ML13153A006

LTR-13-0469 - Don Leichtling Email Operational Difference Between San Onofre Nuclear Generating Station (SONGS) Unit 2

ML13153A006
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Site:	San Onofre
Issue date:	05/29/2013
From:	Leichtling D Public Commenter
To:	Macfarlane A NRC/Chairman
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LTR-13-0469
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Joosten, Sandy From: Capt.D [captddd@gmail.com)

Sent: Wednesday, May 29, 2013 6:16PM To: Capt D

Subject:

Operational Difference Between SONGS Units 2 and 3 There is simply no scientific basis for a no-significant hazards consideration determination in the case of the proposed license amendment for San Onofre Unit 2. (24 Pages)

Operational Difference Between SONGS Units - May 29, 2013

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]

Most of the steam generators operate at void fractions below 98.5%, steam pressures> 900 psi and recirculation ratios >4. This ensures operation of the SG in the nucleate boiling regime (damping of hot SG tubes to prevent FEI), optimum operation of the SG performance and ensures the prevention of adverse effects of FEI (High Dry Steam, high fluid in-plane velocities, Film Boiling), flow-induced random vibrations and excessive dynamic pressures on tube-to-tube wear, tube-to-AVB/TSP wear and retainer bar-to-tube wear. Along with numerous untested and unapproved design changes made under the false pretense of "like for like" to avoid a lengthy NRC 10CFR 50.90 Review and Public Hearings, SCE designed and MHI fabricated 21~ Century Safest and Innovative Replacement Steam Generators, which were then operated outside the above operational parameters to maximize both the SG thermal output and profits. MHI Root Cause states, "Thus, not using ATHOS, which predicts higher void fractions than FIT-Ill at the time of design represented, at most, a missed opportunity to take further design steps, not directed at in-plane FEI, that might have resulted in a different design that might have avoided in-plane FEI. However, the AVB Design Team recognized that the design for the SONGS RSGs resulted in higher steam quality (void fraction) than previous designs and had considered making changes to the design to reduce the void fraction (e.g., using a larger downcomer, using larger flow slot design for the tube support plates, and even removing a TSP). But each of the considered changes had unacceptable consequences and the AVB Design Team agreed not to implement them. Among the difficulties associated with the potential changes was the possibility that making them could impede the ability to justify the RSG design under the provisions of 10 C.F.R. §50.59. Thus, one cannot say that use of a different code than FIT-Ill would have prevented the occurrence of the in-plane FEI observed in the SONGs RSGs or that any feasible design changes arising from the use of a different code would have reduced the void fraction sufficiently to avoid tube-to-tube wear. For the same reason, an analysis of the cumulative effects of the design changes including the departures from the OSG's design and MHI's previously successful designs would not have resulted in a design change that directly addressed in-plane FE I." We now know the end result of that SCE/MHI missed opportunity, destruction of a $1 Billion Dollars in Steam Generators and the Number 1 US Public Safety Concern/Nuclear Scandal, Controversy and Cover-up involving NRC, CPUC, SCE, MHI, Westinghouse, AREVA, lntertek and others. Dr.

Pettigrew, the Worlds authority on steam generators, told Dr. Macfarlane and the other NRC Commissioners that MHI's AVBs simply do not provide a positive restraint against FEI.

Here is a summary of San Onofre Tube-to-AVB Contact Forces and Accident Scenarios for your benefit:

Both Units had circulation ratios of 3.3, narrow tube to pitch tube diameter, excessive number of tubes (9,727), extremely tall tubes (average length of heated tube increased by 50 inches, equivalent to the addition of -650 tubes), 116,000 square feet of tube heat transfer area (increased from 104,000 in the OSGs), and virtually no in-plane restraints.

1. Unit 3 had higher void fractions of 99.6%, high steam flows (film boiling), higher thermal reactor power per RSG (RCS Flows, 76 Million lbs./hr, 1737 MWt plus), higher in-plane fluid velocities (35-50 feet/sec), steam generator operation at too low a pressure (833 psi), insufficient tube-to-AVB contact forces (< 1N per AREVA) and loose supports (larger tube-to-AVB Gaps, Based on ECT Results). These differences caused FEI, Flow-Induced Random Vibrations and Mitsubishi Flowering Effect in Unit 3 @100%RTP. The flow regime in Unit 3 changed from nucleate boiling to film boiling because the Unit 3 RSG heat transfer coefficient was exceeded and the change was attributed to more than 5MWt of SG output in Unit 3 in 4% of the SG U-tube bundle high region of wear on the hot-leg side due to higher RCS flows and lower SG Pressure Operation in Unit 3 RSGs.

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2. Unit 2 had moderate Void fractions of (98-98.9%), lower steam flows (nucleate boiling), lower thermal reactor power per RSG (lower RCS Flows, 74 Million lbs./hr versus 76 Million lbs./hr for Unit 3, lower heat output, 1727 MWt plus versus 1737 plus for Unit 3), high out of-plane fluid velocities (25 feet/sec), steam generator operation at 942 psi (consistent with NRC AIT Report and SONGS SGM procedure) and questionable tube-to-AVB contact forces(< 2N per AREVA) and better supports (smaller tube-to-AVB Gaps, based on ECT results). These conditions caused Flow-Induced Random Vibrations and Mitsubishi Flowering Effect in Unit 2 @1 OO%RTP.

