structure (in this case the tubes) will vibrate with progressively increasing amplitudes, which eventually may lead to the tubes becoming fluid-elastic unstable.

Also, the unstable tubes will excite the surrounding tubes via two-way coupling with the fluid. Therefore, it is more likely for the tubes to vibrate when the void fraction (steam quality) is high." AREVA States, "Contact forces significantly increase as a bundle is heated and pressurized to operation conditions.

A small tube-to-A VB gap is effective against the out-of-plane vibrations forces, but a contact force between 1 N and 1 ON is needed to prevent in-plane vibrations.

A contact force of 1 N is ineffective against in-plane vibrations, but a force of 1 ON is totally effective to prevent in-plane vibrations." Mitsubishi, AREVA and SCE state that the operating and hydraulic conditions in both units were the same, but Unit 2 had double contact force than Unit 3, that is why FEI did not happen in Unit 2. Now Mitsubishi is proposing that Unit 2 and Unit 3 can be repaired to operate at 100% power with 2 new additional thicker AVBs inserted at 45 and 135 degree locations, in even numbered tube columns, between columns 66 and 112. MHI says this option introduces increase tube-to-AVB contact forces in excess of 30N and will prevent in-plane tube-displacement and tube-to-tube contact in high region of wear. MHI further says that the thicker bars will increase tube support effectiveness throughout the bundle, improve stability ratios and greatly reduced the induced random vibrations.

MHI is also recommending that with this repair option, operating conditions be changed to reduce the void fraction from 99.6% to 98.9% by changing secondary SG water level, feedwater and reactor coolant temperatures.

MHI further says that by changing these parameters, Steam pressures and circulation ratios will also increase.

Study of the recently collected data has led to a re-evaluation of the original design basis for the SONGS RSGs. MHI further states, "Several preliminary conclusions have been drawn for developing a design that is resistant to vibration:

1. The "effective zero gap" design concept is effective against "out-of-plane FEI" but for the AVB supports to be active and provide restraint in the in-plane direction requires sufficient tube-to-A VB contact force to generate friction that inhibits in-4 plane tube displacement.

Therefore, the zero gap assembly definition should have included a requirement for small, uniform contact forces (preloads).

2. The magnitude of the required contact force increases in regions of high void fraction (steam quality).

Tubes in the high void fraction (steam quality) region of the tube bundle U-bend are more susceptible to in-plane FEI and random vibration because the higher void fraction (steam quality) reduces the external fluid damping and the liquid film damping (squeeze film damping).

Therefore it is important to assure that upper bound thermal hydraulic values (void fraction, steam quality, flow velocities, damping, etc.) are assumed in the analysis of the design. 3. If small, uniform contact forces are incorporated, the design basis no longer needs to assume inactive supports and the number of supports does not need to be greater than what is needed to prevent out-of-plane FEI (i.e. four sets of AVBs instead of six would be sufficient)." MHI says that "Unit 3 RSGs were designed and fabricated with an "effective zero gap" in order to minimize its potential on tube wear. The technical investigations after the tube leak incident determined that the amount of contact force necessary to prevent in-plane FEI depends on the localized thermal-hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure).

As the steam quality (void fraction) increases, the amount of contact force necessary to prevent vibration increases.

This increase in required contact force occurs because as the steam quality (void fraction) becomes higher, the damping provided by the liquid phase in the form of a liquid film decreases.

The technical investigations after the tube leak incident determined that the amount of contact force necessary to prevent in-plane FEI depends on the localized hydraulic conditions (steam quality (void fraction), flow velocity and hydro-dynamic pressure).

As the steam quality (void fraction) increases, the amount of contact force necessary to prevent vibration increases.

This increase in required contact force occurs because as the steam quality (void fraction) becomes higher, the damping provided by the liquid phase in the form of a liquid film decreases.

The reduced in-plane contact force due to the SONGS "effective zero gap" design and the avoidance of "excessive preload" resulted in lowering the tubes' natural frequency in the in-plane direction.

