ML13155A168
| ML13155A168 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/19/2013 |
| From: | Hawkins B - No Known Affiliation |
| To: | Borchardt R W, Leeds E J, Macfarlane A M Rules, Announcements, and Directives Branch, NRC/Chairman, NRC/EDO, Office of Nuclear Reactor Regulation |
| References | |
| NRC-2013-0070, 78FR22576 00397 | |
| Download: ML13155A168 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 7-0 C')SUNSI Review Complete Template = ADM -013 E-RIDS= ADM-03 Add= B. Benney (bjb) v/Joosten, Sandy From: Bill Hawkins [billlee123456@gmail.com]
Sent: Friday, April 19, 2013 1:13 PM To: CHAIRMAN Resource; Leeds, Eric; Borchardt, Bill; R4ALLEGATION Resource; Lantz, Ryan;Benney, Brian; Hall, Randy; Howell, Art; Dorman, Dan
Subject:
San Onofre -Honorable Dr. Macfarlane
-Please see Recommended and Required Action 1 to justify Your Own Public Statements and Independent Public Safety Position San Onofre NRC/SCE/MHI/Public Awareness Series -Please excuse me for any computer or human performance grammatical or spelling errors.Responsibility:
It is the legal and moral duty of every United States government official and politician to ensure that public safety is not endangered, whether, it is gun violence, terrorist attacks or radiological accidents.
But, that does not seem to be case with NRC preliminary approval of San Onofre proposed License Amendment to Operate Unit 2 at 70% reduced power.Recommend Action 1: NRC Chairman, Dr. Macfarlane is requested to direct Office of the Inspector General (OIG), U.S. Nuclear Regulatory Commission to conduct an inquiry concerning the SCE/MHI handling of issues associated with the San Onofre steam generator tube rupture. This inquiry is required to address concerns raised by 8.4 Million Southern Californians, Numerous Safety Experts, City and State Officials, Director of California Energy Commission, Public Organizations, members of the Congress (Senator Barbara Boxer and Congressman Ed Markey) and to meet the intent of His Excellency, President of the United States, Open Government Initiative and NRC Public Transparent Policy as a result of some of the issues described below.Required Action 1: The NRC Chairman has publically stated that SCE is responsible for the work of MHI, Westinghouse, AREVA and Intertek.
In light of SONGS Units 2 & 3 massive amounts of tube damage (wear) and tube failure in Unit 3, along with incomplete tube inspections for detection of circumferential incubating cracks in Unit 2, based on the review of attached reports and governing standards described below, SCE is legally required to check MHI Fatigue Calculations and post the results on its website before any approval of SONGS proposed New License Amendment for restart of Unit 2, to demonstrate:
That the proposed amendment (1) Would not involve a significant increase in the probability of an accident previously evaluated in the SONGS FSAR; or, (2) Would not create the possibility of a new or different type of accident previously evaluated in the SONGS FSAR; or, (3) Would not involve a significant reduction in the required margin of safety by operating Unit 2 at 70% power However, because of the wear damage previously sustained by Unit 2, some tubes may now be susceptible to rapid fatigue failure.SCE needs to provide a calculation justifying the engineering basis of the above statement to meet the ASME Code, NRC RG 1.121, the NRC Chairman and its own Standards.
The calculation should be performed by a SCE California Licensed Mechanical or Civil Engineer Steam Generator Vibration Expert and Independently Verified (Competent Design Engineering Organization not affiliated with SCE) by a California Licensed Structural Engineer Steam Generator Vibration Expert.Problem Statement:
Two Independent Nuclear Experts Certify that MHI SG Tube Fatigue and Stress Calculations Assumptions are erroneous and based on faulty data. SCE is legally required to certify MHI's calculations to assure that San Onofre Unit 2 does not pose significant radiological risks at 7 0% normal steady state power operations, and during Anticipated Operational Transients and Design Basis Accidents.
SCE Preliminary Response:
Independent Experts' Analysis concerning in-plane tube vibration is significantly flawed in that it applies an unreasonably high stress concentration factor based on solid body geometry rather than the more realistic stress concentration factors for a cylindrical geometry applicable to the SONGS steam generator tubes.1 MHI Response:
MHI did analyze the potential for fatigue failure of the RSG tubes under operating conditions and determined that fatigue was not a credible tube failure mechanism because the stresses sustained by the tubes due to in-plane vibration are well below the stresses that would cause fatigue failure. The analysis that supports this conclusion is contained in Appendix 16 to the"Tube wear of Unit-3 RSG -Technical Evaluation Report." It should be noted that the technical reviews and analysis, both by the NRC and industry experts, have not mentioned fatigue failure of the tubing.FACTS: SONGS Unit 3 SG Tubes leaked and failed due to tube-to tube wear, tube-to-AVB wear, tube-to-TSP wear, Retainer Bar-to-tube wear and high cyclic thermal fatigue induced axial, circumferential, macroscopic and microscopic cracks. Fluid elastic instability, flow-induced vibrations and Mitsubishi Flowering Effect responsible for the above catastrophic effects were caused by exceedingly high steam flows overstretching the thermal performance and reducing significantly the safety margin of SGs tubes to maximize SCE profits, low steam generator pressures, high reactor coolant flows, narrow tube pitch to tube diameter ratio, low tube clearances, extremely tall tube bundle and lack of in-plane restraints.