3. Even though SCE, MHI and AREVA claim that operating and thermal-hydraulic conditions were the same in both units, Unit 2 did not experience tube-to-tube wear because of double tube-to-AVB contact forces and better supports because of inadvertent accidental Unit 2 AVB design. FEI did not occur in Unit 2, which is consistent with Westinghouse report. The NRC AIT Report noted that the operational differences did not make any difference between Units 2 & 3. Throughout this entire paper, we will review SCE, MHI and AREVA claims: (1) About Unit 2 double tube-to-AVB contact forces and better supports because of inadvertent accidental design, and (2) About Unit 3 insufficient tube-to-AVB contact forces and loose supports because of intentional precision manufacturing.

4. According to MHI/ AREVA, a Tube-to-AVB Contact Force of 10N is required to prevent FEI@100%RTP. It is noted that Tube-to-AVB clearances are significantly larger than the SONGS steam generator design clearance of 2 mils diametral. For the present, it is sufficient to note that while the forces at AVB locations needed to prevent the onset of fluid-elastic instability are low, after 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. Calculation of the probability of the onset of in-plane fluid-elastic instability requires information in three areas: stability ratios, contact forces at AVB locations and a criteria for deciding whether AVB supports are effective or ineffective in terms of in-plane support. Stability ratios need to be known as a function of position in the bundle, number of consecutive ineffective supports and power level. Contact forces at AVB locations cannot be determined deterministically since the dispersion of gaps between tubes and AVB supports is random, and thus probabilistic in nature. The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are not 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. MHI has calculated the response of a large U-bend with AVB supports subjected to turbulence and fluid-elastic excitation forces. Various gap (clearance) conditions were included along with contact forces ranging from 1N to 10N. An equal contact force was applied at all12 AVB locations. Given the uncertain nature of fluid-elastic excitation forces, a direct application of the selected excitation function to SONGS at 100% power is problematic. However the scale of the contact force that prevented in-plane vibration is highly useful. A contact force of 1N did not resist in-plane motion but a force of 1ON was completely effective.

5. According to Mitsubishi recent testing data, additional thicker tubes with contact forces in excess of 30N are required in Unit 2 in order to prevent adverse effects of FEI @1 OO%RTP.

6. Based on the best available evidence, existing Unit 2 AVBs have a significantly smaller contact force (2N) than the 30N required in order to prevent FE I.

7. During AOO and MSLB events, Unit 2 at 70% power will experience void fractions of 100%, high steam flows (film boiling),

high in-plane fluid velocities (35-50 feet/sec) and jet impingement from flashing feedwater. With contact forces of 2N, Unit 2 tube bundle would not be able to prevent the adverse effects of FEI, Flow-Induced Random Vibrations and Mitsubishi Flowering Effect. Multiple tube-ruptures can occur due to tube-tube wear, full circumferential rupture of tubes can occur due to incubating cracks and the entire degraded Anti-vibration structure can collapse. The following accident scenarios are considered:

Accident Scenarios (Courtesy of Dr. Joram Hopenfeld, Former NRC Staff)

The Steam Generators in San Onofre reactor Unit 2 (SGE 88 and SGE 89) contain 482 and 563 tubes respectively; with AVB wear ranging from 10% to 34%. The two SGs also contain a total of 515-plugged tubes. These tubes act as multiple sources for leakage during normal operations and during accidents. They must be considered as sources for causing accidents and for propagating the leakage intensity during the accident. An assessment of operations with such degraded tubes must demonstrate that at any time during normal operations and during accidents their local in-plane gap velocities (35-50 feet/sec), the corresponding SR and the burst pressure, will remain at sufficiently low levels to prevent leakages from exceeding acceptable amounts.

The following are examples of accidents that must be included in such assessments.

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.

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Steam Generator overfill occurs relatively frequently in PWRs. An assessment should consider that the DBA SGTR will cause the relief valve to be stuck open during this event. The resulting higher local gap velocities (35-50 feet/second) and the corresponding increase in the SR must not cause additional tubes (both plugged and un-plugged) to rupture.

B. Tube Ruptures from Unplanned closing of an isolation valve.

Closing an isolation valve would lead to an increase in steam flow through the unaffected SG. The corresponding increase in gap velocity would increase the local SR causing tubes, which are close to exhausting their fatigue life to rupture abruptly. This accident is similar to case A above with the exception that the increase in SR is expected to take place at a slower rate.

C. Seismically-Induced Tube Rupture Both plugged and unplugged tubes can potentially lead to large primary to secondary leakage. Plugged tubes would behave differently, firstly because they do not generate a failure signal at the steam ejectors, and secondly, because the natural frequency of a broken tube would be lower than that of an in-service tube.

Reactor experience has demonstrated that tubes that have been plugged due to wear will continue to wear and eventually break, which would then allow them to impact and damage adjacent tubes. Material loss by wear not the mode of failure at plants was studied by EPRI. In their studies combining tube swelling with Fluid Induced Vibration (FIV) led to circumferential fatigue failure. The difference between the cases studied by EPRI and the plugged tubes at SONGS is that at SONGS some plugged tube have already suffered considerable fatigue damage prior to plugging and are prone to fatigue failure. In this regard, EPRI recommends that tubes with pre-existing circumferential cracks be evaluated using linear elastic fracture mechanics. Because some tubes at SONGS used up a significant fraction of their fatigue life they may contain micro cracks of various size. Because such cracks have not been detected at SONGS there is no indication that they do not exist. SCE and the NRC have not address this issue.

EPRI did not assess the effectiveness of tube stabilization in preventing damage to adjacent tubes; neither did SCE provide any information on their criteria for selecting tubes for stabilization.

SCE conclusions that the combined forces of the differential pressure and the seismic loads would not cause any tube to burst cannot be justified when the tubes are also subjected to cyclic loads simultaneously. SCE calculations are based on the tensile strength that would cause tube rupture, a much lower stress, less than half, would be sufficient to sever tubes with cumulative fatigue usage (CUF) near unity (Ref 8).

