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Nuclear Regulatory Commission denies request by environmental group


Thursday, November 8th, 2012
Issue 45, Volume 16.


SAN ONOFRE - Environmentalists who tried to get the Nuclear Regulatory Commission to issue a stay to prevent a restart of the San Onofre Nuclear Generating Station said today their motion was denied.

However, the Friends of the Earth's request for a stay, and to have plant operator Southern California Edison go through a licensing amendment process for new steam generators, will be reviewed by NRC staff and the agency's Atomic Safety and Licensing Board. The five-member, presidentially appointed panel made the decision in Rockville, Md.

The FOE contends the license granted to Edison should have been amended when new steam generators were installed two years ago since they were of a different design than what was in the original permit.

A leak in one of the reactors at the end of January led to the shutdown of the nuclear plant on the northern San Diego County coastline. The other reactor was already offline for maintenance.

An investigation found an unusual amount of wear in many of the thousands of steam tubes in the plant's generating systems. SCE has been conducting tests on the two units at San Onofre.

The environmental group has been fighting for months to keep the reactors from being restarted. The utility and NRC have both said there will be no restart until the plant is deemed to be safe.

The California Public Utilities Commission is also investigating the nuclear plant, including the issue of whether ratepayers should receive refunds for the time it has been out of service. San Diego Gas & Electric owns one- fifth of the facility and receives 20 percent of its power.


Nuclear Regulatory Commission to consider demand by environmental group to keep San Onofre inoperative until license is amended

SAN ONOFRE - The Nuclear Regulatory Commission was scheduled to meet on the East Coast today to Advertisement
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consider a demand by an environmental group to keep the troubled San Onofre Nuclear Generating Station inoperative until its license is amended.

The group Friends of the Earth contends the license granted to plant operator Southern California Edison should have been amended when new steam generators were installed two years ago since they were of a different design than what was in the original permit.

A leak in one of the reactors at the end of January led to the nuclear plant on the northern San Diego County coastline being shut down. The other reactor was already offline for maintenance.

An investigation found an unusual amount of wear in many of the thousands of steam tubes in the plant's generating systems. SCE has been conducting tests on the two units at San Onofre.

The environmental group has been fighting for months to keep the reactors from restarting. The utility and NRC have both said there will be no restart until the plant is deemed to be safe.

Today's meeting starts at 6 a.m. Pacific Time and is being held at a hotel in Rockville, Md., where the presidentially appointed five-member commission is based and typically meets. The meeting is open to the public and will be broadcast over the Internet at nrc.gov.

FOE plans to ask the commission for a stay, which would keep the plant shut down until SCE's operating permit is amended. That wouldn't happen until after a public hearing takes place that includes expert testimony.

The California Public Utilities Commission is also investigating the nuclear plant, including the issue of whether ratepayers should receive refunds for the time it has been out of service. San Diego Gas & Electric owns one- fifth of the facility and receives 20 percent of its power.

SCE did not immediately respond to a message asking for comment.


 

5 comments

Comment Profile ImageBill Hawkins
Comment #1 | Thursday, Nov 8, 2012 at 7:58 pm
PENDING NEWS RELEASE: The DAB Safety Team

San Onfre Latest Problem with Incomplete Unit 2 Inspections

The original Combustion Engineering steam generators (OSGs) for San Onofre Units 2 and 3 were supposed to last for forty years, the design life of the reactors. At Edison’s request, Mitsubishi made numerous untested and unanalyzed changes to the design of the replacement steam generators (RSGS) compared to those originally at San Onofre, such as adding hundreds of more tubes and increasing the height of the U-tube bundle. Yet, by asserting that it was making a “like for like” change, SCE bypassed the normal requirement to apply for a license amendment, which would have entailed a higher degree of scrutiny by the NRC and the opportunity for the public to request an evidentiary hearing. This turned out to be a fateful decision, because a greater degree of review that would have been required with a full license amendment application might have detected the problems that the design changes caused and that have since crippled San Onofre. Palo Verde Nuclear Generation Station (PVNGS) made similar changes as SONGS to their largest Combustion Engineering OSGs in the world, went through an extensive NRC license amendment process and PVNGs 6 RSGs are operating for the last 10 years without any significant problems.

