NRC in the 2.206 Petition Process Was Censoring Me

CEO due diligence and the historic context of SRV problems...


May 11, 2013:

(Re-Published from April 11, 2013)

May 9, 2013 

These documents came out post my Pilgrim 2.206 and post the beginning of refueling outage. It is suspicious as hell.

Right, widespead Curtiss Wright organizational issue with maintaining quality assurance and the quality of nuclear safety components...multiple runaway organizational QA issues with the NRC, Entergy, Curtiss Wright and Target Rock.

They structured the release of these documents till Entergy got into the outage...all this information was available for distribution months before their officials release...

1. CURTISS WRIGHTReactor Coolant Pump Defective Impeller
2. CURTISS WRIGHT-Target Rock SRV Bellows Failure
3. Licensee Event Report 2013-002-00, SRV-3B Safety Relief Valve Declared

April 13:

I am sure the NRC is shocked I actually taped the session and the implication are I taped them all

I could show you how so out of character this session was with the other recordings!
















If  I could just knock off saying my "you knows",  I could save hours of time from my YouTubes. I am talking way too long to explain what happened on my 2.206.



Does the NRC allow a utility to intentionally state incomplete and inaccurate License Event Reports?

My pretend 2.206 presentations.

PRB 2...I don't know what happened to the thumbnail...

NH Union Leader: worst NE ISO grid crisis in memory...how the erratic operating Pilgrim plant and the worst grid crisis in memory are connected.



National or Regional Emergency: secret government policy to weaken safety at nuke plants in a severe grid power crisis?

Commissioner William C. Ostendorff beautiful speech on nuclear safety...and him knowing Region I is a old style Russian Oligarchy.



Where did the good people go



Generalist versus hyper-specialization.



My e-mail today to Mr. Guzman and a additional e-mail to show you its contextuality! My lawyers felt it was less risk to me if I didn't put up their voice recording...it is safer to get it from them for now. 

From: Michael Mulligan
To: "Guzman, Richard"
Sent: Friday, April 12, 2013 10:10 AM
Subject: Re: 2.206: Pilgrim Nuclear Plant SRV Request for Emergency Shutdown

Mr Guzman,

This is a recording of my 2.206 hearing yesterday...I will be shortly putting up a video showing the public how it should have gone.

http://steamshovel2002.blogspot.com/2013/04/i-had-2.html

Want me to put up a prior recording of how a real hearing should have gone? I don't mind disruptions... but all of your disruptions yesterday was about telling me my take wasn't germane to the Pilgrim and a attempt to hurry me up so you could go to your lunch. I would like the public to hear the contemptuous tone of your voice to me.

I request a "do over" and the next hearing run according to management directive 8.11. Request the agency send me the voice recording of this meeting yesterday.

I will be sending  a e-mail to Representative Markey's office also.

Mike Mullian

Hinsdale, NH

From: Michael Mulligan
To: "Guzman, Richard"
Sent: Monday, April 8, 2013 3:27 PM
Subject: Re: 2.206: Pilgrim Nuclear Plant SRV Request for Emergency Shutdown

Mr. Guzman,

Thanks, I see it now today.

Is your agency totally blowing me off on my earlier request to speak with a Pilgrim resident inspector and a senior region 1 official (like region III did for me with palisades)?

Can't even send me an email saying the request was denied, give me a reason...that would be the professional way of handling it?

After all that recent hard talk with the community and the Plymouth selectman about poor NRC communications and agency mistrust during the Pilgrim performance meeting?

Thanks

Mike Mulligan

Hinsdale, NH
 

April 7, 2013

I felt the petition manager as being perturbed, irked at me...he was blatantly contemptuous of me wasting his and the other NRC official's time. I honestly think he was trying to incite my anger...then the NRC could use that as an excuse to prohibit me from participating in any further 2.206s. 

Think about the control of time management with him and the NRC with scheduling another meeting within 10 minutes of the ending of my 2.206...

Honestly, I think that the next meeting after this 2.206 was made up...to use as a tool to incite my anger.

I made two voice recordings. So I have recorded just my conversation during the petition board and then I recorded my conversation and the inputs from the NRC officials...

I had a 2.206 meeting with the NRC over pilgrim. It didn't go like any other. The petition manager asked we could just skip the general explanation of the 2.206 process.. I have never been asked that before. So I agreed...thought that would just give me extra time to explain my position. He has to go by a format  of  2.206 directive...so he has to explain the 2.206 process and a general explanation of my petition. So we just skipped through that step and I now think it was inappropriate.

So he gave me the floor...he kept disrupting me throughout the 45 minutes. He would say that is not germane to Pilgrim...he was censoring me with what I could say. He would constantly interrupt me with "time is" running out" beginning 10 minutes into my speech. He told me more than once I got a prior commitment ten minutes after you get done, basically hurry up.

This petition manager was trying to censure me with what i was saying and he was trying to intimidated me into throwing me off my stride. He was trying to minimize the chance of getting my ideas on the NRC paperwork.

What in the hell is going on at Entergy, Pilgrim and Region!?

The antis recently gave the NRC a hard time at a  yearly pilgrim safety meeting... was he trying to get even to me over that and show me how it feels to be disrupted. The NRC is supposed to be bigger than the antis. I wasn't at that meeting in Pilgrim and I wouldn't handle myself in that kind of disruptive manner.

I had many 2.206s (20-40) and this was run radically different than any in the past.

This 2026 Petition manager with vast experience doesn't have the temperament to interact with the general public!

I am going to ask for a immediate "do over" and for the agency to give me the voice recording of my Pilgrim 2.206..me and the NRC officials in the meeting. You know I am going to stick it out on the internet.

A flag raised upside down is international signal with a person or group being in extreme distress and peril...

The part that is missing is the petition manager saying this meeting is from say 11 am to 12 am...mike you will have something like 35 to 45 minutes to speak. They set the boundaries of the meeting time...they give me a 5 minute warning that your time is almost up.

I think the guy was pissed off this meeting was cutting into his lunch time...

So this is excepts from my most recent Vermont Yankee 2.206 petition

Tuesday, April 2, 2013

I don't believe this was included in today's NRC explanation...it was scheduled for an hour from such and such a time.

As part of the Petition Review Board's review of this petition, Mr. Michael Mulligan has requested this opportunity to address the PRB. This meeting is scheduled from 9:00 to 10:00 a.m.

The below is boiler plate.

The meeting is being recorded by the NRC Operations Center and will be transcribed by a Court Reporter. The transcript will become a supplement to the petition. The transcript will also be made publicly available. 

This process permits anyone to petition NRC to take enforcement-type action related to NRC licensees or licensed activities. Depending on the results of this evaluation, NRC could modify, suspend, or revoke an NRC-issued license, or take any other appropriate enforcement action to resolve a problem. 

