Are your “nads” safe: Seeing the Big Picture with the NRC and Palisades in One Design Inspection Report?

I am a mind reader? This just came out on the internet site Dec 17. See you there.

PUBLIC MEETING ANNOUNCEMENT
Title: Meeting to Discuss Effectiveness of CDBI Inspections
Date(s) and Time(s): January 13, 2015, 12:30 PM to 03:30 PM
Location: NRC Three White Flint North, HQ-3WFN-9A32
11601 Landsdown Street
Rockville, MD
Category: This is a Category 2 meeting. The public is invited to participate in this meeting by
discussing regulatory issues with the Nuclear Regulatory Commission (NRC) at
designated points identified on the agenda.
Purpose: To conduct a public meeting with the industry and staff to share each organization's
perspective on the Effectiveness of the Component Design Bases Inspections

Dec 10:

So I got a question, why after that first CDBI inspection didn’t the NRC tell Entergy if we find another “even one more CDBI violation” we will severely punish you with a prolonged shutdown. Why didn’t Entergy do heavy duty scrub on all their licensing and design issue so they never had another violation? 

How did it ever come to the point where it turned to it’s the NRC responsibility to find CDBI violations at Palisades?  

Personally I think Palisades has become obstinate over finding and correcting CDBI issues.

updated Dec 10:

How did we get here to a component design bases inspection? It came out of the Maine Yankee debacle. The regular inspector staff weren't catching all the violations...the NRC was forced to bring on a heavy duty inspection team for political reasons leading to uncovering a lot of missed violations and the premature shutdown of the plant.  

OCTOBER 3, 2007 

STATEMENT BY DALE E. KLEIN, CHAIRMAN UNITED STATES NUCLEAR REGULATORY COMMISSION TO THE SENATE COMMITTEE ON ENVIRONMENT AND PUBLIC WORKS SUBCOMMITTEE ON CLEAN AIR AND NUCLEAR SAFETY

“Although the Commission is confident that the Reactor Oversight Process is superior to the Maine Yankee Independent Safety Assessment, we continue to improve the process. For example, in 2006, the NRC staff, at the direction of the Commission, significantly enhanced the way the NRC reviews design issues. The resulting Component Design Basis Inspection procedure, which is an important element of the Reactor Oversight Process, is a team inspection to verify that design bases have been correctly implemented for selected risk significant components and that operating procedures and operator actions are consistent with design and licensing bases.” 

Here is a list of NRC component design bases inspection report since 2007. That is when they must have started these inspections. So we got 30 design and licensing violations since 2006 ...the inability to ever meet their plant licensing commitments. 

I think it is the job of the engineering department and the NRC to provide a pristine licensing environment for the licensing operators up in the control room. More, the plant should be perfectly congruent with licensing and we should alway drive the plant licencing towards sustained safety. An operator walks into work for the day, all plant licensing conditions should be pristine and all of the engineering should make sure the components operate better than as designed. The plant needs to keep up in modernity. These staffs should dedicate their lives to support the operating staff. Always, if the plant had any transient or accident...basically in a accident, there should be no surprises with broken equipment and events should never occur outside the well worn path of plant licensing. There never should be any surprises for the operators in any accident and certainly we shouldn't have any events outside licensing and training...plant designs should always be informed with the lessons learned through the history of the plant and throughout the industry. Anything less is a abdication to the dedication of the operating crew...you are setting up the licensing staff to fail. So with any violation of the Maine Yankee Component Design Basis inspection you are screwing the operating staff.

So there were 30 CDBI violations over the course of the five CDBI inspections. Obviously the NRC was doing a political CDBI inspection…partial inspections with the phony idea of clearing out all Maine Yankee CDBI like violations at Palisades or any plant. A type of inspection the NRC implied they were doing would look like this; maybe 30 violations on the first inspection and maybe one or two each of the rest the inspections.

I don't understand why it takes a special CDBI inspection team…is the resident inspector prohibited with uncovering CDBI violations? You get it, the residents are artificially prohibited from being involved in CDBI, 50.59 and licences amendment request violation. They have to bring on a different set of NRC inspectors...I consider this as artificially limiting violations discovery and reporting.

The middle CDBI inspection in 2008, 2009 and 2011 with (4,2 and 4 violations ending in 11 violations in 2014) indicates the agency inappropriately limited the scope of the middle inspections or limited the number of violations the inspector could report in 2008, 2009 and 2011 time frame? 

Right, as the behavior of Palisades was worsening in the middle years of this, the NRC was loosening the screws on Palisades right up to the late Sept 2011 yellow finding DC accident. Palisades was perilously declining all through the prior year and continued on to the other side of the yellow finding and the NRC was looking at them through a inappropriately loosening lens. The NRC needed the high hurdle of a unexpected set of incidences, accidents and plant trips before they could intervene to stabilizes the plant.  

Was the NRC signaling regulatory laxity in the head winds of Palisades worsening behavior, in the lead up to Sept 2011? Was Entergy keying off that and destroying their plant safety culture?