The combination of the localized high steam quality (void fraction) and reduced tube to AVB contact force resulted in exceeding the in-plane critical velocity, which created a condition that led to tube to tube contact. The dominant role played by the low contact force is reflected by the differences in the tube-to-tube wear that was observed in the Unit 2 and the Unit 3 RSGs. Each of the Unit 3 RSGs had approximately 160 tubes that experienced tube-to-tube wear whereas only one of the Unit 2 RSGs experienced tube-to-tube wear in just two tubes, even though the Unit 2 RSGs have operated twice as long as the Unit 3 RSGs. MHI did a comprehensive statistical evaluation of the contact forces between the tubes and the AVBs of the two units and concluded, based on the manufacturing data, that the contact force between the tubes and the AVBs in the Unit 2 RSGs is approximately double the contact force in the Unit 3 RSGs. 5 Unit-3 SGs have slightly larger average tube-to-AVB gaps than the Unit-2 SGs, with the largest in SG-3A. This trend indicates the tube-to-A VB contact force of Unit-3 SGs are smaller than that of Unit-2 SGs. Thus, the lower contact forces in Unit 3 are consistent with the conditions determined necessary to permit in-plane FEI to occur and with the fact that tube-to-tube wear occurred almost exclusively in Unit 3." MHI contradicts itself by stating observation common about Units 2 and 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." Based on the above conditions, AREVA Operational Assessment and Mitsubishi Testing results, one has to conclude that a zero tube-to-A VB gap and contact force between 1 0-30N is required to arrest the in-plane vibrations of large u-bends moving with large amplitudes(>

0.25 inches) to prevent vibrating tubes from hitting the neighboring tubes with violent impact causing tube-to-tube wear and metal fatigue (incubating cracks). In Unit 3, high dry steam with a vapor fraction between 99.6-100%

was produced in 4% area of the high wear of the u-tube bundle due to excessive RCS flows and low steam pressures.

The high steam production was a result of the rejection of the Unit 3 heat RSG heat transfer coefficient with the change of the flow regime from nucleate boiling to film boiling. Since steam travels much faster than water, large U-bends overcame the zero gaps and the low contact forces of 1 N as stated by AREVA of the Unit 3 AVBs due to high in-plane velocities (35-50 feet/second) created by high dry steam with zero water on the tubes for damping. These large U-bends pushed the AVBs out of their way with a force in excess of 30N as demonstrated by MHI recent testing and created tube-A VB gaps larger in Unit 3 than Unit 2 as revealed during ECT inspections.

Now with the AVBs out of their way, vibrating tubes were free to move in the in-plane direction and hit neighboring tubes. Based on the results of ECT and visual inspections, computer simulations, statistical and tube manufacturing AVB twist data, MHI concluded that the all the 12 AVBs were not restraining the tubes in the-in-plane direction.

That is why the tubes with in-plane tube-to-tube wear in the high region of wear in Unit 3 did not show any tube-to-A VB wear, because these tubes only moved in the in-plane direction and did not move in the out-of-plane directions due to flow-induced random vibrations.

Mitsubishi says, even though Unit 2 operational and thermal-hydraulic conditions were same as Unit 3, Unit 2 tubes only experienced tube-to-A VB wear, because the Unit 2 tube-to-A VB contact forces (Assumed 2N) were double that of Unit 3 (Assumed 1 N) and prevented the tubes from hitting each other in the in-plane direction and prevented to-tube wear. SCE U3 RCE states, "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 RSGs. This might have resulted in reduction of the tube "to" AVB contact force and consequently in multiple consecutive AVB supports being inactive.

Inactive tubes supports might have resulted in tube "to" wear." Westinghouse states that "SCE has NOT done an extensive evaluation of Unit 3 6 Manufacturing issues." Westinghouse further 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 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." Westinghouse tests do not represent the conditions experienced in SONGS 3 steam generators, therefore Westinghouse conclusions are considered invalid for SONGS 3 FEI contact force evaluation.

A.

Introduction:

The SG functions as a heat exchanger, by means of which the high temperature pressurized radioactive primary water on the inside of the tubes heats up the non-radioactive secondary water on the outside of the tubes, in order to generate the steam that turns the turbine which in turn generates electricity.

In addition to providing a barrier (Reactor Coolant Pressure Boundary) to radioactivity and producing steam, a steam generator has many other important functions.

It is the major component in the plant that contributes to safety during transients and/or accidents.

A steam generator provides the driving force for natural circulation and-facilitates heat removal from the reactor core during a wide range of loss of coolant accidents.