The NRC staff guidance for the disposition of 2.206 petition's request is Management Directive 8.11, which is publicly available. 

See, the below states I can make any kind of comment or explanation I think would support my petition... but I have to be respectful to the PRB board. I am talking to both the PRB board and to the public at large through the NRC's documented system. I just don't have to meet the so called needs of the PBR board. They were playing me to excite my anger.

The purpose of today's meeting is to give the Petitioner an opportunity to comment on the PRB's initial recommendation to not accept the petition and a second opportunity to provide any additional explanation or support for the petition.

Right, basically no information of any kind on the problems of the SRVs has been disclosed to the public...evidence wise, I don't have a leg to stand on. But that is not my fault...it is the secrecy of the NRC and  Entergy that is causing my lack of evidence.

So all meetings, well most, are scheduled for a hour. Then you got 45 minutes for your presentation. So why all this "get to the point" and hurry up", this is not appropriated material, I got a meeting ten minutes after this ends. Why all this pushing to not waste time...I know it is already rejected?

The Petitioner will have 45 minutes to address the PRB. This meeting is not a hearing, nor is it an opportunity for the Petitioner to request or examine the PRB on the merits or the issues presented in the petition request. No decisions regarding the merits of the petition will be made at this meeting.

Boiler plate:

I would like to describe the scope of the petition under consideration and the NRC activities to date. On December 5, 2012, Mr. Mulligan submitted to the NRC a petition under 2.206 regarding the Vermont Yankee Nuclear Power Station. Mr. Mulligan requested an immediate shutdown of Vermont Yankee because, in Mr. Mulligan's words, "the NRC and Entergy can't keep their nuclear safety paperwork and documents accurate and up to date." 

This is when they began shortening their explanation on my petition back on the Vermont Yankee. I admit this wasn't my best work product to the NRC. I never considered thinking about what is in their interest of shortening their explanation of my petition and the process. What is the agency's hidden agenda with it.

He also requested additional actions, which I will not state now unless someone would like me to on the call.

The management directive exactly means what is says: "allow you to provide any information that you believe the PRB should consider as part of the petition."

They were harassing me with trying control the content "any information" I was providing to the NRC.

Mr. Mulligan, I will now turn it over to you to allow you to provide any information that you believe the PRB should consider as part of the petition.

So all  the NRC's comments and questions were about controlling my speech and hurrying me up....none of the question were about asking me for technical clarifications about my petition.

Basically I got a little riled with what was happening to me ...I clicked a icon on my computer one too many times...my computer froze up on me. I had a set of talking points and I practiced what I was going to say on each point for two days... that is what this cost me. The rest of my presentation came from my head.

Entergy-Fitzpatrick Is Beginning To Be Unreliable?

Right, Pilgrim has had severe capacity factor issues this past this past cycle. Two shutdowns by bad SRV valves and numerous power restrictions.  Arkansas one  dropped a 600 ton stator…knocked off one plant for a month and we don’t know when they will get the second plant running…

I talked to the NRC about Fitzpatrick yesterday. Their power was up and down all this week on the NRC’s internet site. They are at 95% power today.  They are having “conductivity issues” according to the NRC. It sounds like their main condenser is leaking in river water and salt. They got little holes in the main condensers and the vacuum is sucking in the dirty water.  They got a bum main condenser?

Problems with conductivity issues

1)      It causes corrosion on the fuel pins and reactor components…leading to damage and fuel leaks.

2)      Can increase radiation levels throughout the plant.

3)      The replacement of the main condenser can be a 100 million to 400 million dollar job and takes many months to accomplished.

So are Entergy’s  philosophical “economic efficiencies”  and “budget problems” leading to unintended consequences? A drastic decline with their plants staying up at 100% and staying connected to the grid fixed easily preventable problems…  

Unprecedented Bellows Failure In A Pilgrim SRV

Lets see, so Pilgrim had a bellows failure on March 5, 2013, in addition to what caused all the leaks...

They reported it on May 3, 2013.

They safety reported it when Pilgrim was shutdown…just like the LER from Jan discussing the shutdown when in the refueling outage after. See, they are protecting Pilgrim with the timing of the release…making sure no interveners could request to shutdown…

You get it, the standard of safety for Target Rock and Entergy is they have to prove a not transparent component or internal is not safe.

They are not required to prove the material has the required quality for its safety requirements.

So we had some defect in these new valves that caused two plant required shutdowns, some eight leak. In addition to that, we have another defect in the bellows that would have prevented the valve’s automatic relief capability. And they don’t know the mechanism of the failure and the magnitude of it.

Right, that pilot valve was in the plant and then was heading back to the plant after testing…and there was no expectation the bellow corrosion would have been discovered in a regular turnaround.

I will give them the benefit of doubt, got to look at my timeline…it came from the Nemo Nor'easter when they replaced leaking

Holy freaken shit!

My April 2, 2013 petition 

MR. MULLIGAN: “I think that in the beginning, you were -- the NRC. I won't get personal, but the NRC was really rude to me and disruptive because they didn't know all the other ones went (other 2.206's of mine). This was off normal as far as how it was-- you disrupted me and got me off my thought train and stuff. And I interpreted it as being rude and disrespectful and part of a coverup. Thank you.”

INTERIM PART 21 REPORT OFPOTENTIAL DEFECT IN A RELIEF VALVE BELLOWS

March 3, 2013

Part 21 (PAR)

U.S. Nuclear Regulatory Commission Operations Center Event Report

Event# 48996

Rep Org: CURTISS WRIGHT FLOW CONTROL CO. Notification Date / Time:

Emergency Class: NON EMERGENCY ERIC DUNCAN R3DO

10 CFR Section: 21.21 (a)(2) INTERIM EVAL OF DEVIATION

INTERIM PART 21 REPORT OF POTENTIAL DEFECT IN A RELIEF VALVE BELLOWS

The following was excerpted from a fax:

(ii) Identification of the basic component supplied for such facility or such activity within the United States which may fail to comply or contains a potential defect.

Target Rock P/N: 303480-1, Bellows, Manufactured by Target Rock.

(iii) Identification of the firm supplying the basic component which fails to comply or contains a defect. Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation

(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.

During as-found steam testing on March 5, 2013 of a Pilgrim Main Steam Safety Relief Valve (MS-SRV) (TR Model 09J-001, valve assembly SIN 5, pilot assembly S/N 23, bellows PIN 303480-1 SIN 607) a loud pop was heard and as-found testing was secured. Subsequently, the pilot assembly was removed from the valve assembly and

subjected to a leak test and would not hold pressure. The pilot assembly was disassembled and a visual inspection of the PIN 303480-1 bellows convolutions revealed a through wall failure in one of the convolutions. It is noted the steam testing was performed at an offsite test facility and the valve did not fail installed in the plant.