If the agency took a really really hard stance on the CDBI violations…would Palisades have corrected their behavior way before Sept 2012 yellow finding?

These plants with all their components are tremendously complex and the majority of the safety components sit behind a barrier, with the components in operation we are mostly blinded...we can't see what is going on. Of even more complexity, all NRC and plant rules, policies, procedures and multiple organizational complexity. All of this in totality is really not understandable...can't be comprehensively understood in any information system provided to us. It is just the fact of life. At the bottom of this, it a easily apparent in this Palisades inspection and the totality of the component design inspection to date...the massive complexity dwarfs the capabilities of the staff of the NRc and Entergy. If the staff's resources dwarfed the complexity of system, all the known and unknown design component violations would have been cleared out of violations on the first inspection and their would have only been less than a handful after the first inspection to date. If the Palisades and NRC staff dwarfed the complexity of the system all through the design, construction and operations of the plant, then there never would have been the necessity of ever having a Component Design Basis Inspection violation at all...Maine Yankee would have been stall running and there would have never been such a thing as a CDBI.

You can’t understand how we go to the Component Design Basis Inspection and the 2007 commissioner Klein speech, unless you understand the nuclear industry's turmoil all throughout the 1990s and leading up to beyond the 2001 David Besse head problem. 

1992 -I got fired from VY for raising safety issues?  

Jan 1993 -President Clinton  

1993 -Paul Blanch leaves Millstone  

Nov 94 -midterms, house and senate now controlled by republicans  

1995 -Shirley Jackson becomes NRC chairman June 96 -Galantis leaves Millstone and three Units shutdown 

Nov 96 -Clinton reelected with 49% Dec 96 - Maine Yankee shuts down permanently  

Nov 97 -Commonwealth Edison's whole nuclear system on fire and burning in Chicago, largest owner of nuclear plants in nation 

*1998 -Permanent shutdown of two unit Zion  

1999 -Jackson run out of office 

July 1 -Jackson became president of Rensselaer Polytechnic Institute 

Jan 2001 -President Bush 

2002 –Davis Besse head

Do you know what complacency and negligence cost us, make bad actor pay the price with taking shortcuts... it massively drives up complexity not making one more kilowatt for our work. Think of the additional complexity Maine Yankee brought us, it was the symbol nationwide with licensees not maintaining the licensing conditions in their plants and there was tremendous amount of plant operational problems. Think of all the complexity the CDBI brings to us with a gaggle of inspectors dedicated to these inspections, the procedures of the NRC for this, all officials who manage this and the interaction with the licensee, and the effort of the licensees himself. Not one extra kilowatt is made from this. How much does negligence and ignorance really cost us? I think this is model with driving up most of the unnecessary complexity of the whole system...this is the price we a pay with a lesser perfection and the lack of appreciation of a beautiful world. This is the world that short term profits brings us and the quarterly financial reports. A beautiful world is a economical world on the long run, and god serves us all.    

Remember the complexity is so large, I believe the totality the NRC and Palisades staffs and our sampling regulator...we only see and clear a insignificant amount of the licensing and component/procedure defects. So we got 30 CDBI violations to date at Palisade, they must have cleared out a lot of violation before they invented the CDBI inspections and the Maine Yankee shutdown...what violation do we have left in the life of the Palisades? Maybe, thirty, fifty or a hundred... Who is to say, will they are catch all of them before end of life.

Yous see what the staff is facing with obsolete plant? Palisades has never been a pristine operational plant...this plant has been very troubled over decades. This never has been a pristine plant, they have been diry all their lives...this is not about nitpicking the plant on minor licensing and maintaining the quality of component issues here. 

Remember the profound troubles with Palisades and all this massive attention by the NRC to Palisades over the decades is a zero sum game, especially with limited NRC resources...it steals NRC resources from other plants. It allows other plants to decline unseen and get into big troubles. Actually, this is how the NRC explained their actions in Davis Besse. Other bad actors and troubled plants in region III blinded us to the decline of Davis Besse. The refrain was, our inadequate agency processes prevented us from seeing the true conditions of the Davis Besse staff, we through DB was one of the better running plants, we then stuck our extra resources into the other troubles region III plants. We didn't have adequate inspection services on the approach to DB hole in the head at the site and our overwhelmed inspection services and limited agency processes didn't allow the agency see the decline of the plant into the hole in the head.   

And we know, out the NRC's OIG report on San Onofre and events at Millstone, the lot of NRC officials up and down the levels of the NRC spoke over and over again about not having the capabilities to see the big picture because of a artificial ideological limitations on NRC inspection resources and their unconscionable sampling philosophy that only allows them to see less than 10% of the field of play with the little understood complexity of the technology and government. How we defaulted to this kind of regulator. You know that don't you; we are not just talking about the complexity of Palisades...basically the complexity of the NRC organization and their regulation and procedures. It is absolutely clear so much of this complexity is in the NRC itself and this dissociating caused by musical chairs in the commissioners office in recent years, one side of the NRC brain doesn't know what is in the other side of their brain, the hemispheres aren't talking and communicating with each other.   