Proper steam generator operation is of major safety significance and therefore any adverse changes to its design and operation may have significant safety consequences.

A review of operating history by AREVA and MHI Steam Generators shows that void fractions have to be s 98.5% to prevent formation of areas of high dry steam (in-plane fluid elastic instability (FE I)) in nuclear steam generators.

A review of published literature indicates that 1.5% percent of water in the U-tube bundle steam-water mixture is required for nucleate boiling to occur to prevent FE I. According to Dr. Pettigrew, the steam-mixture velocities in U-tube bundle have to be s 20 feet/sec to prevent plane FEI and excessive flow-induced vibrations.

The above parameters are achieved by operating steam generators with circulations ratios > 4 and steam pressures

> 900 psi (e.g., Palo Verde Replacement Steam Generators (RSGs) are the largest steam generators in the world, had similar design changes (except AVB design and tube pitch to tube/diameter ratio) as San Onofre, but have not suffered FEI in 10 years due to very high steam flows because of steam generator operation at 1039 psi and circulations ratios> 4. B. Operational Differences between SONGS Units 2 and 3: The undersigned would honestly and conservatively make an error on public safety side (Unless proven wrong by NRC Commission and NRC OIG). Therefore, the undersigned certifies that based on review of SONGS SGM Procedure (Attached) and discussions with the SONGS Unit 3 Root Cause Team Contractor, that due to higher Unit 2 SG pressure (Range 892-942 psi) compared with Unit 3 (833 psi), lower Unit 2 RCS flows (75.76 Million Lbs/hour) compared Unit 3 RCS flows (79.79 Million Lbs/hour), FEI did not occur in Unit 2. SONGS SGM Procedure

-Figure 8b: (U2) Steam Pressure -942 psi -Blue Curve (Page 95) 7 SONGS SGM Procedure

-Figure 8b: (U3) Steam Pressure-833 psi -Blue Curve (Page 96) SONGS SGM Procedure, Coolant flow rate, each: (U2) 75.76 x 106 lb/hr; (U3) 79.79 x 1 0 6 lb/hr (Page 8) SONGS SGM Procedure, Steam pressure1: (U2) 892 psia; (U3) 833 psia (Page 9) NRC AIT Report, SCE, Westinghouse, MHI and AREVA conclusions on Unit 2 FEI and operational differences between SONGS Units 2 and 3 are as follows: 8.1 -NRC AIT Report Conclusions (Pages 22 & 23): Operational Differences:

The team performed a number of different thermal hydraulic analysis of Units 2 and 3 steam generators.

The output of the various analyses runs where then compared and reviewed to determine if those differences could have contributed to the significant change in steam generator tube wear. It was noted that Unit 3 ran with slightly higher primary temperatures, about 4°F higher than Unit 2. Other differences were noted in steam and feedwater flow but none of the differences were considered sufficient to significantly affect thermal hydraulic characteristics inside the steam generators.

The different analyses included:

• Lower bounding thermal hydraulic analysis using the steam generator base design condition, where primary inlet temperature was 598°F, and an upper bound case where primary inlet temperature was 611 oF as identified in Mitsubishi Document L5-04GA021, Revision 3

• Varying steam generator pressures from 833 to 942 psia

• Steam mass flow rates from 7.59 to 7.62 Mlbm/hr

• Primary loop volumetric flow rate from 102,000 to 104,000 gpm, and

• Recirculation ratio from 3.2 to 3.5. 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.

8.2-Westinghouse Operational Assessment (Page 4, Attachment 6, Appendix D, SONGS Unit 2 Return to Service Report) : 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. There is evidence of proximity in these tubes from pre-service inspection results. Hence, the tube-to-tube wear is most likely a result of out-of-plane vibration of the two tubes in close proximity to the level of contact during operation.

8

• Westinghouse Operational Data for Units 2 & 3, Table 2-7. Summary of ATHOS Results (Page 37) o RCS Flow for Units 2 & 3-79.79 Mlbs/hr o SG Pressure for Units 2 & 3-837.6 psi o Void Fraction for Units 2 & 3-99.6% o Circulation Ratio for Units 2 & 3-3.26 B.3 -AREVA Operational Assessment (Page 15, Attachment 6, Appendix B, SONGS Unit 2 Return to Service Report): Given identical designs, Unit 2 must be judged, a priori, as susceptible to the same TTW degradation mechanism as Unit 3 where 8 tubes failed structural integrity requirements after 11 months of operation.