Eight tube failures and thousands of damaged tubes in SONGS Unit 3 only after 11 months of operation now only means that SCE was operating beyond their Current License Basis (CLB), which they have yet to be penalized for by the NRC, as required by law. This event also offers proof, for the very first time, that had a Main Steam Line Break accident occurred, at least eight RSG tubes would have leaked/ruptured and thereby caused the SONGS reactor to Meltdown, due to loss of radioactive core coolant!

The truth is that San Onofre escaped becoming an International Nuclear Events Scale (INES) Level 7 nuclear disaster by the slightest of margins, unlike Fukushima!

By means of plant technical specifications, licensees are required to assure with high confidence that steam generator tubes have sufficient integrity to survive normal operations as well as possible design basis accidents, such as the rupture of a main steam line outside of the containment boundary. Several steam generator tube rupture and steam line break events have occurred in the last 30 years at nuclear power plants throughout the world and one of the contributor cause of these rupture has been determined to be high cycle fatigue. The source of the loads on tube support plates responsible for tube ruptures is believed to be a combination of a mean stress level in the tube and a superimposed alternating stress (The mean stress is produced by denting of the tube at the uppermost tube support plate, and the alternating stress is the result of out-of-plane deflection of the U-bend portion of the tube above the uppermost support plate, caused by flow-induced vibration). The techniques [used to look for cracked steam generator tubes] are not nearly so reliable for determining the depth of a crack, and in particular, whether a crack penetrates through 40% of the tube wall thickness." NRC's regulations do not allow a nuclear plant to start up with any steam generator tube cracked more than 40 percent of its wall thickness. Bobbin and Rotating Coil (+Point) probe inspections are inadequate to detect circumferential cracking and vibration-induced cyclic fatigue. To address crack detection, the industry relies on surface-riding rotating probes that can detect both axial and circumferential cracks. This is a very time consuming and costly process.

SONGS RCE States, “Visual inspection of the tube sheet primary side of the SG 3E088, with the secondary side pressurized, identified the tube with the leak. Subsequent Eddy Current Testing (ECT) inspection identified extensive unexpected tube free-span wear at the leakage location, typically not seen in recirculating SGs, and tube-to-support wear. A full-length ECT inspection of each tube (100%) in all four SGs using a bobbin coil probe was performed and provided a comprehensive extent of condition evaluation for tube free-span and tube-to-support wear. The bobbin coil probe inspection was supplemented by Rotating Coil (+Point) probe inspection, which provided further confirmation of the extent of condition. This supplemental rotating probe examination covered the U-bend portion of approximately 1300 tubes in each SG.” Thousands of Unit 2 Steam generator tubes, Anti-vibration Bars and Tube support Plates have suffered extensive wear and undetermined amount of cracking (incubation and visible) during 22 months from flow-induced random vibrations and cyclic fatigue. Bobbin and Rotating Coil (+Point) probe inspections are inadequate to detect circumferential cracking and vibration-induced cyclic fatigue. In addition, SONGS has performed remote visual inspection of only 8% of the tubes and this does include inspection of any tube support plates for any visible damage for structural deformation or cracks or any other abnormalities in the degraded Unit 2 RSGs.