Last cycle, what if they had a big accidents? In a accident, it really stresses the components and employees of the plant as never before. All of the 11 violation predisposes to the staff that lots of components will fail in a unexpectant manner in an accident. The violations might give inaccurate indication to the operator in the accident. All these violations and the ones not knowable and fixed, half blinds the operating crew. They are facing a indeterminate amount of failures and control room confusion in big accident. Do you think the staff is at the top of their game or dispirited and demoralized? That will play a huge part in this. A large proportion of the current and future violation are about detecting and correcting the declines of the components in a obsolete plant. I think we are here with a new kind of accident not acted upon by the NRC. Honestly, do you trust government...

I advocated replacing on a emergency bases the 20 oldest plants in the USA with new plants. You know what Fukushima tells us, it would have been tremendously more economical to replace the Daiichi plant and similar plants with new plants. Prior to the accident there just isn't enough money to replace the plant, after the accident it all looks like a once in a lifetime opportunity. You say it is too expensive and the burdens of doing it are too high a hurdle...history teaches us we have no idea what the true cost of complacency is. In the rear view mirror of history, we have wasted so much treasure and opportunity. Honestly, I know this is hard to swallow. i know the system has pretty much got us checkmated on shutting down dangerous plants...maybe if we gave them the opportunity of a future they would do the right thing. i think it is a lot more dangerous if we let the industry decay away as we envision it today. Inaction,complaceny and political gridlock cost us so much.              

I have recently talked to a NRC 50.59 expert in region 1, he reminded me the early plants had skimpily plant licensing and their documentation was atrocious. We should always be thinking of that with these old plants. Keep in mind, the NRC is a shallow sampling regulator. They pick a number sample set...we have no idea of the real size of the cohort of all violation known but not recorded or unknown. They choose what they want us to see so all in the interest of their altruism.

Basically for the most recent component design inspections, these aren't new violations. These are violations and licensing defects that have been mostly around for decades and many come from before the plants first startup. Is this the third world of Guatemala where we are too poor and corrupt to develop their infrastructure...or are we the reflection of the best and brightest nation on the planet. Who are we? We have so much capability and opportunity wasted.      

December 2, 2014 : SUBJECT: PALISADES NUCLEAR PLANT COMPONENT DESIGN BASES INSPECTION 05000255/2014008 

11 violations  

September 12, 2011 SUBJECT:PALISADES NUCLEAR PLANT COMPONENT DESIGN BASES INSPECTION AND TEMPORARY INSTRUCTION 2515/177, “MANAGING GAS ACCUMULATION IN EMERGENCY CORE COOLING,DECAY HEAT REMOVAL, AND CONTAINMENT SPRAY SYSTEMS REPORT” 05000255/2011009 

Four violations 

January 15, 2009: SUBJECT: PALISADES NUCLEAR PLANT NRC COMPONENT DESIGN BASES INSPECTION (CDBI) INSPECTION REPORT 05000255/2008009(DRS)

Two violations 

May 19, 2008: SUBJECT: PALISADES NUCLEAR PLANT NRC FOLLOWUP INSPECTION 05000255/2008008(DRS)

one violations

February 13, 2007 SUBJECT: PALISADES NUCLEAR PLANT NRC COMPONENT DESIGN BASES INSPECTION (CDBI) REPORT 05000255/2006009(DRS) 

11 violations

Based on the results of this inspection, ten NRC-identified findings of very low safety significance were identified. The findings involved violations of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy. Additionally, one licensee-identified violation is listed in Section 4OA7 of this report.

December 2, 2014: SUBJECT: PALISADES NUCLEAR PLANT COMPONENT DESIGN BASES INSPECTION 05000255/2014008

NRC-Identified and Self-Revealing Findings
Cornerstone: Initiating Events 

• Green. The inspectors identified a finding having very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR Part 50.36(c)(3), “Surveillance Requirements,” for the failure to ensure the channel time delay for the degraded-voltage monitor was included in Technical Specification (TS) Surveillance Requirement (SR) 3.3.5.2.a. Specifically, the licensee failed to include in the TS SR the required time delay after the voltage relay trips before the preferred source of power is isolated and 1E electrical loads transferred to the stand-by Emergency Diesel Generators (EDGs). This finding was entered into the licensee’s Corrective Action Program and the licensee’s preliminary verification determined the degraded voltage monitors were still operable but degraded or non-conforming. 

NRC picture...is your gonads protected.

The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to more significant safety concern. Specifically, by not incorporating the total time delay requirements into the Technical Specifications, (TS) the time could be changed without going through the TS change process, possibly leading to spurious trips of offsite power sources or possibly exceeding the accident analysis time is the FSAR. The inspectors determined the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensee’s present performance. (Section 1R21.3.b(9)) 

Cornerstone: Mitigating Systems
• Green. The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, “Design Control” for the failure to ensure the safety-related Engineered Safeguard Systems trains would not be adversely affected by air entrainment when aligned to the Safety Injection and Refueling Water (SIRW) Tank. Specifically, calculation EA-C-PAL-0877D, assumed incorrectly only one train of the Engineered Safeguards System (ESS) was in operation when evaluating if the SIRW Tank reaches the limit for critical submergence during a tank drawdown. As part of their corrective actions, the licensee re-evaluated the scenarios of concern, performed an operability evaluation, and implemented compensatory actions. 