Indeed, the location and orientation of the two shallow TTW indications in Unit 2 are consistent with the behavior observed in Unit 3 and indicates that in-plane fluid-elastic instability in Unit 2 began shortly before the end of cycle 16 operation after 22 months of operation.

It should be noted that this statement is contested by a viewpoint that TTW in Unit 2 is simply a consequence of tubes being in very close proximity to one another with limiting wear produced by a combination of turbulence and out-of-plane fluid-elastic excitation.

This viewpoint has been evaluated completely and is considered to be arguable but not definitive.

The argument that incipient in-plane fluid-elastic has developed in Unit 2 is considered a more logical explanation for the observed TTW but again cannot be stated as definitive.

It is ultimately a moot point since the observations in Unit 3 make TTW via in-plane fluid-elastic instability a potential degradation mechanism for Unit 2. 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.

8.4 -SCE Enclosure 2 (Page 30, Unit 2 Return to Service Report): 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. After the RCE for TTW was prepared, WEC performed analysis of Unit 2 ECT data and concluded TTW was caused by the close proximity of these two tubes during initial operation of the RSGs. With close proximity, normal vibration of the tubes produced the wear at the point of contact. With proximity as the cause, during operation the tubes wear until they are no longer in contact, a condition known as 'wear arrest'.

• Root Cause Evaluation:

Unit 3 Steam Generator Tube Leak and Tube-to-Tube Wear Condition Report: 201836127, Revision 0, 5/7/2012, San Onofre Nuclear Generating Station (SONGS), Page 37 (Attached) o Reactor Coolant Flow (at cold leg temperature), 209,880 gpm o Secondary Side Operating Pressure (@1 00% power), 833 psia 8.5-February 7, 2013 NRC Commission Meeting (Attached-pages 70-75): Honorable NRC Commissioner Mr. Apostolakis was totally confused on Unit 2 FEI inconsistent statements by SCE, Westinghouse and AREVA. 9 C. Contact Force Differences between SONGS Units 2 and 3: NRC AIT, SCE and MHI state that supports were better in Unit 2, so no tube-to-tube wear occurred in Unit 2. Fabrication differences during manufacture of SONGS RSGs caused difference of contact forces in supports between Units 2 & 3. Let us now examine that whether insufficient contact tube-to AVB forces in the Unit 3 upper tube bundle caused elastic instability" which was a significant contributor to the tube-to-tube wear resulting in the tube leak. C.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-A VB 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-A VB gap sizes and decrease of tube-to-A VB 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 FE I, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors)." C.2.-AREVA states-"The primary source of tube-to-A VB 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." C.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 10 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 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/A VB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism.

None were extensively treated in the SCE root cause evaluation." C.4 -John Large States, "Causes of Tube and Restraint Component Motion and Wear: My study of the various OAs leads me to the following findings and opinion that; (i) degradation of the tube restraint localities (RBs, AVBs and TSPs) occurs in the absence of fluid elastic instability (FE I) activity; (ii) TTW, acknowledged to arise from plane FEI activity, generally occurs where the AVB restraint has deteriorated at one or more localities along the length of individual tubes; (iii) the number of tube wear sites or incidences for AVB/TSP locations outstrips the TTW wear site incidences in the tube free-span locations.

I find that the 'zero-gap' 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 combination of manufacturing variations, expansion and pressurization, etc) that determines the in-plane effectiveness of the AVB; (iv) Uniquely, the SONGS RSG fluid regimes are characterized by in-plane activity, which is quite 11 contrary to experience of other SGs used in similar nuclear power plants in which of-plane fluid phenomena dominate.

Moreover, from the remote probe inspections when the replacement steam generator (RSG) is cold and unpressurized, I consider 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., and (5) v) 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. John Large continues, "Phasing of 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 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 FE I, 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. As I have previously stated, I consider that AVB-to-tube wear is not wholly dependent upon FEI activity.