Detection and quantification of additional modes of degradation such as pitting, inter-granular attack (IGA), axial cracking and circumferential cracking require use of specialized probes. Additional inspections need to be carried out with the best available and qualified technology for detection of crack-like flaws. In a Rush to Restart Unit 2, Edison has not inspected more than 2000 Unit 2 Steam Generator tubes with Transmit/Receive (T/R) array and Intelligent probes, and Ultrasonic testing (UT) techniques to resolve detection and quantification of critical modes of degradation such as axial cracking and circumferential cracking at the U-bends and tube-support intersections. By not performing these inspections, Edison has not met the performance for Unit 2 Steam Generator degraded tubes the criteria specified in Title 10 of the Code of Federal Regulations (10 CFR), “Energy”, which establishes the fundamental regulatory requirements for the integrity of the SG tubes. The SG tubes function as an integral part of the reactor coolant pressure boundary (RCPB) and, in addition, isolate radioactive fission products in the primary reactor coolant from the secondary coolant and the environment. Thus, the SG tubing serves a containment function as well as an RCPB function. SG tube leakage (i.e., primary-to-secondary leakage) or ruptures have a number of potential safety implications, including those associated with allowing fission products in the primary coolant to escape into the environment through the secondary system. In the event of an MSLB accident or stuck open SG safety valve, leakage of primary coolant through the tubes could contaminate the flow out of the ruptured steam line or safety valve, respectively. In addition, leakage of primary coolant through the SG tubing could deplete the inventory of water available for long-term cooling of the core in the event of an accident.
The Operational Assessments reports prepared by AREVA, MHI and Westinghouse “conflict and contradict” with each other on the causes and extent of degradation pertaining to the SONGS Unit 2 Steam Generator Replacement Generators. DAB Safety Team Expert Panel, former NRC Staff and SONGS Concerned Insiders opinion is that these reports are full of holes and “Smoking Mirrors”, because: (1) SCE Engineers have either not provided, or they are withholding all the information to these parties because of “The consequences of being Wrong, Terminated or Fired”, (2) Due to competing and proprietary interests between the three NEI qualified, “US Nuclear Plant Designers”, (3) Time Pressure exerted by SCE for rush to Restart Unit 2, and (4) Since no body knows, what really happened, all the Parties have a shared interest to “Operate Unit 2 at reduced power as a “Test Lab to conduct Unapproved Experiments “ to determine, “What really went wrong with unit 3, so SCE can determine the Root Cause, corrective actions, repair and test plans to return both units 2 and 3 to full power operations.” Now SCE is trying to restart the damage unit 2 by bypassing the normal requirement to apply for a license amendment, which would entail a higher degree of scrutiny by the NRC and the opportunity for the public to request an evidentiary hearing.


The DAB Safety Team has concluded that SONGS Unit 2 Replacement Steam Generators (RSG) are in worse shape now than certified by the SCE and their three NEI Qualified, “U.S. Nuclear Plant Designers.” Even at 70% power operations, if a steam line break outside containment were to occur in Unit 2, the depressurization of the steam generators with the failure of a main steam isolation valve to close, it would result in 100% void fraction in the degraded U-Tube bundle and the straight leg portion between the Tube Support Plates. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI) and flow-induced random vibrations, which would then result in massive cascading SG tube failures, involving hundreds of degraded active SG tubes, along with all the damaged inactive (all the plugged /stabilized) SG tubes. With an undetermined amount of simultaneous tube leaks/ruptures, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five minutes from a broken steam line would be EXCEED the SONGS NRC approved safety margins. So, in essence, the RSG’s will become loaded guns, or a nuclear accident waiting to happen. Any failure under these conditions, would allow significant amounts of radiation to escape to the atmosphere and a major nuclear accident would easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor! Since these events would happen at an extremely fast pace, no credit is assumed in the first 5 minutes of the main steam line break accident for: (1) Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item 9.0, and (2) The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with their Emergency Operating - Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item 5.2,2, Probabilistic Risk analysis.

The above statement is consistent with the conclusions and reports provided earlier on this subject by:

1. Fairewinds Associates Internationally Known Nuclear Consultant Arnie Gundersen and his team of Anonymous Industry insiders, who have had lengthy careers in steam generator design, fabrication, and operation.

2. Professior Daniel Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.

3. Dr. Joram Hopenfeld, a retired engineer from the Office of Nuclear Regulatory Research and NRC's Advisory Committee on Reactor Safeguards (ACRS) report issued in February 2001, which substantiated many of Dr. Hopenfeld's concerns,

4. David A. Lochbaum, Director of the Nuclear Safety Project for the Union of Concerned Scientists (UCS).

EXAMPLES OF STEAM GENERATORS TUBE RUPTURES/STEAM LINE BREAKS

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). Analyses carried out by EDF, further to the last two events, resulted in them being attributed to high cycle fatigue of steam generator tubes due to flow-induced vibration.