The performance deficiency was determined to be more than minor because it impacted the Equipment Performance attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, air entrainment into the ESS systems could potentially impact the operability of the system by air binding the pumps, reduce discharge flow, discharge pressure and/or delay injection. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure system or component (SSC) but the SSC maintained its operability. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensee’s present performance. (Section 1R21.3.b(1)) 

• Green. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” for the licensee’s failure to ensure the incoming feeder cables from startup transformer 1-2 to 2400 V safety-related Buses 1C and 1D were sized in accordance with their design basis, as described in Palisades FSAR Section 8.5.2. Specifically, the licensee failed to ensure the ampacity of the cables was at least as high as their maximum steady-state current. The licensee entered this finding into their Correction Action Program and verified the operability of the cables. 

The performance deficiency was determined to be more than minor, because it impacted the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, cables were undersized with respect to the loading that would automatically occur as the result of a design basis accident. The inspectors determined the finding was of very low safety significance (Green) because the SSC maintained its operability and functionality. This finding had a crosscutting aspect in the area of Human Performance, associated with the Design Margin component, because the licensee did not ensure that equipment is operated and maintained within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. [H.6] (Section 1R21.3.b(2)) 

• Green. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” for the licensee’s failure to ensure electric motors are sized in accordance with the design basis, as discussed in Palisades FSAR Section 6.2.3.1. Specifically, the horsepower ratings of certain motors are less than power demands of their driven equipment, and they were not analyzed to ensure overheating would not occur. The licensee entered this finding into their Correction Action Program with a recommended action to analyze the effect of the condition, and has verified the operability of the motors. 

This performance deficiency was determined to be more than minor, because it impacted the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, motors serving loads with power demands in excess of the motor horsepower ratings were not analyzed to ensure that motor damage would not occur. The inspectors determined the finding was of very low safety significance (Green) because the SSC maintained its operability and functionality. This finding had a crosscutting aspect in the area of Human Performance, associated with the Design Margin component, because the licensee failed to ensure that equipment is operated within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. [H.6] (Section 1R21.3.b(3)) 

• Green. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” for the licensee’s failure to ensure that voltages on the 480V system do not exceed equipment ratings. Specifically, the licensee increased the output voltage of the supply transformers to the 480V safety-related buses by 2.5 percent, but failed to ensure the resulting voltages would not exceed equipment ratings when the system is powered from the station power transformer or emergency diesel generator. The licensee entered this finding into their Correction Action Program and verified the operability of the affected equipment. 

The performance deficiency was determined to be more than minor, because it impacted the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to verify or check the voltage increase on the 480V system to ensure the maximum allowable voltage would not exceed equipment ratings. The inspectors determined the finding was of very low safety significance (Green) because the affected SSCs maintained their operability and functionality. The inspectors did not identify a cross-cutting aspect associated with this finding, because the finding was not representative of the licensee’s present performance. (Section 1R21.3.b(4)) 

• Green. The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of Technical Specifications 5.5.7, "Inservice Testing Program," for the failure to perform comprehensive pump testing of Containment Spray Pump P-54A in accordance with the code of record. Specifically, the licensee did not rerun a comprehensive pump test, as required by the code’s ISTB-6300 “Systematic Error” section. As part of their corrective actions, the licensee entered the issue into the Corrective Action Program, and determined the component remained operable. 

The performance deficiency was determined to be more than minor because it impacted the Equipment Performance attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to perform testing as required could result in the degradation of the equipment being undetected. The finding screened as having very low safety significance because the finding was a deficiency affecting the design or qualification of a mitigating structure system or component (SSC) but the SSC maintained its operability. The findings had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee failed to thoroughly evaluate the issue to ensure that resolutions address causes and extents of conditions commensurate with their safety significance. [P.2] (Section 1R21.3.b(5)) 

• Green. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, “Test Control,” for the licensee’s failure to have adequate acceptance criteria in the emergency diesel generator surveillance procedures. Specifically, the licensee failed to ensure the surveillance test procedures for the emergency diesel generator largest load rejection test bounded the power demand of the largest load, as required by Technical Specification SR 3.8.1.5. The licensee entered this finding into their Correction Action Program and verified the operability of the emergency diesel generators.
The performance deficiency was determined to be more than minor, because it impacted the Procedure Quality attribute of the Reactor Safety, Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the surveillance procedure error could result in acceptance of test results that did not satisfy Technical Specification SR 3.8.1.5 for rejection of a load greater than or equal to the emergency diesel generator’s single largest predicted post-accident load. The inspectors determined the finding was of very low safety significance (Green) because the SSC maintained its operability and functionality. This finding had a cross-cutting aspect in the area of Human Performance, associated with the Resources component, because the licensee failed to ensure that personnel, equipment, procedures, and other resources are adequate to assure nuclear safety by maintaining long term plant safety.  