C.5 -Violette R., Pettigrew M. J. & Mureithi N. W. state (Ref. 1 -See below), "In nuclear power plant steam generators, U-tubes are very susceptible to undergo fluid elastic instability because of the high velocity of the two-phase mixture flow in the tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction.

Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction.

Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are axisymmetrically flexible.

In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability." Reference 1: Fluid-elastic instability of an array of tubes preferentially flexible in the flow direction subjected to two-phase cross flow, Violette R., Pettigrew M. J. & Mureithi N. W., 2006, http://yakari.polytechnique.fr/people/revio/masters_research_subject.html C.6 -Dr. Pettigrew (Presentation to NRC Commission, February 2013): So, you notice the U-bend-the plane of the U-bend is being installed, and on top of the U-bends are 12 bars. They are anti-vibration bars. And so you can see here that from the point of view of out-of-plane motion, the tubes are really very well supported because you have a large number of bars all around; but from the point of view of in-plane motion, there's really no positive restraint here to prevent the tube to move in the in-plane direction.

Essentially, it relies on friction forces to limit the vibration.

C.?-Contact Force Definition:

Contact force is the force in which an object comes in contact with another object. Some everyday examples where contact forces are at work are pushing a car up a hill, kicking a ball, or pushing a desk across a room. In the first and third cases the force is continuously applied, while in the second case the force is delivered in a short impulse. The most common instances of contact force include friction, normal force, and tension. Contact force may also be described as the push experienced when two objects are pressed together.

The MHI-designed AVBs had zero contact forces in Unit 3 to prevent in-plane fluid elastic instability and subsequently, wear occurred under localized thermal-hydraulic conditions of high steam quality (void fraction) and high flow velocity.

Large u-bends were moving with large amplitudes in the in-plane direction without any contact forces imposed by the out-of-plane restraints.

The in-plane vibration associated with the wear observed in the Unit 3 RSGs occurred because all of the out-of-plane AVB supports were inactive by design in the in-plane direction.

The Unit 3 tube-to-A VB contact force for the tubes with tube-to-tube wear (TTW) was zero. That is why they did not restrain the tubes in the in-plane direction.

C.8 -Contact Force

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.

I agree with MHI that 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-A VB 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/AVBfTSP 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 area of Unit 3 wear as confirmed by ECT and visual inspections 13 Wright, Darlene From: Sent: To:

Subject:

Vinod Arora [vinnie48in@gmail.com]

Tuesday, May 21, 2013 11:02 PM CHAIRMAN Resource; CMRAPOSTOLAKIS Resource; CMRMAGWOOD Resource; CMROSTENDORFF Resource; CMRSVINICKI Resource; Benney, Brian; Borchardt, Bill; Leeds, Eric; Ryan; Hall, Randy; August, JW; morgan.lee@utsandiego.com; Julie Reeder; Howell, Art; Kendra Ulrich; arnie@fairewinds.com; Ace Hoffman; Joe Hopenfeld; Freedhoff, Michal; Grant_Cope@epw.senate.gov; Walls, William; Bart Ziegler; McGregor, Ellen; Blacher, Mitch; sidneybinder@gmail.com; San Clemente Green; Donna Gilmore; Ray Lutl; Capt D Should SCE/NRC Staff and Attorneys Jointly Appeal against ASLB Ruling To Restricted NRC Commission in a Nationwide Webcast? Would they do it? Truth & Law should prevail? Perplexing Question is why would the NRC Commission rule against their own board. This matter should be referred to the US Congress/US Supreme Court not NRC Commission.

Does NRC Commission and NRC Staff have a possible vested interest or similar thinking as SCE? That is why SCE/NRC Staff and Attorneys probably would not Jointly Appeal against ASLB Ruling To Restricted NRC Commission because of public backlash.

President just fired the IRS Commissioner.

He does not want to face 8.4 Million Southern Californians because ofNRC Commission's Mistakes or perception of vested interests.