The bellows acts as a pressure sensor responsible for initiating the opening of the MS-SRV at set pressure. Failure of the bellows does not directly impact the integrity of the Reactor Coolant System (RCS) pressure boundary, which is maintained by the bonnet assembly that surrounds it, but does impair the ability of the MS-SRV to provide over-pressure protection of the RCS. This technology has an extensive history of reliability in nuclear power systems and has been used in Commercial Nuclear Power Plants (NPPs) since the 1970s. This is the first reported incident regarding a thru wall bellows failure.

Target Rock initiated a comprehensive root cause evaluation pursuing several areas of investigation. In parallel, Entergy is conducting an independent investigation and we are cooperating with them. A complete review of our paperwork confirms all manufacturing procedures and processes were performed in accordance with all specified requirements. This includes:

- Raw material analysis

- Dimensional inspections

- Cleaning

- Heat Treatment

- Manufacturing processes

- Testing

- Review of design stresses

Preliminary metallurgical analysis of the failed bellows indicates cracks forming in an inter-granular manner as would be expected from Inter Granular Stress Corrosion Cracks (IGSCC) originating at pit like location on the interior pressurized surface. The source of this cracking is the focus of on going investigations. Target Rock has also visually inspected two other bellows of the same part number, one manufactured from the same material lot and another manufactured from an earlier material lot. Both of these bellows were installed in valves steam tested at Target Rock. One of these valves bellows was also full flow tested at Wyle Labs. Neither of these additional bellows contained pit-like locations and may indicate this potential failure mechanism is an isolated incident. However, to date, neither Target Rock nor Pilgrim can draw final conclusions with the information collected and analyzed.

The mode of failure has not been determined; however, in order to address the potential for a common mode failure, Target Rock is continuing metallurgical testing of the failed bellows and the two other bellows with the same part number. Based on these results, it is likely we will need to evaluate bellows that have been installed in other NPP as they become available.

(v) The date on which the information of such defect or failure to comply was obtained.

The as-found steam test and identification of the potential defect occurred on March 5,

2013.

(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part.

The following plants are running with bellows P/N 303480-1 installed: Limerick 1 & 2, Pilgrim, and J.A. Fitzpatrick.

(vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

The following plants are running with bellows P/N 303480-1 installed:

Limerick:  

28 valves installed,  3-Stage MS-SRVs, Units 1 and 2, with bellows installed 1999 

 Pilgrim:   

4 valves installed, 3-Stage MS-SRVs, with bellows installed in 2011  

Fitzpatrick: 

 3 valves installed, 3-Stage MS-SRVs, with bellows installed 2011

(vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

The root cause of the potential defect is not yet known as of the date of this report. Therefore, no specific corrective actions have been initiated. Target Rock Corrective Action Request CAR 13-013 will document the corrective actions when they are determined. This determination will be based on further mechanical and material evaluations. TR anticipates completing these evaluations within 45 days; however, in the event the evaluations are not completed, TR will forward another interim report within 45 days.

 (viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees.

Target Rock will recommend that the end user perform a detailed visual inspection of the interior convolutions of installed bellows P/N 303480-1 at the next opportunity to determine if any areas of pitting or cracking exist on the interior wails of the bellows. This is a difficult inspection to perform due to the following: internal geometry of the convolutions, a trained inspector is required and specific inspection technology is needed to yield reliable results, (ix) In the case of an early site permit, the entities to whom an early site permit was transferred.

Are Radiation Reports At nuclear Plants Not Trustworthy?

With Fitzpatrick and ANO...the NRC made entergy notify their fleet that falsification will not be tolerated.

Indian Point supervisor arrested on federal charges

WHITE PLAINS – The former chemistry manager at the Indian Point nuclear power plant was arrested on Tuesday and charged with falsifying records to conceal a violation of Nuclear Regulatory Commission requirements of the facility. Daniel Wilson, 57, of Walden, was responsible for making sure particulate matter in the diesel fuel used to power emergency generators at the facility did not exceed a set limit. In 2011, tests of the diesel fuel showed the particulate matter exceeded the NRC limit. In February 2012, Wilson concealed facts from his employer and the NRC by fabricating test data for non-existent resamples of the diesel fuel, falsely showing that the resamples of diesel fuel tested below the applicable NRC limit. Investigation revealed that no such samples were taken and the purported test data were fabrications. Later in February, Wilson responded to questioning from other employees at Indian Point in advance of an inspection by the NRC and wrote a report in which he gave a false explanation for the lack of supporting documentation for his fabricated tests. In a later interview with NRC personnel, he admitted that he had fabricated the test results so that Indian Point would not have to shut down. Wilson was charged in a two-count federal complaint with willfully violating rules of the NRC by engaging in deliberate misconduct and with making false statements in a matter within the jurisdiction of the NRC. If convicted, he could face up to seven years in prison.

...Palisades is along Lake Michigan in Van Buren County's Covert Township, 80 miles east-northeast of Chicago. It's had nine shutdowns since September 2011.

...Leaks have been an ongoing issue at Palisades, owned by New Orleans-based Entergy Corp., which shut down four times in 2012 and twice so far this year. Most recently, in February, the plant shut down for six days to repair a component cooling water heat exchanger and replace a damaged switch.

 This indicates massive issues with integrity with radiation readings at all nuclear power plant. I takes a lot of skill and high education to run a radiation department and a large nuclear utility!

So four plants, Farley, Perry,  Arkansas One and the Fitzpatrick...these are all employees employed or contracted by radiation protection departments. These are the guys who train, measure and record plant radiation and employee radiation levels. It is three large nuclear corporations. With this widespread national level of employee radiation related falsification...if a dire nuclear meltdown happened...could you trust the radiation employees like would develop for the public?

And what the hell is going at the Perry plant with the radiation protection department....just plain incompetence. I think the fish incident was maliciously trying to besmirch the reputation of the Perry plant and FirstEnergy. The employees are fed up with craziness of the bureaucracy of the Perry Plant and the hapless NRC. And the NRC sent in an unheard of special four man radiation protection inspection team in to the middle of the recent refueling outage and a fired whistle blower singing into the media on TV.

The Terrible Perry nuclear plant:

PERRY, Ohio - Safety issues at the Perry Nuclear Power Plant will be outlined Wednesday evening during a public hearing in Painesville by top Nuclear Regulatory Commission officials.

The hearing follows an exclusive 5 On Your Side investigation in February 2012 detailing concerns over radiation exposure and workers inside the plant.

Overall, the NRC found the Perry Plant operated safely in 2012 but it will remain under "increased NRC oversight" for the remainder of 2013.

The NRC also will announce an extensive series of inspections that will continue throughout 2013.

Wednesday night's meeting will be held from 6 p.m. to 8:30 p.m. at the Quail Hollow Resort, 11080 Concord-Hambden Rd. in Painesville.

At the time, NRC Regional Director Chuck Casto said, "it starts to worry you about the spread--are problems spreading...There needs to be procedure changes and they need to be sustainable."