SCE calculations are based on a non-conservative model and therefore their conclusions in the FSAR (5.4.2.2.1.3) regarding the ability of degraded tubes to withstand seismic loads anfnot valid.

D. Station Blackout, SBO Severe accidents are not considered design basis accidents, nevertheless when changes in system operations are contemplated those changes must not increase safety risk. The operation of San Onofre reactor Unit 2 with a large number of fatigued tubes *represents a new accident that has never been previously analyzed.' All the analysis to date was based on tube failure by creep at high temperature. The fact that the tubes were fatigue-damaged demonstrates they can fail earlier due to natural flow instabilities in the steam generator. The SBO accident is briefly described below.

In this accident the primary system remains pressurized following the core becoming uncovered. After the core is uncovered the secondary sides of all steam generators are dry while on the primary side, steam flows by natural convection from the core to the steam generators and back to the core. The high pressure, high temperature steam will cause the weakest component in the system to fail thereby depressurizing the primary side. In this regard the hot leg surge line and the SG tubes are the weakest components in the reactor coolant system. If the high hoop stress on the hot leg surge line causes it to fail, the release of the highly radioactive gases will be contained within the containment. If on the other hand, the high pressure, high temperature steam opens up existing cracks in the steam generator tubes or ruptures the tubes, the primary side will be depressurized, by-passing the containment and allowing the highly radioactive gases to escape directly to the environment through the SG relief valve. The above scenario, also known as the high/dry core damage sequence, represents an early containment failure, which significantly increases the large early release frequency (LERF). When the containment fails early, the release to the environment is several thousand times larger in comparison to the release when the containment is intact. Most importantly, this early release occurs prior to the evacuation of the close population and therefore may cause early health effects (prompt fatalities).

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Conformance to 10 CFR 50, Appendix B Criterion 16 dictates that operation with fatigued tubes will not increase the probability that fatigued tubes will fail before the surge line. Appendix B dictates that to maintain its licensing basis the licensees must provide measures to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective materials and equipment, are promptly identified and corrected. Fatigued tubes definitely represent conditions, which are adverse to quality.

E. Main Steam Line Break, MSLB The fact that San Onofre reactor Unit 2 can pass the existing performance criteria from the in-situ tests results of San Onofre reactor Unit 3 provides no assurance at all that during a spontaneous MSLB accident the leakage will not exceed the DBA leakage. The in-situ tests only show that if the tubes were only exposed to tested pressure they would not leak if they maintained their wall geometry as tested. The in-situ tests were intended to determine leakage on the basis of tube weakening by actual loss of material and inclusions of stress corrosion cracks. In contrast to static pressure tests, fatigue failure due to high cycle FIV would result in a fast propagating circumferential crack at relatively low stresses. Leakage from degraded tubes must be assessed in terms of the mechanism that has the potential to cause the largest leakage.

If SCE wants to base their calculations on a realistic accident scenario, it must first demonstrate that the wear equation that was developed for laboratory data would be applicable to a tube that experienced impact wear in the SONGS steam generators. As discussed in Appendix A, the wear equation which was used by SCE to calculate wall thickness did not properly incorporate the effects of impact wear. Secondly and more importantly, SCE must demonstrate that their burst pressure mechanism of determining leakage is conservative in comparison to the leakage that would occur during the fast MSLB depressurization.

The fast depressurization of the secondary side following an MSLB will lead to rapid increases in local gap velocity and steam quality, thereby significantly increasing the stability ratio (SR). The higher SR would, in-turn, increase the stress on the tube leading to rapid circumferential crack propagation as occurred in North Anna (Ref 2).

Summary- Tube-to-AVB Contact Forces: It is concluded that FEI and MHI Flowering effect redistributed the tube-to-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 other 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 and unproven Unit 2. Even at 70% power, under Anticipated Operational Transients and Main Steam Line Breaks, the 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 to prevent adverse effects of FEI. Best on the best available evidence, existing Unit AVBs have a significant smaller 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 flow-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 catastrophically, it appears that there is a much larger group of tubes, which is out of alignment and is in direct contact with the support plates in Unit 2. SCE, MHI, AREVA, lntertek, Westinghouse and NRC Staff are completely ignoring these logic cracks in their analyses. The difference in management of Steam Generator Tube Rupture between Finland and USA is, that no primary coolant (liquid and/or steam) releases to the environment are allowed in Finland, while in USA, primary steam releases are not forbidden, so Utilities can seek profits by conducting risky experiments with people's lives, because this situation is unique to San Onofre 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 DAB Safety Team fully endorses the Friends of the Earth (FoE) and the Natural Resources Defense Council (NRDC) request that the proposed SCE "No significant hazards" consideration determination should be withdrawn because (1) The NRC Staffs proposal exceeds the authority granted to it by the Sholly 4

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-AVB 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 1N did not resist in-plane motion but a force of 1ON was completely effective", and (2) MHI states, "Tube-to-AVB contact forces in excess of 30N will prevent in-plane tube-displacement and tube-to-tube contact in high region of wear." The number of Unit 2 tube-to-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 tube-to-AVB wear indications with less wear rates than that of Unit 3.

In the 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-AVB contact forces and better supports for prevention of FEI in Unit 2 are incomplete, inconsistent, confusing, inconclusive and are based on faulty computer simulations and/or 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-to-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, even NRC Commissioner Mr. Apostolakis was totally confused on Unit 2 FEI inconsistent statements by SCE, Westinghouse and AREVA.