NEWS: The chairman of the Nuclear Regulatory Commission is asking for patience amid calls for public hearings on plans to restart the San Onofre nuclear reactor. Allison Macfarlane, the top U.S. nuclear safety regulator, issued a written statement Monday that says she and four fellow nuclear commissioners are restricted in their ability to comment about the possibility of public hearings related to the restart of the plant until at least June 7. That is the deadline for appealing a decision calling for the opportunity for court-like hearings on the restart. "The NRC understands that there is significant public interest in the opportunity for a hearing, and we will provide more information about the Atomic Safety and Licensing Board decision after further review of that decision and after the appeal period has elapsed," Macfarlane said. Federal regulators have indefinitely delayed a decision on the proposed restart of the shuttered San Onofre nuclear power plant. Deadlines for a final decision were removed at least a week ago from the nuclear commission's webpage on the San Onofre outage, now in its 16th month. The Atomic Licensing and Safety Board found that the destructive vibrations among steam generator tubes that have sidelined San Onofre are not accounted for in the plant's official safety blueprint, known as the updated Final Safety Analysis Report. An evaluation by nuclear commission staff of plans to restart the plant at partial power amounts to an amendment of the operating rules -creating the opportunity for public hearings, the board found. Plant operator Southern California Edison or nuclear commission staff can file an appeal that might bring the matter before Macfarlane and her colleagues.

Neither party has said whether they will. 1 CHAIRMAN Resource From: Sent:

Subject:

Vinod Arora [vinnie48in@gmail.com]

Friday, May 24, 2013 12:27 AM Dangers of San Onofre Unit 2 Restart-Updated Version -Submitted for NRC Staff Kind Review-For Information Only-No Response Required San Onofre Sad Saga Continued-NRC/SCE/MID/SCE Experts/CPUC and Public Awareness Series. Excuse me for the formatting, misspellings or grammatical errors.

Reference:

Nuclear Regulatory Commission

[Docket No. 50-361; NRC-2013-00701, Application and Amendment to Facility Operating License Involving Proposed No Significant Hazards Consideration Determination; San Onofre Nuclear Generating Station, Unit 2] I. Because of operational differences between Units 2 (Steam Pressure 942 psi, RCS Flows-74 MLbs/Hr.)

& 3 (Steam Pressure 833 psi, RCS Flows-76 MLbs/Hr.), FEI did not occur in Unit 2 (out-of-plane vibrations and/or may be in-plane vibrations existed far below the level ofFEI to cause NO tube-to-tube wear, existing Unit 2 imprecisely manufactured AVBs with contact force of2N could not have prevented FEI-Contradicted by AREVA OA and John Large). This finding is consistent with Westinghouse OA. Unit 3 FEI occurred in 4% area of the tubes in the Hot leg due to high steam flows (SG Heat Transfer Coefficient Exceeded by 5 MWt, Change from Nucleate Boiling to Film Boiling), high in-plane fluid velocities (35-50 feet/sec), low tube clearances (0.05-0.25 inches), extremely tall tubes, low steam pressures, high RCS flows and Mitsubishi Flowering Effects (Tubes moving with high in-plane velocities between 35-50 feet/second and undetermined impact forces pushed the perfectly designed out-of-the plane AVBs out of their way due to unconvincing SCE/MHI/NRC AIT highly speculative low contact forces of< IN and increased the tube-to A VB Gaps in Unit 3 compared to Unit 2 as measured by ECT). SONGS Original Combustion Engineering steam generators were operated at a void fraction of96.1%, fluid velocities of22 feet/sec and steam pressures of900 psi, and a circulation ratio of3.3. That is why FEI did not happen in Original San Onofre Units 2 & 3 for 28 years, but, these generators did suffer from flow-induced random vibrations.

According to Dr. Pettigrew, for optimum steam generator operation, operations and design engineers are advised to keep fluid velocities

< 20 feet/sec to avoid FEI and flow-induced random vibrations in nuclear steam generators.

2. MHI calculated a fatigue stress of-4 ksi on tubes due to in-plane vibrations based on wrong assumptions and hideous data. MHI fatigue calculations are significantly flawed based on calculations performed by Dr. Joram Hopenfeld and submitted to CPUC and NRC Office oflnspector General. The correct value of the fatigue stress is> 16 ksi, which exceeds the ASME Limit of 13 ksi. Review of 170,000 San Onofre Tube Inspections indicates that SCE and its vendors have not used the latest technology probes used by the Canadian and Finland Engineers for detection of incubating and circumferential cracks. These cracks can cause instantaneous tube ruptures during SONGS Unit 2 normal 70% steady state power (at any time in the 5 month operation), anticipated operational occurrences, inadvertent equipment manipulations and Design Basis Accidents.