...NRC: The NRC has sent four additional inspectors – in addition to the two Resident Inspectors – to the Perry Nuclear Plant in Ohio to watch and evaluate how the plant is ensuring the safety of these workers.

Sending these extra inspectors to monitor outage activities reflects the measure of our concern with Perry’s occupational radiation safety program – which is supposed to make sure workers don’t get exposed to unnecessary levels of radiation. The plant is under increased NRC oversight because of deficiencies in this program. Even though these issues have not resulted in any overexposures to workers, we want to make sure the plant fixes the weaknesses in this vital area.

In June, we will conduct a thorough inspection to determine if plant owner FENOC has understood the extent of the weaknesses in occupational radiation safety at Perry and has taken what we call “sufficient and sustainable actions” to fix the problems and prevent them from happening again.

These got to be smart people who did this. Sounds like the licenced operators sending t FirstEntergy a their special memo through the media. Get the metaphor...cooked with high radiation like steamed fish.  It is sophisticated  message.  They placed the fish there so they wouldn't be caught and trusted someone to report it. They chose this issue maliciously so it would question getting contraband by the security umbrella. They figured the fish story in a radiation place and radioactive water would get amplified in the media... media attractive. They knew it wasn't against the rules and knew where to get the radioactive water. A contractor couldn't do all that. That was thoughtful in such a lonely place...a boyfriend and girlfriend goldfish. They are making a cry for help!

...“Federal regulations do not prohibit workers from bringing goldfish into the plant,” Mitlyng said.  

...An investigation is underway by the U.S. Nucleaer Regulatory Commission after a pair of radioactive goldfish were discovered swimming in a lemonade pitcher in the steam tunnel of the Perry Nuclear Power Plant in Ohio.

The Perry Nuclear Power Plant is located about 40 miles northeast of Cleveland in North Perry, Ohio, and is situated near the coast Lake Erie.

Jennifer Young, a spokeswoman for the plant, said:

Farley speak for itself...

NRC: Farley Nuclear Plant security employees cheated on tests

May 8, 2013

WASHINGTON — Security workers at Farley Nuclear Plant in Alabama cheated on their training exams in 2010 and 2011, the Nuclear Regulatory Commission announced Tuesday.

NRC officials said Southern Nuclear Operating Co., which operates the plant, avoided civil fines by negotiating a settlement with federal regulators.

A testing proctor at the plant near Dothan gave security officers exam answers or took the test for them, according to the agreement. Managers at the plant, which is owned by Alabama Power Co., investigated the cheating incidents and disciplined the employees involved.

All employees at the plant are required to take radiation-worker training exams regularly. (Basically like training with Arkansas one and Fitzs.)

Rather than go through a formal NRC sanctions process for violating rules about testing security, Southern Nuclear opted for mediation. The company re-tested all employees, overhauled the testing process and began randomly observing workers as they took exams.

As part of the settlement, the company is studying ways to improve testing procedures at all of its nuclear plant sites and launching new training on professional integrity….

So this is Arkansas one in a March 13, 2013 NRC Inspection report.

(Basically training related and accident preparitness) Yea, you will get the accurate accident radiation levels to panic you sufficiently.)

During the NRC OI investigation, the senior emergency planner at ANO admitted to generating false documentation over a period of four years. The false documentation does not meet the requirement under 10 CFR 50.9(a), Completeness and Accuracy of Information. This regulation states, in part, that information required by the Commission's regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. The false documentation included 2 miscellaneous drills involving the Post Accident Sampling (PAS) system, as recorded on December 14, 2010 and December 7, 2011, and 2 drills involving environmental monitoring, as recorded on December 14, 2010, and December 6, 2011. The drills were required by the licensee's procedure number 1903.004, "Admin and Maintenance of the Emergency Plan and Implementing Procedures," which fulfills the requirement under 10 CFR 50.47(b)(14). In addition, it was determined that the senior emergency planner at ANO falsely documented 3 surveillances required by EP-010, "Emergency Response Facility Walkthrough Surveillance, Technical Support Center (TSC)" on May 12, June 4, and September 30, 2008. The TSC surveillance required checking the operation of the NRC management counterpart link (MCL) telephone line in the TSC. The false documentation indicated that the NRC MCL line was operable. The investigation determined that the NRC MCL line in the TSC was inoperable from February 2008 through November 2008. This surveillances was also required by the emergency plan, to meet the regulatory requirement under 10 CFR 50.47(b)(8).

Operators, firefighters, maintenance, chemistry and radiation employees wear the majority of the radioactivity related respirators and forced air masks. The majority of plant employee never wear respirators of any kind. It is policy that the fat asses of senior executives are never allowed in areas that need respirators.

You get it, they slit the career and family throats of of the employee at the bottom who are trying to make their jobs more efficient for the corporations...trying to please their bosses and get promoted...while the senior executives who oversee this only get a hand slapping and generally are held unaccountable.

4 fired, 34 disciplined at James A. FitzPatrick Nuclear Power Plant

September 09, 2011

A series of investigations at the James A. FitzPatrick Nuclear Power Plant in Scriba has resulted in four workers being fired and 34 being disciplined, a spokeswoman for the plant owner said Thursday.

Meanwhile, federal prosecutors announced that one of the fired workers has pleaded guilty to falsifying tests of safety equipment at the plant.

Also Thursday, the Nuclear Regulatory Commission notified the owner of the plant, Entergy Nuclear Northeast, that it could face civil actions in the wake of the investigations.

Results of the three investigations were handed over to the U.S. Attorney’s Office in Syracuse, which brought criminal charges against Michael McCarrick, 56, of Oswego, a former radiation protection technician at the plant. McCarrick admitted to falsifying records relating to more than two dozen plant worker.

In his guilty plea, he admitted he failed to adequately perform tests to make sure the workers’ emergency respirators were properly fitted and sealed, and then falsely documented that they were.

The plant is required to refit workers’ emergency respirators every year. The respirators, which protect against chemical releases or other fumes during emergencies, must fit snugly, and the fit can change as workers gain or lose weight or otherwise change.

Assistant U.S. Attorney Craig Benedict said that on 32 documented occasions between 2006 and 2009, McCarrick falsely claimed he had completed such tests. Benedict said the incidents were investigated by special agents from the NRC.

No known injuries occurred as a result of the falsified tests

McCarrick pleaded guilty to one felony count of violating the Atomic Energy Act. He could receive up to two years in prison and a $250,000 fine when he is sentenced Jan. 10.

The NRC also found that two unidentified “staff level individuals” acted with “careless disregard” by not following through on their suspicions that the respirator fit tests were inadequate.

The NRC informed Entergy that the second investigation found that McCarrick deliberately failed to document required surveillance of air samples or to make sure workers leaving the radiologically controlled area went through contamination monitors.