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-and extremely low tube clearances (ranging from 0.05- 0.25 inches)

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 (equivalent to the addition of 650 tubes)

6. 11% increase in tube heat transfer area

7. Tubes-AVB zero gap during bundle in the hot and pressurized condition and AVBs with a contact force of>

1N (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 5

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-AVB gap is effective against the out-of-plane vibrations forces, but a contact force between 1Nand 1ON is needed to prevent in-plane vibrations. A contact force of 1N is ineffective against in-plane vibrations, but a force of 1ON is totally effective to prevent in-plane vibrations." Mitsubishi, AREVA and SCE state that the operating and thermal-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 increases tube-to-AVB contact forces in excess of 30N and will prevent in-plane tube-displacement and tube-to-tube contact in regions of high wear. MHI further says that the thicker bars will increase tube support effectiveness throughout the bundle, improve stability ratios and greatly reduce flow-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-AVB contact force to generate friction that inhibits in-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-and 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. 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-AVB 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-AVB gap and contact force between 10-30N 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 a 4% area of 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 reduction of the Unit 3 6

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 1N 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-AVB 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-AVB 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 the same as Unit 3, Unit 2 tubes only experienced tube-to-AVB wear, because the Unit 2 tube-to-AVB contact forces (Assumed 2N) were double that of Unit 3 (Assumed 1N) and prevented the tubes from hitting each other in the in-plane direction and prevented tube-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" tube wear." Westinghouse states that, "SCE has NOT done an extensive evaluation of Unit 3 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 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."

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::; 98.5% to prevent formation of areas of high dry steam (in-plane fluid elastic instability (FEI)) 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 ::; 20 feet/sec to prevent out-of-plane FEI and excessive flow-induced vibrations. The above parameters are achieved by operating steam generators with circulation 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 DAB Safety Team believes 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)

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 1O* 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 were 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:

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• 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.

B.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.

• 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 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 self-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.

B.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

• Secondary Side Operating Pressure (@100% power), 833 psia B.5- February 7, 2013 NRC Commission Meeting (Attached- pages 70-75): NRC Commissioner Mr. Apostolakis was totally confused on Unit 2 FEI inconsistent statements by SCE, Westinghouse and AREVA.

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 whether insufficient contact tube-to AVB 8

forces in the Unit 3 upper tube bundle caused "fluid-elastic instability" which was a significant contributor to the tube-to-tube wear resulting in the tube leak.

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-AVB gap in hot condition was sufficient to prevent the tube from becoming fluid-elastic unstable during operation based on the MHI experiences and contemporary practice. MHI postulated that a "zero" gap in the hot condition does not necessarily ensure that the support is active and that contact force between the tube and the AVB is required for the support to be considered active. The most likely cause of the observed tube-to-tube wear is multiple consecutive AVB supports becoming inactive during operation. This is attributed to redistribution of the tube-to-AVB-gaps under the fluid hydrodynamic pressure exerted on the tubes during operation.

This phenomenon is called by MHI, "tube bundle flowering" and is postulated to result in a spreading of the tube U-bends in the out-of-plane direction to varying degrees based on their location in the tube bundle (the hydrodynamic pressure varies within the U bend). This tube U-bend spreading causes an increase of the tube-to-AVB gap sizes and decrease of tube-to-AVB contact forces rendering the AVB supports inactive and potentially significantly contributing to tube FEI. Observations Common to BOTH Unit-2 and Unit-3: The AVBs, end caps, and retainer bars were manufactured according to the design. It was confirmed that there were no significant gaps between the AVBs and tubes, which might have contributed to excessive tube vibration because the AVBs appear to be virtually in contact with tubes. MHI states, 'The higher than typical void fraction is a result of a very large and tightly packed tube bundle, particularly in the U-bend, with high heat flux in the hot leg side. Because this high void fraction is a potentially major cause of the tube FEI, and consequently unexpected tube wear (as it affects both the flow velocity and the damping factors)."

C.2.- AREVA states- "The primary source of tube-to-AVB contact forces is the restraint provided by the retaining bars and bridges, reacting against the component dimensional dispersion of the tubes and AVBs. Contact forces are available for both cold and hot conditions. Contact forces significantly increase at normal operating temperature and pressure due to diametric expansion of the tubes and thermal growth of the AVBs. After fluid elastic instability develops, the amplitude of in-plane motion continuously increases and the forces needed to prevent in-plane motion at any given AVB location become relatively large.

Hence shortly after instability occurs, U-bends begin to swing in Mode 1 and overcome hindrance at any AVB location."

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 excitation. Hence, a tube that may vibrate in-plane (IP) would definitely be unstable OP. A small AVB gap that would be considered active in the OP mode would also be active in the IP mode because the small gap will prevent significant in-plane motion due to lack of clearance (gap) for the combined OP and IP motions. Thus, a contact force is not required to prevent significant IP motion. Manufacturing Considerations: There are several potential manufacturing considerations associated with review of the design drawings based on Westinghouse experience. The first two are related to increased proximity potential that is likely associated with the ECT evidence for proximity. Two others are associated with the AVB configuration and the additional orthogonal support structure that can interact with the first two during manufacturing.