[H.1] (Section 4OA2.1.b(1))
Cornerstone: Barrier Integrity
• Green. The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” for the licensee’s failure to correctly translate design valve leakage limits into the applicable test procedure. Specifically, the acceptance criterion for emergency core cooling system (ECCS)/containment spray (CS) recirculation isolation valves CV-3027 and CV-3056 had not been correctly adjusted to account for the higher differential pressure associated with ECCS operation under post-accident conditions. The licensee entered this finding into their Corrective Action Program to correct the valve leakage limit. 

The performance deficiency was determined to be more than minor because it impacted the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that containment could protect the public from radionuclide releases caused by accidents or events. Specifically, leakage approaching the procedural values would exceed analyzed dose calculations. The finding screened as of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined this finding did not have an associated cross-cutting aspect because it was not representative of present performance. (Section 1R21.3.b(6)) 

• Green. The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion VII, “Control of Purchased Material, Equipment, and Services,” for the licensee’s failure to identify non-safety-related sub-components improperly supplied with safety-related valves. Specifically, ECCS/CS recirculation isolation valves CV-3027 and CV-3056, which were installed in 2007, were supplied with non-safety-related sub-components. These components were identified as non-safety-related on the vendor drawings. In addition, the licensee later installed a section of non-safety-related tubing on valve CV-3027 based on the incorrect vendor drawing. The licensee entered this finding into their Corrective Action Program to correct the valve drawings and replace the non-safety-related parts. 

The performance deficiency was determined to be more than minor because it impacted the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that containment could protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to identify non-safety-related sub-components improperly supplied with safety-related valves which would form part of the containment barrier under post-accident conditions. The finding screened as of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined this finding did not have an associated cross-cutting aspect because it was not representative of the licensee’s present performance. (Section 1R21.3.b(7)) 

• Green. The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, “Test Control,” for the licensee’s failure to establish an adequate test program for the Shutdown Cooling (SDC) Heat Exchangers (HXs) to demonstrate they can perform as designed. Specifically, the licensee failed to take actions to ensure the SDC HXs’ heat transfer capability met its design bases, as assumed in design bases calculations. 

The performance deficiency was determined to be more than minor because it impacted the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that containment could protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to verify the SDC HXs heat transfer capability met their design bases, as assumed in design bases calculations, to limit containment temperatures and pressures during an event. The finding screened as of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined this finding had an associated cross-cutting aspect, Conservative Bias, in the Human Performance cross-cutting area. Specifically, on several occasions when the licensee identified the need to perform testing and/or inspection of the SDC HXs, the licensee did not take actions because they did not believe any regulatory requirements or technical issues existed that required the testing and/or inspections. [H.14] (Section 1R21.3.b(8)) 

B. Licensee-Identified Violations
Violations of very low safety or security significance or Severity Level IV that were identified by the licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensee’s Corrective Action Program (CAP). These violations and CAP tracking numbers are listed in Section 4OA7 of this report.

NRC Checking if Gonads protected With Titanium Vest On Employees at Palisades (offical NRC pictures).

Entergy-Palisades replaced leaking CRMD housings more often than you change the oil in your car. To do this job, they take off the reactor vessel head and place it aside on the refueling floor. It is pretty radioactivity hot job and I think they are in respirators. They had lots of preventable fuel failures in the recent past and that makes this job particularly radioactivity hot.

Hmm, does these employees feel so insecure about the high radiation levels, the "not required vest" acts as a toddler blanket comforter?

Is it for beta or alpha protection?  

The below translated, Entergy said these titanium vest gave them more protection than they actually provided...meaning Palisades put down on the paperwork for these employees a radiation dose lower than what they actually got. They collectively got 98 rems more dose than what they first put down on paperwork. I am wondering of the HP contractor talked Entergy into this artificial titanium shield reduction in dose...hey, you can report your dose to the NRC lower than it actually is. How much of this titanium dose crap is going on throughout the industry. Palisades and Entergy is supposed to have highly educated and experience professional people on their staff such that this never could occur.

Well, it is a lot more complicated than I stated it. It is basically they got a tremendous radiation gradient, if they put  radiation detectors in the upper chest and their ass is sitting on a hot spot without a detector there…the dose detected is not accurate. Their fingers might be taking off a nut in a hot zone, with the rest of the body not being in that field. They basically got computer programs that can take these peripheral reading and come up with a whole dose.  

That reactor head and its control rod drive mechanism was defective and obsolete since the first day that reactor came into operation. Palisades should have designed, manufactured and installed a new head and CDRM package decades ago. But they been always been crying poor boy and one step away from permanent shutdown.