The third investigation found that McCarrick and another radiation protection technician failed to conduct other leak testing and surveillance duties.

McCarrick was the only worker charged by the U.S. Attorney’s Office . However, the NRC notified Entergy that it could face civil action pending the result of either an enforcement conference with the NRC or a mediation session, whichever the company chooses.

Entergy spokeswoman Tammy Holden said the plant conducted an internal investigation in June 2009 after learning the NRC had received a phone call alleging a potential violation at the plant.

Most of the 34 workers who were disciplined were workers who should have known that their “fit tests” for the respirators were either not done or were incomplete, Holden said. She said those workers were removed from the site during the investigation and later received either suspensions of pay or verbal or written warnings. They were also retrained.

Holden stressed that the masks are rarely required, and that none of the workers who had inadequate fit tests did any work during that period that would have required the masks to be used.

“At no point was there any risk to the public health or safety at any time during this process,” she said

Was Waterford A Precursor To Arkansas Nuclear One?

June 12:

ARKANSAS NUCLEAR ONE - NRC AUGMENTED INSPECTION TEAM

REPORT 05000313/2013011 AND 05000368/2013011

Basically the temporary crane was supposed to be tested with 125% of the weight of the stator...it wasn't.

You get it; the crane company isn't participating with the Entergy investigation...
You see how the Waterford event covered all the bases of the Arkansas event?

What i see is, they made the complexity of the heavy lift rules so massive...that no one can understand the extremely infrequent lift rules or follow them. The extreme complexity of the rules basically makes the heavy lift rules unenforceable and inherently dangerous.




So this caused the death of one employee, the injuries of eight, and rumors of the amputation of a leg.

Is the 2012 Waterford Inspection Report actually the 2013 dropped stator...

　
　
Is this how they game rules for short term advantage?

I suspect May 9 would go like, we overly depended on the contractors to do the heavy lifting...

November 14, 2012

SUBJECT: WATERFORD STEAM ELECTRIC STATION, UNIT 3 – NRC INTEGRATED INSPECTION REPORT 05000382/2012004

Heavy Lifts and Failure To Perform Proper Risk Assessment.

"Description. On August 15, 2012, the licensee conducted heavy load lifts over the B train of the dry cooling tower area in order to assemble portions of a temporary work platform (TWP) used to support steam generator replacement maintenance activities. The licensee used procedure EN-WM-104, “Online Risk Assessment,” to perform an initial risk for this activity while at-power and determined that the risk was a normal level with no additional risk management activities needed. The inspectors reviewed the online risk assessment and noted that the risk assessment associated with the lifts failed to identify that some of the activities associated with the assembly of the TWP met the definition of a non-standard lift. The risk assessment procedure EN-WM-104 identified that non-standard lifts should be considered as high risk and additional requirements for preparation, approval, and oversight of such activities are needed.

The inspectors noted that to determine if a lift is non-standard, the licensee should use procedure EN-MA-119, “Material Handling Program.” Numerous lifts associated with assembling the TWP met the definition of a “critical lift” given in EN-MA-119. Specifically, the lifts involved handling large equipment “over spaces in which high value or safety-related equipment or systems are located.” EN-MA-119 defines all critical lifts as non-standard lifts. However, the licensee used the contractor’s assessment of what constitutes a critical lift. The inspectors determined that the licensee did not follow processes in place to properly assess and manage the risk associated with performing non-standard lifts. Due to this failure, the licensee inappropriately categorized the activities as having normal risk, rather than high risk, when performing EN-WM-104. The licensee also noted at that time the licensee scheduled reactor trip breaker testing. However, since this activity was deemed normal, no other risk management actions were in place due to the inadequate assessment. The categorization of the activities as having normal risk resulted in the licensee’s failure to implement the more stringent risk management actions required by EN-WM-104 for high-risk activities.

The licensee entered this condition into the corrective action program as CR-WF3-2012-4195 and CR-WF3-2012-4489. The immediate corrective action taken to restore compliance was to re-evaluate and change the integrated risk classification from a normal risk to a high-risk level and implement the required risk management actions.

Nuclear Diaphragm Valve Product Line

Nov 7: Here is the update.

As with VY’s SRV’s actuator seals, a formulation had gone obsolete...so they replaced it with lesser materials.

So 10% of their customers need an equivalent or similarly durable material...ITT just threw these guys under the bus.   

What sticks out is how amazingly complex a issue this is over a simple diaphragm. Think about this with millions of components in the industry...

And a limited cycle rate or life time is mind blogging complex problem when you consider the 18 month operating time of our reactor and the limited opportunities to replace these guys while up at power in a radiation field.

How many venders are throwing the whole nuclear industry under the bus...

While the Ml test program was considered a success and the new compound was launched in 2008, there were still 10% of those conditions for which ITT was unable to meet the full qualification target of 7,500 cycles, and this was a source of some concern for certain ITT customers. Those customers had purchased the previous MI diaphragm made from the polymer that had gone obsolete, and they would contact ITT for replacement valves or diaphragms subject to those conditions. ITT would note that while the data did not permit the diaphragm to be sold for that condition, the diaphragm could be used at a restricted service life base on the limited data that did exist. This was a common practice that had been used successfully with the previous MI formulation, and customers with those certain operating conditions agreed to use the diaphragms with a specific reduced life.

May 15:
Yea, can you even imagine what a phone call by the NRC to ITT raising questions on the diaphragms. No matter what was said, it would imply the NRC is taking a interest in it and it's going into a  official investigation. They would be doing back flips trying to prepare for the upcoming investigation.

?

May 14:

I talked to two part 21 NRC specialist today. They were very friendly. They could have talked for months about the part 21 process. I told them I am not concerned with the radiation qualification of the diaphragms. I showed them what I thought were shortcomings in the part 21 process.     

I explained this is what I am concerned about:

Code Case N31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad).

1)      What is code case n31?

2)      The high rads imply the diaphragms are going into containment and the 250 degrees is incongruent with containment.  More like 370 degrees.   

3)      Specifically what does 250?F mean.

4)      The fundamental question,  is there non-qualified diaphragms in containment.

Basically they told me that I should trust the process, they do…I told them where there are current holes in the agencies part 21 process…so I am testing you guys.

They explained we need to wait until the 60 day update…that is when they will specify what valves and plants are involved. They seem really hesitant to call ITT. They didn’t seem concerned the 1E8 rads and 250 degrees in the containment didn’t mix.  But they said they would get answers for me. They said they will call me back when the update came in.

I just was surprised they couldn’t immediately tell me these valves don’t go into containment or nobody knows what the 250?F means, but all diaphragms are good to 370 degrees. The best they would say is we don’t know what "code case n31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad)” means. I was surprised they couldn’t throw some facts back at me to blow me out of the water.

May 3: SO, a Watts Bars inspector called me up about the ITT diaphragm issue.