Another relates to AVB fabrication tolerances. These potential issues include: (1) The smaller nominal in-plane spacing between large radius U-bend tubes than comparable Westinghouse experience, (2} The much larger relative shrinkage of two sides (cold leg and hot leg) of each tube that can occur within the tubesheet drilling tolerances. Differences in axial shrinkage of tube legs can change the shape of the U-bends and reduce in-plane clearances between tubes from what was installed prior to hydraulic expansion, (3) The potential for the ends of the lateral sets of AVBs (designated as side narrow and side wide on the Design Anti-Vibration Bar Assembly Drawing that are attached to the AVB support structure on the sides of the tube bundle to become displaced from their intended positions during lower shell assembly rotation, (4) The potential for the 13 orthogonal bridge structure segments that are welded to the ends of AVB end cap extensions to produce reactions inside the bundle due to weld shrinkage and added weight during bundle rotation, and (5) Control of AVB fabrication tolerances sufficient to avoid undesirable interactions within the bundle. If AVBs are not flat with no twist in the unrestrained state they can tend to spread tube columns and introduce unexpected gaps greater than nominal inside the bundle away from the fixed weld spacing. The weight of the additional support structure after installation could accentuate any of the above potential issues. There is insufficient evidence to conclude that any of the listed potential issues are directly responsible for the unexpected tube wear, but these issues could all lead to unexpected tube/AVB fit-up conditions that would support the amplitude limited fluid-elastic vibration mechanism. None were extensively treated in the SCE root cause evaluation."

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) nw, acknowledged to arise from in-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 nw 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 contrary to experience of 9

other SGs used in similar nuclear power plants in which out-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 AVB-TSP Wear -v- TTW: I reason that, overall, the tube wear process comprises two distinct phases: First, the AVB (and TSP) -to-tube contact points wear with the result that whatever level of effectiveness is in play declines. Then, with the U-bend free-span sections increased by loss of intermediate AVB restraint(s), the individual tubes in the U-bend region are rendered very susceptible to FEI induced motion and TTW. Whereas the OAs commissioned by SCE broadly agree that the wear mechanics comprises two phases, there are strong differences over the cause of the first phase comprising in-plane AVB wear: AREVA claim this is caused by in-plane FEI whereas, the contrary, Mitsubishi (and Westinghouse) favor random perturbations in the fluid flow regime to be the tube motion excitation cause. Put simply: (i) if AREVA is correct then reducing the reactor power to 70% will eliminate FEI, AVB effectiveness will cease to decline further and TTW will be arrested; however, to the contrary, (ii) if Mitsubishi is right then, even at the 70% power level, the AVB restraint effectiveness will continue to decline thereby freeing up longer free-span tube sections that are more susceptible to TTW; or that (iii) the assertion of neither party is wholly or partly correct. 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 U-tube region and also because of their low natural frequencies in their out of plane modes. In nuclear power plant steam generator design, flat bar supports have been introduced in order to restrain vibrations of the U-tubes in the out of plane direction. Since those supports are not as effective in restraining the in-plane vibrations of the tubes, there is a clear need to verify if fluid elastic instability can occur for a cluster of cylinders preferentially flexible in the flow direction. Almost all the available data about fluid elastic instability of heat exchanger tube bundles concerns tubes that are 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 :1/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 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-AVB 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. The DAB Safety Team agrees 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-AVB wear and insufficient contact forces between tubes and AVBs. Tube bundle distortion is considered a major contributing cause to the mechanism of tube-to-tube/AVB/TSP wear seen in the Unit 3 SG's. After 11 months of wear, contact forces were virtually eliminated between the tube and AVBs in the areas of highest area of Unit 3 wear as confirmed by ECT and visual inspections.

The Proposed License Amendment Presents New and/or Increased Risks That Endanger Public Health and Safety (Courtesy of The Friends of the Earth)

In considering the criteria of 10 C.F.R. § 50.92, the Staff could not determine that the proposed license amendment for San Onofre entails no significant hazards consideration. If the proposed change in the license fails to meet any one of the three criteria in 10 C.F.R. § 50.92, the NRC must withdraw the proposed no significant hazards consideration determination. As demonstrated in the technical analyses appended to these comments, and by the ASLB's recent decision on San Onofre, the proposed amendment does not satisfy any of the three criteria.

10

To assess whether the change proposed by SCE creates a significant hazards consideration, the appropriate comparison is between the operation of the unit with undamaged steam generators as assumed in SCE's current license, on the one hand, and the operation at 70% of power with damaged steam generators that Edison now proposes. SCE's rationale for concluding that no significant hazards consideration is presented is apparently based on comparing operation with undamaged tubes at 100% and 70%, completely ignoring the current highly-damaged state of the steam generators in Unit 2. As Dr. Hopenfeld states, SCE's evaluation of the § 50.92 criteria "is based on the presumption that change in power level can be discussed without giving any considerations to the physical conditions of the tubes before and after the change."

The NRC cannot, despite its best efforts, ignore the events of the past 16.5 months. Major defects causing unprecedented tube wear have been discovered in the replacement steam generators at San Onofre, and while the mechanical force that inflicted the wear has been identified as primarily in-plane FEI, neither the NRC nor the licensee has yet determined the root cause of the FEI, let alone a remedy for it. Instead, SCE and the NRC propose to simply restart Unit 2 and operate it at reduced power for one shortened test cycle as an experiment on steam generators with demonstrated design flaws in components and systems critical to the safety of San Onofre Unit 2.

i. The Proposed Finding of No Significant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve a Significant Increase in the Probability or Consequence of an Accident Previously Evaluated.

Staff addresses the first criterion of 10 C.F.R. § 50.92 by simply restating SCE's analysis, which concludes that the proposed license amendment would not involve a significant increase in the probability or consequence of an accident previously evaluated "because there is no adverse effect on plant operations or plant conditions."24 SCE relies on its response to Requests for Additional Information (RAis) 11-14 as its basis for this assertion.25 SCE, however, fails to make the appropriate comparison when applying this first criterion.