SONGS Units 2 & 3 Tube Leak and tube-to-tube wear in 2012 -See NRC AIT Report.Defect or Deviation:
In San Onfre replacement steam generators, the relative motion between the tubes and the anti-vibration bars (AVBs), the tube support plates, and the retainer bars have resulted in tube wear and fatigue damage in tubes due to fluid elastic instability, flow-induced random vibrations, excessive fluid hydrodynamic pressures and Mitsubishi Flowering Effect. These adverse phenomena can produce relatively quick tube failures when the stresses generated during vibrations are sufficiently large. As described in Unit 2 Return to Service Report, Attachment 4, MHI Document L504GA564, Appendix 16, page 459 of 474, MHI used a finite element model ("FE"), to calculate that the tubes were subjected to a stress of 4.2 ksi (kilopounds per square inch).Consequently, MHI concluded that the stress on the tube due to in-plane vibration is under fatigue limit (13.6 ksi) and the structural integrity of the tube is confirmed from the view point of fatigue due to in-plane vibration (page 470 of 474). MHI results are based on two erroneous assumptions described below.The source of MHI's error described below resulted from how they calculated the increase in the local stress at geometrical discontinuities (notches), which are formed when two metal surfaces come in contact during vibration.
Since the worn surfaces of the tubes inside the steam generators U-Tube Bundle region of tube-to-tube wear could not be seen during SONGS Tube Inspections, MHI made two incorrect key assumptions, which are inconsistent with the observation that both the tube and the supporting bar are worn into each other. First, MHI assumed that the ASME endurance limit could be applied directly to the notched tube surfaces.
Since it is commonly known in the nuclear and commercial industry that surface roughness significantly reduces fatigue life and since the ASME data is for smooth polished surfaces, this assumption would underestimate the amount of fatigue damage. Secondly, when using the Peterson chart, MHI assumed an unrealistically large fillet radius for sharp corners (Should be zero according to basic knowledge of geometry) and consequently derived a low concentration stress factor. Large radii would decrease the local stress and cause the tube to fail at a higher level of stress, thereby increasing its fatigue life. Only by using these two, arbitrary, non-conservative assumptions was MHI able to conclude that Unit 2 did not suffer any fatigue damage. When these assumptions are corrected, the opposite conclusion is reached, which is that the tubes will be susceptible to failure from fatigue.
Background:
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 changes to its design may have significant safety consequences.
Out-of-plane fluid-elastic instability has been observed in nuclear steam generators in the past and has led to tube bursts at normal operating conditions.
However, the observation of in-plane fluid-elastic instability in steam generators of a nuclear power plant is a true paradigm shift. The combined effects of tube-to-tube wear and high cycle thermal fatigue cracks caused by fluid-elastic instability and/or flow-induced random vibrations have been witnessed as sudden tube ruptures in the following nuclear power plants: North Ana in 1987- NRC Bulletin No. 88-02: Rapidly Propagating Fatigue Cracks In Steam Generator Tubes: MHI SG Japan 1991 -On February 9th, 1991, leakage of about 55 tons of primary coolant occurred due to the failure of one SG tube in a steam generator built by Mitsubishi in the No. 2 pressurized water reactor at the Mihama nuclear power station in Japan. The 2 tube had been severed, causing the massive leakage of contaminated cooling water. At the same time, water pressure in the core had dropped drastically and the ECCS kicked in, flooding the reactor and shutting it down. If the core had been left exposed, a meltdown -an overheating of the fuel that can, if uncontrolled, lead to a large release of radio-activity
-could have occurred.Following week an estimated 7 million Becquerels (Bq) had been released into the sea and an estimated 5 billion Bq of radioactive gas had been released into the atmosphere.
This tube rupture caused the first INES level 3 nuclear incident in Japan, which ignited social concerns all over Japan because it shattered the nuclear industries myth of 100% safe reactors!
The failed tube was removed from the heat exchanger, and the fracture surface was examined by a scanning electron microscope.
Striations, which are a characteristic of fatigue failure, were observed on large portions of the fracture surface, and dimples showing tensile fracture were also observed.However, few traces of stress corrosion cracking and corrosion were found on the fracture surface of the tube. Stress amplitude of the failed tube estimated based on the striation spacing was found to be in the range of around 51 to 60 Million Pascal's (7391-8702 psi > 1.5 X 5,200 psi, SONGS 3 MSLB Test Pressure).
Indian Point 2, 2000 (See NRC Office of Inspector General Report Below)Comanche Peak 1, 2002 -Routine inspections at the Comanche Peak nuclear power plant failed to detect a damaged steam generator tube that later ruptured, forcing a shutdown.
The flaw in the tube was "clearly identifiable and missed" about 18 months ago by workers for TXU Energy, the plant's owner and operator, according to the preliminary findings of a special inspection team of the Nuclear Regulatory Commission.
Oconee 2, 2002 -Failure to meet structural integrity performance criteria in fall 2002, as determined by in situ pressure testing during condition monitoring.
Craus NPP: Between 2004 and 2006, three primary-to-secondary leaks occurred at the Cruas NPP: unit 1 in February 2004 and unit 4 in November 2005 and February 2006. The three leaks were all the result of a circumferential crack in the tube at the location where the tube passes through the uppermost tube support plate (TSP #8).3