I told him this reads as qualified for class 1 nuclear safety components and 1E8 implies it is near the core. The "250?F" implies that is the highest temperature rating. Told him it should be 370 degrees qualified. Reminded him about the Fort Calhoun, Peach Bottom and Vermont Yankee issue with putting in intentionally improper material in valves. This looks to me like to me they are putting improper diaphragms in nuclear plant containments in general. He told me they have rules and regulations to prevent this, besides it was caught in a surveillance. I told  him that was a hell of a way to run a Navy.

The only way the agency is going to Catch it is by doing a inspection on containment diaphragms and the NRC then asking a set of utilities if all containment diagrams are fully environmentally qualified. 

"It only applies to those that were sold for a particular service condition of Code Case N31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad)."

But he seem more interested on how I got the phone number of the first inspector. It must be a secret phone number or such. This is two NRC officials that seemed to be really interested in how I got this special phone number.

He wasn't even familiar with what ASME code case N31 means...

I think it defines the quality of class I, II and III elastomer 3 or 4 inch diaphragms......Big picture, based on the VY safety relief valves threaded seals and current events in the industry, the NRC lost control of maintaining the design quality and environmental considerations within the containment to withstand the worst accident of design. They don't know how the plants will respond in a accident with their containment at 370 degree to 400 degree. And this is all wrapped up the TVA Watts Bar's commercial dedication crisis...the inability of vendors to provide quality replacement parts in the industry. It is in every plant in the nation.

I asked the NRC at VY's annual public meeting and they seemed to promise me a response:

1) Does Vermont Yankee have any code case n31 diaphragms in Vermont Yankee, specifically diagrams only qualified to 250 degrees?

2) Does any plant in the USA have code case n31 diaphragms in their plant, specifically diaphragms only qualified to 250 degrees, which should be 370 degrees... 

I have complained about components in nuclear plants being not qualified for the designed accidents in containments environments.

I specially complained about pneumatic safety relief valve o rings, seals, gaskets and diagrams. I worried they wouldn't be qualified for the containment environments...specifically for temperatures and radiation.

Code Case N31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad):

Based on the VY SRV unqualified buna-n threaded seals, 250F isn't qualified for any containment in the USA. They should be qualified for 400 degrees F.

They could be for pneumatic actuators?

Devices that measure flow and difference of pressure (d/p)...

What the hell does "250?F" mean? What does "Code Case N31" mean and are diaphrams that meet this allowed in containments.

So why is the requirements 250 degrees and 40 year radiation exposure of 1E8 Rad?

ASME Section III Component Replacements 

N31 (1540-2)
Elastomer Diaphragm Valves, Section III, Class 2 and 3

7/18/85-Each applicant who applies the Code Case should indicate in the referencing safety analysis report that the service life of the elastomer diaphragm should not exceed the manufacturer’s recommended service life. This recommended service life should not exceed 1/3 of the minimum cycle life as established by the requirements of paragraph 3 of the Code Case. In addition, the service life of the elastomer diaphragm should not exceed 5 years, and the combined service and storage life of the elastomer diaphragm should not exceed 10 years. 

Class 1Components (III, Subsection NB)-Those components that are part of the primary core cooling system 

Components (III, Subsection NH)-Those components that are used in elevated temperature service

Class 2Components (III, Subsection NC)-Those components that are part of various important-to-safety emergency core cooling systems

Class 3Components (III, Subsection ND)-Those components that are part of the various systems needed for plant operation

When you worry about this kind of  rads for1E8, you are talking about power operations near the core and the potential of nuclear meltdown. That be 400 degree F.

Sounds like this comes from TVA, Watts Bar and not qualified nuclear parts...

Rep Org: ITT ENGINEERED VALVES, LLC Licensee: ITT ENGINEERED VALVES, LLC Region: 1 City: LANCASTER State: PA County: License #: Agreement: Y Docket: NRC Notified By: STEPHEN DONONHUE HQ OPS Officer: BILL HUFFMAN	Notification Date: 04/26/2013 Notification Time: 17:25 [ET] Event Date: 04/26/2013 Event Time: 13:54 [EDT] Last Update Date: 04/26/2013
Emergency Class: NON EMERGENCY 10 CFR Section: 21.21(d)(3)(i) - DEFECTS AND NONCOMPLIANCE	Person (Organization): JUDY JOUSTRA (R1DO) MARVIN SYKES (R2DO) DAVID HILLS (R3DO) JACK WHITTEN (R4DO) PART 21 GROUP (RX) (E-MA)

Event Text

DIAPHRAGMS MAY NOT BE QUALIFIED FOR SPECIFIC RADIATION DESIGN CONDITIONS The following report was received from ITT Engineered Valves, LLC via facsimile: "It is my duty as the Responsible Officer of ITT Engineered Valves, LLC (ITT) to inform the Nuclear Regulatory Commission of a defect with certain items of our nuclear diaphragm valve product line which may be considered Basic Components. The components are ITT's Nuclear M1 diaphragms, sizes 3 inch and 4 inch that may have been sold to certain customers for specific design conditions. The defect does not affect all 3 inch and 4 inch M1 diaphragms that have been sold. It only applies to those that were sold for a particular service condition of Code Case N31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad). "The nature of the defect is best described by 10 CFR Section 21.3 Defect Definition #5, as 'an error, omission or other circumstance in a design certification or standard design approval that... could create a substantial safety hazard.' In this case, ITT inadvertently qualified the 3 inch and 4 inch M1 diaphragms for a design condition that includes the effect of radiation when in fact our recommendation was erroneously based on diaphragm testing that did not include irradiated diaphragm test results for those sizes. The potential safety hazard stems from the fact that if one of these diaphragms sees radiation in this particular service, there is no data to indicate that the diaphragm will perform its function in that service condition. Until such time that we can conduct additional irradiated diaphragm testing to additional sample diaphragms and test for this condition, we need to consider the parts that are in this service as potentially unsafe. "ITT is in the process of identifying all facilities for which the diaphragms were sent, either as spare parts or diaphragms incorporated into valve assemblies. We are also preparing to do further verification tests of the 3 inch and 4 inch M1 diaphragms in an attempt to ascertain the true performance rating at the noted condition. "Per 10 CFR 21 policy guidelines, this initial notification will be followed by a written notification by May 27, 2013."

Exelon-LaSalle's Two Plant Trip Is A Mess Over A Lightening Strike (really, lightning)

I am just saying...a known not corrected defect in the industry just lead leads to worst and worst outcomes.   

Vermont Yankee switch yard insulator defect was a precursor to the double hitter Byron insulator failure.  Think about how much money this cost by not forcing a industry wide response.