SCE's apparent position is that the relevant comparison is between the operation of fully functional undamaged steam generators originally licensed to run at 100% power and operation of those same steam generators at 70% power. That comparison is incorrect. The proper consideration is whether operating at 70% power with defective, damaged, and unrepaired steam generators involves a significant increase in the probability or consequence of an accident previously evaluated, as compared to the risk of operating at 100% of power with fully functional, undamaged steam generators. In this context, operating the steam generators in their present condition at 70% of power creates a significant increase in the probability of a release of radioactivity and in the consequences - exposure of potentially millions of people to increased radioactivity.

The three impartial experts who wrote the ASLB's recent decision on San Onofre found that operating the replacement steam generators at 70% would significantly increase the probability and consequences of a previously analyzed accident.z6 For example, the replacement steam generators can no longer meet 10 C.F.R. Part 50, App. A- General Design Criterion (GDC) 14 (Reactor Coolant Pressure Boundary), which requires "an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture." SCE's own tube-to-tube wear assessment, as the ASLB order notes, shows that "one unstable tube can drive its neighbor into instability through repeated impact events."z7 Given this condition, there is no longer "an extremely low probability" of the kind of tube failure GDC 14 is meant to guard against.

Nuclear engineers Mr. Large and Dr. Hopenfeld show in the attached declarations that the proposed amendment would involve a significant increase in the probability or consequence of an accident previously evaluated. Mr. Large explains that the excitation forces present in the steam generators exist due to pressure and temperature conditions that will not be affected by reducing the power from 100% to 70%. Thus, contrary to the assertions of SCE, operating Unit 2 at 70% of power during Cycle 17 would not reduce the forces exerted on the tubes during Cycle 16 that caused the unprecedented rapid tube wear and deterioration.29 Both of SCE's operational assessments agree that the damage will continue at an unprecedented pace, differing only between 6 months and 16 months as the remaining life expectancy of the Unit. Even at 70% of power, large numbers of tubes in the replacement steam generators will continue to wear and degrade and, as a consequence, significantly increase the probability of tube rupture.

Dr. Hopenfeld asserts that the probability and consequences of a previously considered accident are significantly increased because, in addition to the fact that operating at 70% of power will not reduce the excitation forces that cause tube wear, SCE also failed to take into account metal fatigue caused by fretting, which is brought on by the FBI-induced vibration. Tubes in Unit 2's steam generators used up a large fraction, if not all, of their allowable "fatigue life" during the last cycle of operation, Cycle 16. Dr. Hopenfeld asserts:

The number of tubes which are susceptible to rupture by fatigue during a given accident scenario must be known if one is required to predict accident consequences. Until this is done the present pressure based burst performance criteria cannot be used as a reliable indicator of risk.

As a result it must be conservatively concluded that allowing Unit 2 to operate at any power level could significantly increase the consequences of the accidents, which were evaluated by SCE and were described in the UFSAR.

SCE and its consultants have inspected the steam generators for tube surface wear and tube wall thickness but have failed to account for metal fatigue, which cannot be discerned by inspection. Technical Specification Task Force (TSTF) 449 requires SCE to evaluate additional 11

loads on the tubes that could contribute to burst or collapse, even ifthey cannot be physically measured. SCE's analysis ignores the increased probability or consequences of an accident contributed to by metal fatigue in the tubes of the steam generators.

Tube fatigue increases the probability of an accident. It also increases the consequences, because tube failure owing to metal fatigue happens more suddenly than failure owing to stress corrosion cracking (SCC). A tube failure from fatigue, such as that experienced at the North Anna Generating Station Unit I on July 15, 1987, occurs suddenly and quickly.33 In the event of a main steam line break, for example, accompanied by the rupture of five or more fatigue- weakened tubes, the operator's inability to control the loss of coolant rapidly enough would lead to a significant increase in the probability of uncovering the core, with major increases in the consequences of a previously evaluated accident, including the exposure of millions of Californians to radiation.

Dr. Hopenfeld therefore concludes that restarting the plant for another cycle would place Unit 2 outside of the bounds of accidents evaluated in the updated final safety analysis-report (UFSAR) by significantly increasing the probability and consequences of a main steam line break (MSLB) accident.3s Similarly, Mr.

Large found that a single tube burst caused by an MSLB that damages the fuel core could result in severe consequences beyond those considered in the UFSAR.

The NRC's proposed finding of no significant hazards consideration addresses none of the issues identified by Friends of the Earth's experts, as summarized above. Thus, the proposed NRC finding must be withdrawn and a hearing on the proposed license amendment held before a decision is made on the proposal.

ii. The Proposed Finding of No Significant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated.

Significantly, the UFSARfor the original steam generators for SONGS Units 2 and 3 excluded the possibility of in-plane vibrations caused by fluid elastic instability when evaluating the conditions necessary to maintain steam generator tube integrity[,] .. .[an assumption that is] '

demonstrably unjustified for the replacement steam generators.37 --- ASLB Opinion, May 13, 2013 The NRC's regulations do not allow the Staff to make a no significant hazards consideration determination if it finds that the proposed license amendment would create the possibility of a new or different kind of accident not previously evaluated. The Staff restates in the Federal Register notice proposing the license amendment the licensee's position that "the proposed changes do not require a change in any plant systems, structures, or components or the method of operating the plant other than to reduce power for the duration of Cycle 17.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated."

Edison's "therefore" is misplaced: the conclusion of the second sentence does not follow from the statement in the first. The premise of"no change" that SCE relies on for this conclusion, however, is erroneous because it ignores the change that shut the plant down more than a year ago: that an abnormally high amount of tube wear has occurred in the replacement steam generators, and, in particular, the unprecedented fretting fatigue caused by massive FEI and its impact on the steam generator tubes.

First, the UFSAR does not consider the possibility of accidents caused by tube wear from in-plane FEI because it is based on an assumption that in-plane FEI will not occur. UFSAR section 5.4.2.3.1.3, which analyzes steam generator tube integrity, is therefore inadequate and demonstrates that operation at 70% of power presents new and different kinds of accidents from those previously evaluated.