On July 15, 1987, a steam generator tube rupture (600 gallons per minute) event occurred at North Anna Unit 1 shortly after the unit reached 100% power. The cause of the tube rupture has been determined to be high cycle fatigue. The source of the loads is believed to be a combination of a mean stress level in the tube and a superimposed alternating stress (The mean stress is produced by denting of the tube at the uppermost tube support plate, and the alternating stress is the result of out-of-plane deflection of the U-bend portion of the tube above the uppermost support plate, caused by flow-induced vibration). Denting also shifts the maximum tube bending stress to the vicinity of the uppermost tube support plate. The rupture extended circumferentially 360ø around the tube. Based on available information, the staff concludes that the presence of all the following conditions could lead to a rapidly propagating fatigue failure such as occurred at North Anna: (1) Denting at the upper support plate, (2) A fluid-elastic stability ratio approaching that for the tube that ruptured at North Anna, and (3) Absence of effective AVB support.


At around 13: 50, 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. 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 that an estimated 7 million Bq had been released into the sea and that 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 raised social concerns. 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 MPa. The ruptured tube may have been dented by the metal plate, which could have swelled from rusting over the years.

On Monday, 9 August 2004 a fatal accident happened at the Mihama No. 3 nuclear power station in Fukui prefecture, Japan. The plant is owned and run by Kansai Electric Power corporation (KEPCO), the major power utility in Western Japan. Four workers were scalded to death by superheated steam, seven other workers were injured. The accident happened when the reactor was about to undergo routine maintenance. The accident was caused by a bursting steam pipe in the non-radioactive part of the reactor. In 27 years of operation that 56 cm diameter pipe had not once been checked for corrosion, let alone replaced. By the time it burst, its walls had worn down from an initial 10 mm of carbon steel to a mere 1.4 mm. Regulations required the pipes to be replaced when the walls were eroded to a thickness of 4.7 mm. Nine months before the accident a subcontractor company had alerted the operators to the need for inspections, but the warning was ignored.

Non-Isolable Main Steam Line Break Outside Containment: Main steam line breaks (or equivalent ruptures in attached piping or equipment) may be caused by a combination of stresses from restriction of pipe thermal expansion by pipe supports, weld defects, lack of pipe stress relief, age-related erosion/corrosion, vibration-induced cyclic fatigue, or repeated safety valve operation causing fatigue cycles to the piping and tubes and increasing the likelihood of a safety valve sticking open. Relatively large steam line breaks have occurred outside the containment, upstream of the MSIV, during hot functional testing at Robinson 2 and Turkey Point 3. These resulted in collateral valve, piping, and equipment damage; blowdown of the affected SGs; and excessive cooldown of the RCS. In addition, large amplitude vibrations of components and structures, water hammers, and sonic booms that affected operator communication and actions were observed. The Turkey Point 3 event involved SG re-pressurization shortly after the initial blowdown as a result of collateral damage.
Comment Profile Imagegme
Comment #2 | Thursday, Nov 8, 2012 at 10:29 pm
I think it's going to take legal action by environmental groups to force the closure of a plant that should never have been built in an earthquake zone in the first place. Remember Japan, NRC!
Comment Profile ImageJust the facts
Comment #3 | Friday, Nov 9, 2012 at 8:39 am
Hey Bill,
What is the DAB safety team? Who is on the team and what are their credentials? Has this group actually been given access to test the equipment at SONGS that they are reporting on here or is this speculation?
Comment Profile Imagereally?
Comment #4 | Friday, Nov 9, 2012 at 5:58 pm
That dangerous, unstable and terrifying plant provides 20% of our power? One fifth. There was a t-shirt in the 60's when the beginnings of this nuclear insanity began:

If you don't want Nukes, turn them OFF. The switch is on your wall.