See, you can’t see this in its proper context. I bitterly complained about the dangerous and poorly designed primary coolant pumps during this very same outage. The NRC has been letting them get away with operating not according to plant designs and licensing for decades…thus causing impeller blades to fly out of the housing and get flung all around the coolant and stuck in areas in the reactor.

 It is like the tire nuts on your right front wheel being loose and you feel front end shaking as you drive the car. The NRC says this it is safe to drive until you get to the next oil change. If the NRC allows 100 nuclear plants to operate with that kind of attitude, then our nation is in deep trouble when eventually one plant pops its cork. 

Basically this is a slap on the wrist of Entergy having no meaning to interrupt the work on the CRDM job. This happened around early this year in Feb/Mar...why is this inspection report especially so late? Typically the NRC makes these reports like this, it is harmless to the big boy players… but it acts as protection against somebody calling it as a cover-up. This is how you do cover ups today…disclose it as a shield against somebody calling it coverup in in the future. It is corruption and cover-up protection. It is pretty smart really.

These guys are on artificial respiration and one breath away from a permanent plant shutdown…the NRC now has a policy to handle these guys with kit gloves..

December 2, 2014

EA-14-168

SUBJECT: PALISADES NUCLEAR PLANT, NRC INSPECTION REPORT 05000255/2014010, PRELIMINARY WHITE FINDING

Preliminary White. The NRC identified a finding and two apparent violations of NRC requirements associated with the replacement of CRD housings between February 6 and March 8, 2014. Specifically, the inspectors identified an apparent violation of Title 10 of the Code of Federal Regulations (CFR) Part 20.1201, "Occupational Dose Limits for Adults," because the licensee failed to ensure that radiation worker dosimeters calibrated to the Deep Dose Equivalent (DDE) were located at the highest exposed portion of the respective compartment, a condition of the NRC-approved method for determining effective dose equivalent external (EDEX). The inspectors also identified an example of an apparent violation of Technical Specification 5.4 "Procedures," associated with this finding. Upon identification of this issue, the licensee suspended the use of EDEX and tungsten shield vests. The licensee re-calculated the dose received for the workers involved and updated the nuclear power industry’s dose tracking system with the revised dose results. Additionally, a root cause evaluation was initiated under Condition Report CR-PLP-2014-04683.

Definitely gonads not protected.

What’s the NRC’s proof that body got gonads? 

Wall Street Scam: Relicensing Nuclear Plants for 80 Years and Nuclear Grade Titanium Gonad Protection Shields.

Wall Street Scam: Relicensing nuclear plants for 80 years and titanium gonad protection shields.

I think the aims of this is to artificially increasing the value of the utility company and its stock price. They get a free couple of pennies worth of corporate valuation over this lie.

A lot of these plants are destroying the long term value of these plants by tolerating gross fuel failures aiming to buttress short term profits over the long term viability of the plant.   

Bet you before the 50 year old mark most of these nuclear utilities will find it at the uneconomical tipping point to keep up with the maintenance of these nuclear plants. The primary core and the coolant system radiation and contamination levels constantly build up in the life of these plants.  With the high radiation levels, these guys will find it impractical (without harming their employees) to service and perform maintenance in their primary system. They  will waste a lot of money trying to protect their employees against radiation instead of maintaining and increasing safety through upgrades and maintenance.

The Palisades CRDM housing replacement job is a prime example. Did you ever see the “Silver Lining Playbook” movie. The Entergy employees at Palisades working on the CRDM housing job wore greater than 30 pound titanium chest vest in order to protect themselves from radiation. Do you remember Pat Solatano Jr wearing a black plastic trash bag chest vest? The must also have been wearing 50 pounds lead diapers to protect their gonads.

   

Nope,  the NRC dinged Palisades on not protecting their employee's gonads. How would you like that job, an NRC radiation titanium shield gonad  protector inspector? 

NRC inspector: "Furthermore, some photos depicted workers sitting in the work area in an orientation that negates the protective qualities of the tungsten shield vest for the gonads while the dosimeter for those organs remained shielded by the tungsten shield vest."

I think this is a male sitting on a radiative hot nuclear reactor head with a titanium gonad shield...this is how not to protect you gonads!!!

Gonads definitely not protected! 

Titanium vest shield and gonads not protected(one with black vest on)

60.

Although no applications for an SLR have been submitted, several utilities are evaluating whether to apply for one, including Dominion Resources for its Surry Power Station Units 1 and 2 in Virginia (current license expiration dates of 2032 and 2033) as well as Exelon for its Peach Bottom Atomic Power Station Units 2 and 3 in Pennsylvania (current license 

Nearly all U.S. nuclear power plants need extension past 60 years to operate beyond 2050 

December 8, 2014

When nuclear power plants are built, the Nuclear Regulatory Commission (NRC) has the authority to issue initial operating licenses for a period of 40 years. Beyond that, the reactors need license renewals, and the NRC has granted 20-year license renewals to 74 of the 100 operating reactors in the U.S., according to the Energy Information Administration.