The ASLB agrees. In its recent opinion on SCE' s proposed restart plan under the CAL, the ASLB found that operating the replacement steam generators in their current degraded condition is a test or experiment as described under 10 C.F.R. § 50.59(C)(2).39 By definition then, the proposed license amendment cannot possibly meet the second criterion for a no significant hazards consideration determination. Operating at 70% for any length of time with the replacement steam generators in their current condition is an experiment, the outcome of which has not been analyzed in the UFSAR.

Second, the UFSAR currently considers only accidents resulting from excessive pressure loads, not fretting fatigue. During Cycle 16, the tubes in Unit 2's steam generators experienced fretting previously not experienced in the history of any U.S. steam generator.4o To allow the operation of the steam generators without repairs would, because of this unanalyzed fatigue, create the possibility of a new or different kind of accident from any accident previously evaluated.

Accidents caused by fretting fatigue are different from accidents caused by stress corrosion cracking (SCC). As described above, unlike sec, metal fatigue is difficult to detect through in-service inspections, and near or at the end of a tube's fatigue life cracking propagates much more quickly than SCC. There is no available data correlating field measurements to leakage from fatigued tubes during a design-basis accident.43 Thus, any safety analysis that is based on fatigue failures relates to a new and previously unanalyzed accident.44 SCE has yet to 12

perform an analysis of probable accidents owing to fretting fatigue failures, which it must do before the proposed license amendment could possibly satisfy the second criterion of I 0 C.F.R. § 50.92.

Specifically, Dr. Hopenfeld discusses five possible accident scenarios owing to fretting fatigue not considered by the existing UFSAR. In other words, the risk of these accidents arises from the fact that the tubes have already been substantially fatigued and will experience further fatigue at 70% operation:

I. Fretting fatigue rupture of a tube in the free span with a relief valve stuck open or a broken header;

2. Unplanned closure of an isolation valve, increasing steam pressure abruptly, causing rupture of tubes on the border of exhausting their fatigue life;

3. Seismically-induced ruptures of both plugged and unplugged tubes near the end of their fatigue life;

4. Severe accident causing rupture of tubes near the end of their fatigue life; and

5. Main steam line break accident: in situ tests of tube integrity show only the tendency of tubes to leak on the basis of loss-of-wall-material or weakening by stress corrosion cracks. Fatigue failure would cause propagating circumferential crackS.46 Little data is available to factually assess the safety risks presented by these damaged replacement steam generators due to the unprecedented and unique nature and extent ofthe damage to the tubes in Unit 2's steam generators. Dr. Hopenfeld calculates that ofthe nearly 1100 tubes susceptible to fatigue failure, the probability of only 5 tubes rupturing during Cycle 17 exceeds the NRC's safety goals by a factor of 5.

Thus, the proposed license amendment involves serious risks that SCE and the NRC have not considered, precluding a finding of no significant safety hazards consideration. The risks associated with fretting fatigue are serious, and must be evaluated under the TSTF 449.

Mr. Large also raises a number of considerations not taken into account by the Staff in its no significant hazards consideration determination. While Mr. Large's technical analysis is presented in detail at section 8.6 of his attached declaration, the key points are summarized here. Foremost, Mr. Large emphasizes a critical omission in SCE's analysis: SCE did not adequately consider-despite the evidence of extensive damage to literally hundreds of tubes- the possibility of a multiple tube failure, which would greatly exceed the design basis accident of a single tube burst. When evaluated against the current condition of the steam generators in Unit 2, Mr. Large details a number of situations with the potential for multiple tube failure that were ignored by SCE.

The first of these situations is a scenario in which one of the restraining structures (the anti-vibration bars, or "A VBs"), some of which are already significantly worn, physically detach, more likely to result in multiple tube failure, particularly in the event that fatigue-weakened tubes come into contact with either shrapnel from a single burst tube or the severed tube itself.

Having failed to properly address even the issue offatigue, SCE could not have evaluated, as it must, the effect of fatigue on a new or different type of accident involving multiple tube failures. Last, and significant for the purpose of evaluating the proposed license amendment, fatigue can run its course to failure within a single operation cycle, underscoring the importance of taking this factor into account in accident scenarios.

The fundamental point here is that the damage to the tubes and tube restraint components that occurred during the previous operating cycle at San Onofre Unit 2 was so substantial that the response of these structural components to both normal-as well as possibly adverse-operating conditions have not been accounted for, either in the original design accident cases, nor in the analyses SCE relies upon to justify restarting Unit 2 at 70% of power.

Accordingly, SCE's analysis cannot purport to demonstrate that running the plant at 70% power will not involve the possibility of a new or different kind of accident from the types considered previously. Remember, it is precisely this type of accident, such as a multiple tube failure, that would result in the most severe consequences for public health and safety, a large nuclear accident.

iii. The Proposed Finding of No Significant Hazards Consideration Should Be Withdrawn Because the Proposed License Amendment Would Involve a Significant Reduction in a Margin of Safety.