It's still on your wall. Every family in CA could cut back 1/5th of their power usage--or more-with really very little inconvenience. And then we could all sleep better while Sun and Wind and Water sources of power are developed.
Comment Profile ImageBill Hawkins
Comment #5 | Friday, Nov 9, 2012 at 7:07 pm
OFFICIAL NEWS RELEASE
DAB Safety Team
Media Contact: Don Leichtlin (619) 296-9928 or Ace Hoffman (760) 720-7261

Don Leichtling and Ace Hoffman are the spokesmen of the DAB Safety Team, who along with the support of an ever-growing number of SONGS Concerned Insiders and Whistleblowers have prepared the following analysis.


Subject: SONGS Unit 2 RSG’s Incomplete and Inadequate Tube Inspections

San Diego, CA (November 9, 2012) – The failure of eight Replacement Steam Generator (RSG) tubes (something which has never happened before) plus the structural integrity of thousands of additional damaged tubes in both SONGS Units 2 and 3 RSG’s are now in question and affect the safety of about 8 million Southern Californians. This design failure, which has been termed by the NRC as “a very serious safety issue” because of the unprecedented tube damage: Unit 3 has 807 tubes plugged and Unit 2 has 510 tubes plugged – which makes them the 2 WORST RSG’s in the history of the entire U.S. Nuclear “Fleet”. Which means that SCE was operating beyond their Current License Basis (CLB) and Safety Analysis Limits, plus SCE has yet to be penalized by the NRC for these violations, as required by law. These RSG failures also offers real proof for the very first time, that if a Main Steam Line Break accident had occurred, at least eight RSG tubes would have leaked/ruptured and potentially caused the SONGS Unit 3 reactor to become a nuclear disaster (e.g., like Fukushima, Chernobyl or Three Mile Island) due to the loss of undetermined amount of radioactive core coolant!

The truth is that San Onofre escaped becoming an International Nuclear Events Scale (INES) Level 7 nuclear disaster by the slightest of margins, unlike Fukushima!

SONGS Reactor Compatibility Experiment (RCE) States, “Visual inspection of the tube sheet primary side of the SG 3E088, with the secondary side pressurized, identified the tube with the leak. Subsequent Eddy Current Testing (ECT) inspection identified extensive unexpected tube free-span wear at the leakage location, not typically seen in recirculating SGs, and tube-to-support wear. A full-length ECT inspection of each tube (100%) in all four SGs using a bobbin coil probe was performed and provided a comprehensive extent of condition evaluation for tube free-span and tube-to-support wear. The bobbin coil probe inspection was supplemented by Rotating Coil (+Point) probe inspection, which provided further confirmation of the extent of condition. This supplemental rotating probe examination covered the U-bend portion of approximately 1300 tubes in each SG.” Thousands of Unit 2 Steam generator tubes, Anti-vibration Bars and Tube support Plates have suffered extensive wear and undetermined amount of internal cracking during 22 months from flow-induced random vibrations and/or cyclic fatigue. In addition, SCE has only performed remote visual inspection of 8% of the tubes in their damaged Unit 2 RSG’s plus this does include inspection of any tube support plates for any visible damage, structural deformation, cracks and/or any other abnormalities.

The methods employed by SONGS using bobbin and rotating coil (+Point) probes to inspect the Unit 2 damaged tubes cannot reliably determine the depth, extent and location of these cracks, so the actual condition of the tubes remains a dangerous unknown! In a Rush to Restart Unit 2 and cut their costs, Edison has not inspected more than 2000 Unit 2 Steam Generator tubes with T/R single-pass array probes, laser-scanned penetrant inspection and ultrasonic detection technologies to accurately identify and determine the depth, extent and location of these internal cracks at the U-bends and tube-support intersections. NRC's regulations do not allow SONGS to start up with any steam generator tube cracked more than 35% of its wall thickness. Therefore by not performing these accurate inspections, Edison has not met the performance criteria specified in 10 CFR Part 50, Appendix A, “General Design Criteria for Nuclear Power Plants,” Criterion 14, 15, 30 and 32, which establishes the fundamental regulatory requirements for the integrity of the SG tubes.

