I realize the document below, which was left at the NRG BLOG web site this morning by a friend of mine, is fairly technical, so I've added a GLOSSARY at the bottom to help out.
But nevertheless I highly recommend reading it (it's fairly short) and discussing it with as many people as possible in (and especially out of) the movement to shut San Onofre. It is the most succinct technical analysis I've seen to date on why it is unsafe to restart San Onofre Unit 2, even at 70% power.
(Note #1: Minor corrections to the version posted at the NRC BLOG have been made to the version below. Please refer to my animation for additional parts placements: http://goo.gl/CRlQl and to yesterday's newsletter for additional discussion: http://goo.gl/q33yL )
(Note #2: Another important point, courtesy Donna Gilmore, http://SanOnofreSafety.org "If the NRC approves restart, Edison will be able to run the plant longer than 5 months. The Confirmatory Action Letter (CAL) gives Edison the discretion to make that decision after the 5 month test period ends. The NRC has refused to modify the CAL to require NRC approval after the 5 month test.")
Subject: US NRC Blog
DATE: March 19, 2013 at 1:12 am
NRC STATUS: Your comment is awaiting moderation.
Sincere Thanks to Mr. Victor Dricks, Mr. Cale Young, Mr. Ryan Lantz, Mr. Randy Hall and entire NRC Staff. Thanks to NRC posting this blog
San Onofre NRC/SCE/MHI/ Public Education Series by HAHN BABA Statement of facts unless proven wrong otherwise ……… Southern California Edison Submits Operational Assessment Requested by NRC NRR RAI 32 Putting Production/Profits over Safety
1. BACKGROUND: ROSEMEAD, Calif., March 18, 2013 — A new technical evaluation of San Onofre Nuclear Generating Station Unit 2 demonstrates that the Unit 2 steam generators could be operated safely at 100 percent power and reinforces Southern California Edison's (SCE) more conservative plan to begin operating Unit 2 at 70 percent power for five months. SCE submitted the operational assessment of potential Unit 2 steam generator tube wear to the Nuclear Regulatory Commission in response to NRC questions. The new evaluation determined Unit 2 could operate at full power for 11 months with full tube integrity. The assessment was performed by Intertek APTECH of Sunnyvale, CA, and supplements Intertek's earlier assessment of Unit 2 operation at 70 percent power. Intertek performs operational assessments relating to steam generators for many nuclear power plants around the U.S. "This evaluation confirms the structural integrity of the Unit 2 steam generators at 100 percent power, as requested by the NRC," said Pete Dietrich, SCE senior vice president and chief nuclear officer. "While we have no intent to restart Unit 2 at full power, this demonstrates the amount of safety margin we have built into our analyses. We welcome this additional safety analysis but remain steadfast in our commitment to restart Unit 2 at only 70 percent power."
2. Facts to Dispute/Refute SCE Claim
A. San Onofre Unit 3 operation @100 Power 11 Months- SONGS Unit 3 Failed In-situ Wear Data Unit 3 SG 3E088 (www.nrc.gov). The following tube wear data is based on a result of actual tube degradation in SONGS Unit 3 SG 3E088 caused by fluid elastic instability.
Row 106 Column 78, 100 percent through wall wear, length of wear 29 inches
Row 102 Column 78, 99 percent through wall wear, length of wear 23 inches
Row 104 Column 78, 99 percent through wall wear, length of wear 27 inches
Row 100 Column 80, 81 percent through wall wear, length of wear 28 inches
Row 107, Column 77, 80 percent through wall wear, length of wear 34 inches
Row 101, Column 81, 78 percent through wall wear, length of wear 26 inches
Row 98, Column 80, 72 percent through wall wear, length of wear 29 inches
Row 99, Column 81, 72 percent through wall wear, length of wear 27 inches
B. Mitsubishi Root Cause Document UES-20120254, Rev 0, page 13 of 64, Item 1, "Tube to Tube Wear due to in-plane FEI" states," Tube to tube wear was found in the U-bend region, located between AVBs, in the free span. Many of the tubes exhibiting tube to tube wear also exhibited wear at the AVBs and TSPs, in particular at the top tube support plate. For tubes with wear at the top tube support plate, it is considered that the entire tube, including its straight region, is vibrating. Tube to tube wear occurs when there is tube in-plane motion (vibration) with a displacement (amplitude) greater than the distance between the tubes in the adjacent rows, resulting in tube-to-tube contact. Some of the tubes with tube-tube wear did not experience large amplitude vibration but were impacted by tubes that did experience large amplitude vibration. Also the two tubes in Unit 2 with tube-to-tube wear had different wear characteristics than the Unit 3 tube-to-tube wear."
C. Based on Dr. Pettigrew's Research and other papers published between 2006 -2011 on fluid elastic instability experimental data, "The high dry steam velocities differ in the in-plane and out-of-plane directions. For the SONGS RSG tube geometry, based on experimental data, it is conservatively estimated that the high dry steam velocities for in-plane FEI are at least 200 % of the high dry steam velocities for out-of-plane FEI."
D. The SONGS Unit 2 SG tube wear rates calculated by AREVA, Westinghouse and Intertek Operational Assessments and Work Rates are based on the results of out-of-plane velocities, which are under conservative based on FEI Observations in SONGS 3 and Dr. Pettigrew's Research acknowledged by MHI and NRC Chairman and Commissioners.