The assessment in this [Hopen.fold 's} report does not support SCE 's position that operation of Unit 2 for jive months at 70% power will not affect safety. It is shown that SCE conclusions are not conservative. Operation of Unit 2 even for one month at any power level would present a safety risk.6J ---Dr. Joram Hopenfeld NRC's regulations at 10 C.F.R. § 50.92 prevent the Staff from making a finding of no significant hazards consideration where the proposed amendment would involve a significant reduction in a margin of safety. As an initial matter, the ASLB's decision raises a number of serious safety considerations that are evidence that the Staffs position on the no significant hazards consideration is indefensible. SCE's optimistic Operational Assessment estimates ofthe margins of safety of operation at 70% of power are not justified by experience, as the ASLB pointed out:

SCE's prediction that accelerated tube wear will be precluded by plant operations limited to 70% power is grounded on theory that is not yet supported by actual experience .... [T]here is a dearth of applicable experiential data available for in- plane vibrational motion, because, as conceded by SCE, tube-to-tube wear due to in-plant [fluid elastic instability] ha[s] not been previously experienced in U-tube steam generators."

The ASLB further held that the in-plane vibrations caused by FEI were never considered in the UFSAR. The analyses in the UFSAR provide the basis for operating the plant within an acceptable margin of safety. Restarting a reactor unit with known defects caused by mechanisms (e.g., in-plane FEI) that were not analyzed in the UFSAR thus significantly decreases the margin 13

of safety provided for by the UFSAR.

FoE's experts agree that SCE and the Staff cannot show that SCE's license amendment proposal would maintain the required margin of safety in the current license. Dr. Hopenfeld, for example, concludes that operating Unit 2 at 70% of power for Cycle 17 would not be in compliance with ASME code, as required by 10 C.F.R. § 50.55(a), because many ofthe tubes in Unit 2's steam generators have exhausted their fatigue life. An increased risk of a MSLB accident is an obvious example ofthe significant reduction in the margin of safety posed by the license amendment request, since such an accident would cause the largest leakage from the fatigued tubes.

According to Dr. Hopenfeld's analysis, the proposed license amendment would increase the Large Early Release Frequency (LERF) of radiation escaping to the environment to a level five times greater than the Commission's stated safety goals.66 A five-fold increase in risk with potential for large-scale human exposure and the evacuation of southern California is undoubtedly a "significant reduction in the margin of safety."

Mr. Large similarly rejects SCE's conclusion that the proposed amendment would not involve a significant reduction in a margin of safety on the grounds that when it was originally determined, the safety margin67 required by the NRC assumed that the functionality of the replacement steam generators complied with the design specifications.6s The fact that they do not, is now evident. Critically, the import of this is that "any detriment arising from a design omission or design shortcoming," such as those discussed above, would not have been included tubes would not fail from fatigue."73 MHI's analysis was based on erroneous assumptions, however. When corrected, MHI's model would predict tube failure from fatigue because the stress on the tubes exceeds the ASME Endurance Limit.

Taken together, these analyses by FoE's experts show that the proposed amendment would involve a significant reduction in the margin of safety of Unit 2.

iv. Summary In order to issue a finding of no significant hazards considerations, the NRC Staff bears the burden of showing that the hazards considerations raised by Friends of the Earth's experts in these comments and by 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.

c. National Environmental Policy Act The proposed license amendment should not be considered prior to a public hearing because the proposal presents a significant hazards consideration. The National Environmental Policy Act of 1969 (NEP A), 42 U .S.C. § 4321 et seq., requires NRC Staff in such circumstances to at least prepare an Environmental Assessment (EA), which the Staff has not yet done. NEPA requires federal agencies such as the NRC to examine and report on the environmental consequences of their actions. NEPA is an "essentially procedural" statute intended to ensure "fully informed and well considered" decisionmaking.1s Under NEPA, each federal agency must prepare an Environmental Impact

• Statement ("EIS") before taking a "major Federal action[] significantly affecting the quality of the human environment."

An agency can avoid preparing an EIS, however, if it conducts an Environmental Assessment "EA) and makes a Finding ofNo Significant Impact ("FONSI").77 Specifically, no EIS is required if the agency conducts an EA and issues a FONSI sufficiently explaining why the proposed action will not have a significant environmental impact.7s However, in deciding whether to prepare an EIS, the agency must 1)

"accurately identifly] the relevant environmental concern," 2) take a "hard look at the problem in preparing its EA," 3) make a "convincing case for its finding of no significant impact," and 4) show that even if a significant impact will occur, "changes or safeguards in the project sufficiently reduce the impact to a minimum."79 An agency's decision not to prepare an EIS must be set aside if it is "arbitrary, capricious, an abuse of discretion, or otherwise not in accordance with law."

The Federal Register notice is silent as to the application ofNEPA to this case. One can only conclude that the Staff is relying on the categorical exemption from the procedural requirements ofthe NEPA, as described in NRC's regulations at 10 C.F.R. § 51.22(c)(9),

available when the Staff makes a finding of no significant hazards consideration. However, as FoE and NRDC demonstrate in these comments, the NRC Staff cannot make such a finding in this instance, unless they deny scientific reality.

At the very least, an EA and subsequent FONSI must be completed because the proposed amendment would allow steam generators with a severe and dangerous level of wear to operate without repair. Since the leak of radioactive steam in January 2012 resulting from rapid wear in the steam generator tubes, yet, the licensee has proposed no actions to prevent the conditions that caused the leak from reoccurring. The proposed license amendment therefore poses great potential risk to the environment, as shown by the analyses of FoE's experts and the recent ASLB decision, and thus requires the NRC to follow the procedures under NEPA to address that risk.

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IV. CONCLUSION For the foregoing ~easons,.the Staff's proposed finding of no significant hazards consideration should be withdrawn and the significant hazards consideration instead referred to an ASLB, with an attendant public adjudicatory hearing held prior to a decision on SCE's proposed license amendment. 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 ofthe 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."

There is simply no scientific basis for a no-sif(nificant hazards consideration determination in the case of the proposed license amendment for San Onofre Unit 2.

This document will be posted on the web at this link: DAB Safety Team Documents.

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 May 29, 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.

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