Due to the amount of abnormal and unprecedented degradation reported in thousands of Unit 2 tubes and inadequacy of Unit 2 AVBs to prevent FEI, inspections beyond the current NEI Steam Generator Management Program are required to assure adequate protection of health and safety of8.4 Million Southern Californians and minimize Environmental, Ecological and Economic Damage from potential nuclear accidents.

The following types of scenarios are possible to inflict the above damage: A. Spontaneous fretting fatigue rupture of a single steam generator tube in the free span with a stuck open relief valve or a broken header B. Tube Ruptures from Unplanned closing of an isolation valve. C. Seismically -Induced Tube Rupture D. Station Blackout, SBO E. Main Steam Line Break, MSLB From any tube rupture and leakages, concurrent with containment bypass, these events might cause offsite radiation doses in excess of 10 CFR Part 100 as evaluated in the SONGS FSAR. Any of these two events would cause a simultaneous reactor, turbine, feedwater and reactor coolant trips. Due to feedwater pump trip, the RSG U-bundle secondary water level will shrink and tubes will be uncovered for a period of at least 10 minutes and experience a sharp drop in secondary side pressure.

The entire sub-cooled feedwater inventory contained in the faulted RSG will instantaneously flash to high dry steam. The combination of resonant, out-of-plane, plane vibrations, jet impingement forces, broken tube fragments and RSG debris will cause large axial, bending, dynamic and cyclic loads on all the tubes, tube support plates, retainer bars and anti-vibration structure.

The strength of the welded and mechanical connections of these low frequency retainer bars, retaining bars and bridges have not been analyzed for the effects of these cumulative loads to prevent A VB structure displacement, deformation or collapse during anticipated operational transients and main steam line breaks. The displacement, deformation or collapse of A VB structure along with the large axial, bending, dynamic and cyclic loads can potentially cause thousands of worn, cracked, plugged and stabilized tubes to exceed several times the allowed tube ASME Endurance 1

Limit of 13.6 ksi. If this happens, multiple tube ruptures will occur at tube-support plates, mid-spans, free spans and vibration bar notched interfaces.

Since all the water from the RSG would escape to the environment, the iodine-131 from partitioned reactor coolant leaking out the rupture tubes will also escape to the environment in less than I 0 minutes with 60 tons of radioactive coolant. Consistent with Fukushima Task Force Lessons Learnt and NRC Commissioner Meeting Transcripts, this event will be considered as a beyond design basis event, and SONGS Operators will be unable to take any timely mitigation actions to stop a severe nuclear accident in progress.

If the prevailing winds are towards San Clemente, consistent with NRC Inspector General Reports, NRC Studies and observations of SONGS Emergency Plan Drills for the last six years, SCE and Offsite agencies would not have time to respond, notify, evacuate, shelter or give Potassium Iodide to the affected residents within the 10-mile affected emergency planning zone. The casualties, and short, long-term cancer affects to the affected population will depend upon the iodine spiking factor and the duration ofblowdown, but will significantly exceed the NRC approved SONGS Control Room limit of5 Rem Total Effective Dose Equivalent (TEDE), and the Exclusion Area Boundary and Low Population Zone limit of2.5 Rem TEDE 3. MHI tube-to-A VB contact forces to prevent FEI and reduce flow-induced random vibrations based on ECT results, Visual Inspections, Quarter Bundle Model, Statistical Analysis, Manufacturing Dispersions, AVB Twist Forces Testing and New Vibration Test Data range from 2 N to > 30 N . According to Mitsubishi recent testing data, additional thicker tubes with contact forces in excess of30N are required in Unit 2 are required to prevent adverse effects ofFEI@ @100%RTP.

Best on the best available evidence, existing Unit 2 A VBs have a significant smaller contact force (2N) than 30N required to prevent FEI. This data appears to be contradicting and significantly flawed. NRC needs to question MHI and AREV A to determine correct tube-to-A VB contact force number to prevent FEI with tube-bundle uncovered and depressurized during a potential MSLB with Unit 2 at 70% power? 4. The in-plane critical velocities based on latest 2011 research papers, Dr. Pettigrew's and Dr. Mureithi Testing and MHI Root Cause Data range between 35-50 feet/sec.