Vermont Yankee Switch Yard Insulator Defect LER 2005-001-00

 "A root cause investigation team determined that the MOD failure was caused by the failure of a porcelain electrical insulator as a result of a manufacturing defect. A laboratory examination of the insulator was performed by an off-site lab. The examination revealed a void area in the cement that attached the failed section of the insulator to the metal flanges and a geometric off-set in the placement of the insulator in the flanges. Close examination of the void surfaces showed that this void was pre-existing and occurred during the manufacturing of the assembly. These conditions caused a stress riser to occur on the northwest side when wind and other cyclic loads were applied to the insulator. The repeated cyclical loading and unloading produced a stress crack in the porcelain, weakening the insulator and ultimately leading to failure, prior to it's design lifetime of 40 years. The insulator was original plant equipment."

Byron Switch Yard Insulator Defect LER 2012-001-01

Cause of the Events Event l
The Unit 2 SAT-1/2 insulator failure was caused by service propagation of a large manufacturing material defect that covered approximately 40% of the fracture cross-section in one section of the insulator stack. The defect was characterized as poorly vitrified porcelain, which contained a high density of porosity and micro-cracks. 

Additionally design vulnerabilities existed in the protective relaying schemes regarding the lack of single open phase detection that complicated plant and operator response by not automatically isolating all three phases on the affected line.

Event 2
The Unit 1 SAT insulator failure was caused by service propagation of a large manufacturing material defect that covered approximately 25% of the fracture cross-section in one section of the insulator stack. The defect was characterized as poorly vitrified porcelain, which contained a high density of porosity and micro-cracks. Moreover, a second insulator section, which fractured as a result of the fall, exhibited the same poor vitrification as did the section that initially fractured.

Sounds like a thunderstorm wind could undermine the structure of a defective switch yard insulator?

VY: "These conditions caused a stress riser to occur on the northwest side when wind and other cyclic loads were applied to the insulator. The repeated cyclical loading and unloading produced a stress crack in the porcelain, weakening the insulator and ultimately leading to failure, prior to it's design lifetime of 40 years."

Can't start the plants up sequencually ...what not try and start them up at the same time?

Exelon Illinois LaSalle reactors 1 And 2 ramped up early Wed.

Wed May 1, 2013 9:03am EDT

May 1 (Reuters) - Exelon Corp's 1,118-megawatt Unit 1 at the La Salle nuclear power plant in Illinois ramped up to 70 percent power early Wednesday from 22 percent power Tuesday, the U.S. Nuclear Regulatory Commission said in a report.

Its 1,120-megawatt Unit 2 at plant was operating at 28 percent power, up from 1 percent power on Tuesday.
 

...Originally posted on April 18...

Exelon's Guatemala and oligarchy fleet of nuclear plants...  

NRC: “When we looked at all lightning-related events at U.S. nuclear power plants from 1992 to 2003, we identified a total of 66 such events, he said. “Twenty-one of those involved a loss of one or more offsite power sources but no equipment damage. There were no events that involved a loss of offsite power that resulted in plant equipment damage. Of the 66 events, 48 (or about 73 percent) involved no reactor trip, or shutdown.”Sheehan added, “Most lightning strikes do not cause a plant to shut down.”

So let me get this straight...the LaSalle facility has two nuclear power plants. They got a so called lightning strike on April 17 leading a special NRC inspection. It tripped both plants. The special inspection team is on site right now. They first attempted to start-up Unit 2 last week (April 25). They had a bad circ water pump or something, it took out the plant. Over the weekend with the NRC on site, they attempted to start-up Unit 1. They had a RCIC steam line leak in the containment and it caused them to shut down on Saturday. 

So Thursday Unit 2 was shut down, then within two days Unit 1 was shut down for a steam line leak. Today both plants are shut down again while the NRC is on site for a special inspection. Buddy, the heat is on. Is Exelon collapsing like they did in the middle and late 1990s...   

It is kind of amazing, on restart both plants had to be shut down before they even reached 100% power...within just hours of the startup. Two plant startups within the last few days and double 0% power in today's NRC's current reactor status report. 

Sequentially in early 2012 both Byron 1 and 2 tripped within a month of each other on bad switch yard insulators, led to LOOPs...a bad protective safety circuit created a severe vulnerability for both plants. The NRC issued a serious Guatemala engineering style warning to all plants to check their LOOP protective circuits. This is how a two nuke plant facility would behave if they were located in Guatemala?   

Exelon issued a recent severe financial warning over wind and natural gas making their nuclear fleet extremely vulnerable. Huge nuclear plant budget cutback and extreme employee disillusionment! Had to cut their dividend also! And Exelon is home-ported in Chicago Illinois and its one of the most politically corrupted states in the USA.    

Japan Times wrote a article about severe and dangerous employee issues at Byron...the culture of intimidation at Exelon. What crap these employees have to go through to make more than $100,000 a year. If they were making minimum wages. they could just say fuck this stuff and quit. They are all slaves to feeding their families and making money...having a career.     

For eight minutes, you’ve raised your middle finger to the meltdown gods,” one reactor operator said, speaking on condition of anonymity. “If anything else happened in that window — and it’s a safe bet one problem causes another — you’re screwed.”

“Those eight minutes symbolize over a decade of abuse,” said a plant source. “And you can never undo it. And it’s never forgotten.”

So in a LOOP over the failed insulator without diesel generators, Byron failed to automatically start up their emergency diesel generators for eight minutes. You get it, the extremely technical Exelon employees are pleading and crying for help? 

Gets you wondering if Exelon and Entergy are exhausting and overwhelming the NRC? 

Power Reactor	Event Number: 48977
Facility: LASALLE Region: 3 State: IL Unit: [1] [ ] [ ] RX Type: [1] GE-5,[2] GE-5 NRC Notified By: JIM SPIELER HQ OPS Officer: JOHN SHOEMAKER	Notification Date: 04/28/2013 Notification Time: 00:48 [ET] Event Date: 04/27/2013 Event Time: 21:24 [CDT] Last Update Date: 04/28/2013
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): DAVID HILLS (R3DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	6	Startup	0	Startup

Event Text

TECHNICAL SPECIFICATION REQUIRED PLANT SHUTDOWN "This notification is being provided in accordance with 10CFR50.72(b)(2)(i), Plant Shutdown required by Technical Specifications, and 10CFR50.72(b)(3)(ii)A, Degraded or Unanalyzed Condition. "At 2245 CDT on 04/27/13, LaSalle Unit 1 commenced a Technical Specification required plant shutdown, due to identification of pressure boundary leakage. At 2124 CDT on 04/27/13, a through-wall leak was identified in the body of 1E51-F076, Reactor Core Isolation Cooling system steam supply inboard isolation bypass warmup valve. This qualifies as pressure boundary leakage, which requires entry into Technical Specification 3.4.5, Reactor Coolant System Operational Leakage, Required Action C, to be in Mode 3, Hot Shutdown, by 0924 [CDT] on 04/28/13, and Mode 4, Cold Shutdown, by 0924 [CDT] on 04/29/13. This leakage is significantly less than 10 gpm and therefore does not meet the threshold for entry into the Emergency Action Plan. At the time of discovery, Unit 1 was in startup mode following a forced outage. A unit shutdown has been initiated. A repair plan is being prepared at this time, and the unit will remain in Cold Shutdown until repairs are complete." The leak is located inside the primary containment and was visually identified during a containment walk-down. The licensee has notified the NRC Resident Inspector.