The DAB Safety Team has concluded that SONGS Unit 2 Replacement Steam Generators (RSG) are in worse shape now than certified by SCE and their three NEI Qualified, “U.S. Nuclear Plant Designers.” The accident scenario of concern consists of two events: (1) a non-isolable secondary system break or rupture that is outside containment; and (2) a coupling of this break with the rupture of, or significantly increased leakage from, affected SG tubes. Even at 70% power operations, if a steam line break outside containment were to occur in Unit 2, the depressurization of the steam generators with the failure of a main steam isolation valve to close would result in 100% void fraction in the degraded U-Tube bundle and the straight leg portion between the Tube Support Plates. This condition of ZERO Water in the steam generators would cause fluid elastic instability (FEI) and flow-induced random vibrations, which would then result in massive cascading SG tube failures, involving hundreds of degraded active SG tubes. Fluid elastic instability (FEI) and flow-induced random vibrations can progress through a buffer zone of plugged tubes to reach pressurized, in-service tubes and create additional SG tube failures. The resulting SG secondary side blow-down could further increase tube leakage due to resonance vibrations within the affected SG tube bundle. With an undetermined amount of simultaneous tube leaks/ruptures, approximately 60 tons of very hot high-pressure radioactive reactor coolant would leak into the secondary system. The release of this amount of radioactive primary coolant, along with an additional approximately 200 tons of steam in the first five minutes from a broken steam line would EXCEED the SONGS NRC approved safety margins. So, in essence, the RSG’s will become loaded guns, or a nuclear accident waiting to happen. Any failure under these conditions, would allow significant amounts of radiation to escape to the atmosphere and a major nuclear accident would easily result causing much wider radiological consequences and even a potential nuclear meltdown of the reactor! Since these events would happen at an extremely fast pace, no credit is assumed in the first 5 minutes of the main steam line break accident for: (1) Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item 9.0, and (2) The differential pressure across the SG tubes necessary to cause a rupture will not occur if operators prevent RCS re-pressurization in accordance with their Emergency Operating Procedures - Enhanced Unit 2 Defense-In-Depth Actions - SCE Restart Plan Enclosure 2, Item 5.2.2 - Probabilistic Risk analysis.

In Emergency Planning Space, decisions have to be Accurate and Timely. Under-conservative, rushed and profit-motivated analyses based on limited facts, biased and ambiguous operational data, untested deterministic and probabilistic risk analysis, conflicting theories and differing operational assessments of degraded equipment at even reduced power operations for 150 days with conditional monitoring along with unproven and unreliable compensatory actions represent enormous risks to public safety, the environment and our nation’s economy.

The NRC must REALLY resolve the concerns stated above as soon as possible. In the interim, the NRC must stop making favorable decisions to SCE (especially when it lacks defensible technical and inspection bases), which affect the lives of millions of Southern Californians."

The above analysis is consistent with the conclusions and reports provided earlier on this subject by:
1. Fairewinds Associates Internationally Known Nuclear Consultant Arnie Gundersen and his team of Anonymous Industry insiders, who have had lengthy careers in the design, fabrication, and operation of nuclear steam generators.
2. Professor Daniel Hirsch and Internationally Known Nuclear Consultant Dale Bridenbaugh.
3. Dr. Joram Hopenfeld, a retired engineer from the Office of Nuclear Regulatory Research and NRC's Advisory Committee on Reactor Safeguards (ACRS) report issued in February 2001, which substantiated many of Dr. Hopenfeld's concerns.
4. David A. Lochbaum, Director of the Nuclear Safety Project for the Union of Concerned Scientists(UCS).


CPUC: California Public Utilities Commission
DBA: Design Basis Accident
ECT: Eddy Current Testing
FEI: Fluid Elastic Instability
MHI: Mitsubishi Heavy Industry
MSLB: Main Steam Line Break
NRC: Nuclear Regulatory Commission
RCE: Reactor Compatibility Experiment
SCE: Southern California Edison
TTW: Tube-to-Tube Wear

________________________________________________________________

Copyright November 9, 2012 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 Attorneys.

Article Comments are contributed by our readers, and do not necessarily reflect the views of The Fallbrook Village News staff. The name listed as the author for comments cannot be verified; Comment authors are not guaranteed to be who they claim they are.

 

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