5. AREVA, Westinghouse, MHI and SCE conclusions on Unit 2 FEI are conflicting, contradicting, smoking mirrors and ambiguous based on a review of SCE Unit 2 return to Service Reports and NRC Commissioners Transcripts.

6. SCE, NRC AIT, Westinghouse, AREVA, MHI and Intertek have not addressed the combined effects of tube-to-tube wear, circumferential and incubating cracks caused in tubes due to tube-to-tube wear and high cycle metal fatigue caused by fluid elastic instability.

One European Nuclear Site experienced 3 tube leaks between 2004-2006 due to fluid elastic instability and high cycle fatigue. San Onofre Unit 3 did. San Onofre Unit 2 one tube had 90% wear, not disclosed to the public. Now Unit 2? 7. Based on Unit 3 tube leak and MSLB in-situ testing, SCE has not addressed the effects of fluid elastic instability on multiple SGTRs concurrent with a MSLB in the Updated UFSAR, IOCFR 50.59 and proposed IOCFR 50.92 No Significant Hazards Analysis License Amendment.

Operating Unit 2 degraded RSGs @70% power due to the above described potential accidents results in multiple SGTRs due to FEI and incubating cracks and the consequences are as follows: A. The Proposed License Amendment Would Involve a Significant Increase in the Probability or Consequence of an Accident Previously Evaluated.

B. The Proposed License Amendment Would Involve the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated.

C. The Proposed License Amendment Would Involve a Significant Reduction in a Margin of Safety. In order to issue a finding of no significant hazards considerations, the NRC Staff bears the burden of showing that the hazards considerations as a result of the ASLB 's recent decision in the CAL proceeding are insignificant.

The Staff cannot make that showing, and consequently the proposed finding must be withdrawn and a hearing on the proposed license amendment held by an ASLB before the amendment may be approved by the NRC. As the ASLB recently held with respect to San Onofre Unit 2: "We conclude that until the tube degradation mechanism is fully understood, until reasonable assurance of safe operation of the replacement steam generators is demonstrated, and until there has been a rigorous NRC Staff review appropriate for a licensing action, the operation of Unit 2 would be outside the scope of its operating license because the replacement steam generator design must be considered to be inconsistent with the steam generator design specifications assumed in the FSAR and supporting analysis." May 23, 2013: California Democratic Sen. Barbara Boxer told Nuclear Regulatory Chairwoman Dr. Allison Macfarlane that she wants two things before the restart of the San Onofre Nuclear Generating Station is considered:

completion of an investigation and a public hearing. Boxer made her comments Thursday as part of the Senate reconfirmation hearing for Macfarlane, who is seeking another term as chair of the NRC. Boxer scoffed at Southern California Edison's (SCE) plan to change its operating license to restart the number two reactor at San Onofre at partial capacity.

She repeated the U.S. Atomic Safety Licensing Board's description of the plan as an "experiment." Boxer commented on SCE's plan to operate the reactor at 70 percent. "We'll see what happens, we'll see how it goes," Boxer said during the hearing. "That's like saying I think I fixed the damaged brakes on your car, but don't drive it over 40 miles per hour." 2 Boxer repeatedly brought up the 8 million people Jiving within 50 miles of the nuclear plant, saying if someone came to the NRC today and asked for a license to operate a nuclear power plant at that site, "in a seismic and a tsunami zone, we all know every single commissioner would say, 'don't you think you could find a better place for it?'" 8. SCE has not addressed the True Root Cause of Unit 3 tube-to-tube wear (Untested and unanalyzed design changes, adverse operational and thermal-hydraulic parameters, human performance errors (Avoidance of 1 OCFR 50.90 by portraying RSGs as "like for like" replacement, rejecting SCE/MHI A VB Team proposed changes to reduce void fractions by improving circulations ratios, failure to review ofFEI research papers by Dr. Pettigrew, Dr. Ivan Cotton, Dr. Dhir, etc., failure to benchmark other successful CE replacement generators (Palo Verde)) and actions to prevent tube-to-tube wear in Unit 2 as required by CAL. 9. [Redacted]

tube wear rates calculations appear to be non-conservative and based on old SGs testing, which significantly differ in design compared with San Onofre RSGs. 3