E. Deterministic Analysis Uniform Linear Tube-to Tube Wear Rate in Unit 2 based on Unit 3 Benchmarking = 100%/11 months = 9%/month, consistent with item D and Actual Observations in SONGS 3
F. Westinghouse in SONGS Unit 2 Return to Service Report, Attachment 6, Appendix D, (www.songs.community.com), page 91 states, "Table 3-2. Wear Projection Results for Active Tubes with Limiting AVB Wear Indications" shows the following active tubes in Unit 2 SG 3E089 with the following data:
Row 119 Column 89, 28 percent ECT reported through wall wear,
Row 121 Column 91, 28 percent ECT reported through wall wear
Row 131 Column 91, 21 percent ECT reported through wall wear
Row 129 Column 93, 22 percent ECT reported through wall wear
Row 126 Column 90, 21 percent ECT reported through wall wear
G. Calculate SG 3E089 tube rupture time @ 9% wear/month for 100% Tube wear @ full power operation
Row 119 Column 89, 28 percent wall wear + 72 % in 8 months = 100 % wear = Tube Rupture
Row 121 Column 91, 28 percent wall wear + 72 % in 8 months = 100 % wear = Tube Rupture
Row 131 Column 91, 21 percent wall wear + 81 % in 9 months = 101 % wear = Tube Rupture
Row 129 Column 93, 22 percent wall wear + 81 % in 9 months = 102 % wear = Tube Rupture
Row 126 Column 90, 21 percent wall wear + 81 % in 9 months = 102 % wear = Tube Rupture
H. Intertek APTECH Operational Assessment referenced in item 1 above, page I-iv states, "Two OA analysis cases were evaluated based on the sizing techniques used to define the Unit 3 TTW depths. Case 1 evaluated the situation where voltage based sizing for Eddy Current Testing Examination Sheet (ETSS) 27902.2 was used to establish the TTW depth distributions and the correlated wear rate with wear index. The results for Case 1 indicate that the Structural Integrity Performance Criteria (SIPC) margin requirements are satisfied for an inspection interval length of 0.94 years (11.5 Months) at 100% power level. For Case 2, where the TTW depths were resized by AREVA using a more realistic calibration standard, the SIPC margins will be met for an inspection interval length of 1.04 years at 100% power level. The plan for Unit 2 is to operate for an inspection interval of 5 months at a 70% power to provide additional margin to the industry requirements for tube integrity. Tube burst at 3xNOPD (Normal Operating Pressure Differential) is the limiting requirement for inspection interval length. Therefore, the accident-induced leakage requirements will be satisfied provided that burst margins at 3xNOPD are maintained during the inspection interval.
I. Deterministic Analysis results shown in item G shows that all the five tubes can rupture in 9 months or less than shown by Intertek in Probabilistic Analysis of 11 months. This Probabilistic Analysis does not meet the intent of NRR RAI 32, in which SCE promised to provide an OA that includes an evaluation of steam generator TTW for operation up to the RTP.
CONCLUSIONS: SCE is once again trying to circumvent and gaming the NRC RAI #32, just like avoiding 10 CFR 50.90 for the Brand New $570 Million RSGs . MHI Anti-vibration bar structure, designed for out-of plane vibrations, is incapable of preventing the adverse effects of tube-to-tube wear or fluid elastic instability (high dry steam) at 100% power operation or main steam line break. We saw the destruction of SONGS Unit 3 RSGs due to tube-to-tube wear or fluid elastic instability (high dry steam) at 100% power operation or Main Steam Line Break Testing. According to the analysis of Unit 2 Plant Operational Data/Procedures and Westinghouse Operational Assessment, fluid elastic instability (high dry steam, high fluid velocities, in-plane vibrations) conditions did not occur in Unit 2. Therefore, this insufficient contact tube-to-AVB forces in Unit 3 causing the FEI is based on hideous data and unreliable MHI Computer Modeling once again. Taking credit for double contact tube-to-AVB forces (Better supports), which prevented in-plane vibrations or Tube-to-tube wear in Unit 2 by NRC Region IV AIT Team/SCE/MHI directly contradicts and conflicts with statements made by Dr. Pettigrew, Westinghouse, AREVA, John Large and inconsistent with Unit 2 Operational data. This analysis by SCE does not meet the intent of Federal Regulations, NRC Steam Generator Tube Structural Integrity Criteria, SONGS NRC Approved Technical Specifications, NRC Reasonable Assurance Criteria, NRC Chairman's Standards and SCE's Overriding Obligation for Public Safety. A Lot More to Come… Thanks NRC Staff… HAHN BABA
Glossary for above:
AIT: Augmented Inspection Team (of the NRC)
AVB: Anti-Vibration Bar
AREVA: French nuclear state-owned corporation
CAL: Confirmatory Action Letter (sent from the NRC to SCE)
CFR: Code of Federal Regulations
ECT: Eddy Current Testing
ETSS: Eddy Current Testing Examination Sheet (27902.2)
FEI: Fluid Elastic Instability
MHI: Mitsubishi Heavy Industries
NOPD: Normal Operating Pressure Differential
NRC: Nuclear Regulatory Commission
NRR: Nuclear Reactor Research division of NRC
OA: Operational Assessment
RAI: Request for Additional Information
RSG: Replacement Steam Generator
RTP: Reactor Thermal Pressure (standard thermal pressure)
SCE: Southern California Edison
SG: Steam Generator (for example, SG 3E089 and SG 3E088 are Unit 3)
SIPC: Structural Integrity Performance Criteria
SONGS: San Onofre Nuclear [Waste] Generating Station
TSP: Tube Support Plate
TTW: Tube-to-Tube Wear
Contact information for the author of this newsletter:
The original author of the above statement which was left at the NRC web site wishes to remain anonymous, but he can be contacted through Ace Hoffman.