These reactors may now operate for a total period of 60 years. They represent a cumulative capacity of a little more than 69,000 MW. The NRC is currently reviewing license renewal applications for an additional 17 nuclear reactors, and expects to receive seven more applications in the next few years. 

With the bulk of the existing nuclear fleet licensed before 1990, nearly all existing reactors will be more than 60 years old by 2050. Unless a utility applies for and receives a subsequent license renewal that could further extend the operating lives of their reactors up to 20 additional years, the reactors will not generate power beyond age 60...

Entergy-River Bend fined $70,000 for Improper Security Department Decorium.

I had a high NRC OIG official asked me if I know about any more security department events at the Palisades plant. Now we know why she asked me that. What a terribly troubled company. How many of these Confirmatory Orders with the repeated basic aims of training the plant staff and flee, all surrounding ethics and NRC requirements have we seen in recent years at Entergy-nuclear? There has been a tremendous amount of similar Confirmatory Orders and the focus always is telling the truth, following NRC regulations and ethics fleet wide. It is call crazy doing the same wrong thing over and over again without change.

I bet you this is about offensive materials in the security watch stations or offices. You notice Entergy's management and the  NRC can't figure out who are the culprits with violating the decorum not stated policies.  I thinking it is either porn, racial intimidation and it even might be right wing anti government propaganda. Is these even against any NRC rule.

You get it, it is national security terrorism secrecy information where they can't publicly describe what this $70,000 decorum violation is all about?  

Basically we are talking about two systems of accountability...a NRC secret system of withholding information that obviously doesn't work and the transparently shaming a bad utility actor with openly disclosing all security violations. If you can’t control these guys through your regulations, we can control them through public shaming and shutting down nuclear plants.

But mike, that would give information to terrorist. The vulnerabilities are giving terrorist information or allowing these utilities in a secret system of bloodless paper cuts of no NRC accountability ...not requiring a change in behavior that allows the security staff to cycle down into hidden dysfunction. We have seen this as example at Vermont Yankee. I think we would all be a lot more safer if we'd seen all the security violations and we effectively shamed a corporation publicly with their bad behaviors. Then the public and community could drive the NRC and utility into behavior change with a bad actor utility.

As this system of security secrets stands now, is these secret protect the industry and the NRC from accountability. The system allows these licensees to spend less money on security. It is a tragic abuse of using national security as a tool to protect these corporations...it is in our highest societal interest to maintain faith in a centralized organizing force or government in time of a crisis. If these security staffs at these nuclear plants are more dysfunctional than disclosed through the paper cut punishment NRC bad behavior change...then we will lose faith in government if the security staff collapses in the face of insignificant terrorist like pressure.

In other words, shaming these utilities with the truth on violations would create a system that bolsters and increases the public safety more than the vulnerabilities of giving terrorist insignificant advantages with insider security information prior to an attack. The sin would be if the NRC gave River Bend the old wink wink and documentation paper cut punishment of no bother...their security department cycles down unseen into dysfunction more that the community can see and correct. Then a terrorist attack occurs and the security department isn't really prepared for it.
The last time I put terrorism and nuclear plant in the same sentence, I had two FBI Joint Terrorism Task Force agents at my door asking questions and threatening to out me into jail. 

Democracy in not a system for cowards!!!    

CONFIRMATORY ORDER, NOTICE OF VIOLATION, AND CIVIL PENALTY- NRC SPECIAL INSPECTION REPORT 05000458/2014407 AND NRC INVESTIGATION REPORT 4-2012-022- RIVER BEND STATION

On September 22, 2014, the NRC and Entergy met in an ADR session mediated by a
professional mediator, arranged through the Cornell University Scheinman Institute on Conflict
Resolution. ADR is a process in which a neutral mediator, with no decision-making authority,

OFFICIAL USE ONLY SECURITY RELATED INFORMATION
OFFICIAL USE ONLY - SECURITY RELATED INFORMATION

assists the parties in reaching an agreement on resolving any differences regarding the dispute.
During the ADR session, a preliminary settlement agreement was reached. The elements of
that preliminary agreement, with the exception of the section that includes SRI, are described
below. The portions of the agreement that contain SRI, as well as the sections of this
Confirmatory Order that address SRI, are described in the aforementioned non-public
Attachment. The following description of the preliminary ADR agreement, and the required
actions described in Section V of this Confirmatory Order, include references to the non-public
Attachment to allow for public release of this Confirmatory Order. The publicly available
elements of the agreement consist of the following:

The NRC recognizes the corrective actions that Entergy has already implemented associated
with the apparent violation and preliminary finding. Entergy's corrective actions are described in
the non-public Attachment.

A. The NRC and Entergy agree that a willful violation of Title 1 0 Code of Federal
Regulations (1 0 CFR) Part 73 occurred on March 18, 2012, at River Bend Station.
However, the NRC and Entergy disagree on the specific aspects of that willful
characterization of the violation. The details regarding these aspects are described in
the non-public Attachment.