April 26:

Special investigation and another trip on unit two?

They are sticking with the lightning  strike...

The issue involved a lightning strike that resulted in the loss of external power to Unit 1 and 2. Both units automatically shut down and all control rods were inserted.

April 19:

“When we looked at all lightning-related events at U.S. nuclear power plants from 1992 to 2003, we identified a total of 66 such events, he said. “Twenty-one of those involved a loss of one or more offsite power sources but no equipment damage. There were no events that involved a loss of offsite power that resulted in plant equipment damage. Of the 66 events, 48 (or about 73 percent) involved no reactor trip, or shutdown.”
Sheehan added, “Most lightning strikes do not cause a plant to shut down.”

"So what would prevent all the defective switchyard insulators from breaking and shorting in a heavy storm, tornado or earthquake and then causing another LOOP..."

The LaSalle Nuclear Plant Cooling Lake and NRC Idiocracy
　I meant to say, what would prevent more strategic insulators from breaking in the weather causing more LOOPs?

It would have been nice to included thunderstorms...

Another two plant LOOP...Loops going on all over the place. On-site disconnection of off site electricity.

Right, whatever happens to their nuclear plant no matter what the evidence, it is always a total act of god.

Based on Exelon,  Briadwood and Byron, I'll bet you it was high voltage switchyard bad insulators... a thunderstorm wind came up and broke a defective insulator. The winds shook the lines/insulators and then broke the lines. Then a humorous short damaged the rest...

The theme is these plants aren't designed for the climate and the poor maintenance that doesn't address obsolete and defective component.

Towers and lines get hit all the time by lightening...and there is not much plant trips on lightening and damaging insulators nationwide.

I'll will bet a protective switchyad relay didn't work....

"This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side."

Facility: LASALLE Region: 3 State: IL Unit: [1] [2] [ ] RX Type: [1] GE-5,[2] GE-5 NRC Notified By: DAN SZUMSKI HQ OPS Officer: DONG HWA PARK	Notification Date: 04/17/2013 Notification Time: 16:59 [ET] Event Date: 04/17/2013 Event Time: 15:11 [CDT] Last Update Date: 04/18/2013
Emergency Class: UNUSUAL EVENT 10 CFR Section: 50.72(a) (1) (i) - EMERGENCY DECLARED 50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): STEVE ORTH (R3DO) JEFFERY GRANT (IRD) DAVID SKEEN (NRR) JENNIFER UHLE (NRR) ANNE BOLAND (R3)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	A/R	Y	100	Power Operation	0	Hot Shutdown
2	A/R	Y	100	Power Operation	0	Hot Shutdown

Event Text

NOTIFICATION OF UNUSUAL EVENT DECLARED DUE TO LOSS OF OFFSITE POWER FROM A LIGHTNING STRIKE "LaSalle Unit 1 and LaSalle Unit 2 have both experienced an automatic reactor scram, in conjunction with a loss of offsite power. This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side. "The plant systems have all responded as expected. All five diesel generators started, and have loaded on to their respective buses as designed. All rods went full in on both units during the respective scrams. HPCS [High Pressure Core Spray] system was started on each unit and automatically aligned for injection for initial level control." The MSIVs [Main Steam Isolation Valves] are shut on both units with decay heat being removed via the safety relief valves. Suppression pool cooling is in progress. The licensee will notify the NRC Resident Inspector and has notified the State. Notified DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email. * * * UPDATE FROM DON PUCKETT TO VINCE KLCO AT 2113 EDT ON 4/17/2013 * * * "In addition to information [previously provided], LaSalle Unit 2 received a high drywell pressure signal [1.77 psig] due to loss of containment cooling from the loss of power. At the time of this high drywell pressure signal, high pressure core spray pump and 2B residual heat removal [RHR] pump was already in operation, the low pressure core spray system and 2A residual heat removal system was secured and [placed] in pull to lock. When the signal was satisfied the ECCS [Emergency Core Cooling Systems] signal was processed but only the 2C RHR pump would have started. In this case, the 2C RHR pump tripped when the signal was received. There is no evidence of reactor coolant leakage. There was no additional ECCS systems discharging into the RCS [Reactor Coolant System]. As [initially stated], level was controlled using High Pressure Core Spray and level control is now being maintained using the Reactor Core Isolation Cooling [RCIC] systems. The 2C RHR pump trip is under investigation. "Due to the initial loss of offsite power for both Unit 1 and Unit 2 reported at 1511 [CDT], multiple containment isolation valves isolated and closed as expected. Once initial containment isolations were verified, two Unit 2 primary containment vent and purge valves were opened to vent the Unit 2 containment. Once Unit Two containment pressure reached 1.77 [psig], these two vent valves isolated as expected. "Due to the loss of offsite power, the Station Vent Stack Wide Range Gas Monitor (WRGM) and the Standby Gas Treatment Wide Range Gas Monitor (VGWRGM) also lost power. Manual sampling has been implemented and power is restored to the VGWRGM, however the VGWRGM has not been declared operable yet. Normal radiation levels have been reported from the manual sampling. [This is being reported in accordance with 10CFR50.72(b)(3)(xiii).]" The licensee notified the NRC Resident Inspector and the State of Illinois. Notified the R3 IRC, NRR EO(Skeen), IRD MOC (Grant). * * * UPDATE AT 0057 EDT ON 04/18/13 FROM MIKE LAWRENCE TO S. SANDIN * * * "After the Unit 2 primary containment vent and purge system isolated on the Unit 2 containment High Pressure signal, Venting of the Unit 1 primary containment was commenced. At 2005 CDT, Unit 1 primary containment pressure reached the Group 2 primary containment isolation system setpoint (1.77 PSIG) causing the primary containment vent and purge valves being used to vent the Unit 1 containment to isolate. Unit 1 primary containment venting was being performed through the Standby Gas Treatment system which is a filtered system. "In addition to the primary containment isolation signal on high drywell pressure, an ECCS initiation on high drywell pressure also occurred. The ECCS signal resulted in an auto start of the 1C RHR system. The 1B RHR system was already running in suppression pool cooling mode. 1A RHR and LPCS had been secured to prevent overloading the common diesel generator for division 1. The common diesel generator supplies both Unit 1 and Unit 2 division 1 ESF busses." The licensee informed the NRC Resident Inspector. Notified NRR EO (Skeen), IRD MOC (Grant) and R3IRC (Louden).