1. The NRC concluded that the security-related violation occurred because of the
deliberate misconduct of an unidentified security officer at River Bend Station.
2. Entergy does not believe that willful intent was involved in all aspects of the violation.
3
OFFICIAL USE ONLY SECURITY RELATED INFORMATION
OFFICIAL USE ONLY SECURITY RELATED INFORMATION

B. Within 4 months from the date of this Confirmatory Order, Entergy will revise its security
procedures.

C. Within 3 months from the date of this Confirmatory Order, Entergy will, at each of its
nuclear plants, conduct a review of its controls for SRI and communicate to the NRC the
results of the review. Within 6 months from the date of this Confirmatory Order, Entergy
will establish new controls and will provide its proposed controls to the NRC for its
review. The NRC will communicate to Entergy any concerns regarding the controls
within 60 days of submittal for resolution in a manner acceptable to both parties.
Entergy will implement the controls within 15 months from the date of this Confirmatory
Order. The details regarding these controls are described in the non-public Attachment.
D. Within 9 months from the date of this Confirmatory Order, Entergy will review and
evaluate the location and storage of SRI at each of its nuclear plants. The details are
described in the non-public Attachment.

E. Entergy will develop a "commitment to compliance" statement or a similar document
highlighting the special responsibilities of nuclear security personnel. This document will
explain that nuclear security personnel need to comply with regulations and procedures,
and it will describe the potential consequences if compliance does not occur. Within
12 months from the date of this Confirmatory Order, Entergy will require at each of its
nuclear plants that nuclear security personnel read and sign the statement (subject to
any collective bargaining obligations it may have). Entergy will include the reading and
signing of this statement in the initial qualification process of nuclear security personnel.
The details are described in the non-public Attachment.
4
OFFICIAL USE ONLY SECURITY RELATED INFORMATION
OFFICIAL USE! ONLY SEiCURITY REiLATEiD INFORMATION

F. Within 6 months from the date of this Confirmatory Order, Entergy will identify those
security posts in each of its nuclear plants that should be subject to certain decorum
standards that will ensure a professional environment in those areas. Once identified,
Entergy will establish decorum protocols for those security posts. In addition, within
6 months of the date of this Confirmatory Order, Entergy will provide its proposed
decorum protocols to the NRC for its review. The NRC will communicate to Entergy any
concerns regarding the proposed decorum protocols within 60 days of submittal for
resolution in a manner acceptable to both parties. Entergy will implement the decorum
protocols within 12 months from the date of this Confirmatory Order.

G. Within 4 months from the date of this Confirmatory Order, Entergy will prepare a
"lessons learned" presentation to be delivered to Entergy nuclear employees at each of
its nuclear plants describing the event that formed the basis for this violation. Prior to
making the presentation, Entergy will provide its proposed presentation to the NRC for
its review. The NRC will communicate to Entergy any concerns regarding the
presentation within 30 days of submittal. Entergy will deliver the presentation to Entergy
nuclear employees within 12 months of this Confirmatory Order.

H. Within 4 months from the date of this Confirmatory Order, Entergy will prepare a
presentation describing the event that formed the basis for this violation. The
presentation will be delivered to the Nuclear Security Working Group and the National
Nuclear Security Conference (subject to acceptance of the conference-organizing
committees). This presentation will include, among other subjects, the subjects covered
in the non-public Attachment to this Confirmatory Order. Prior to making the
presentation, Entergy will provide its proposed presentation to the NRC for its review.

OFFICIAL USE! ONLY SEiCURITY REiLATED INFORMATION
OFFICIAL USE ONLY SECURITY RELATED INFORMATION

The NRC will communicate to Entergy any concerns regarding the presentation within
30 days of submittal. Entergy will deliver the presentation within 12 months of this
Confirmatory Order.

I. Within 6 months from the date of this Confirmatory Order, Entergy will ensure that an
independent third party conducts a safety culture assessment of the Security
organization at River Bend Station. The results will be incorporated into Entergy's
corrective action program as appropriate. A copy of the completed assessment will be
made available for NRC review.

J. Within 4 months from the date of this Confirmatory Order, Entergy will prepare refresher
training on the provisions of 10 CFR 50.5 and 50.9 for Entergy employees at each of its
nuclear plants. Prior to conducting the training, Entergy will provide its proposed
refresher training plan to the NRC for its review. The NRC will communicate to Entergy
any concerns regarding the plan within 30 days of submittal for resolution in a manner
acceptable to both parties. Entergy will complete administration of this refresher training
within 12 months of

( North Anna) Dominion's Intentionally Running Their Nuclear Reactor to Failure Philosophy

Dec 14

WAM-E.1   
08:30  Tungsten Shield Vest Barbara Thompson*, Dominion - North 
Anna Power Station 


Abstract: This paper 
will discuss the evaluation and implementation of the tungsten shield vest as an 
ALARA measure to reduce personnel exposure. The tungsten shield vest was 
invented in 2009 by an industry veteran from Entergy Arkansas Nuclear One. The 
form-fitting shield vests are 
similar to lead vests worn to protect dental and medical patients from 
radiation exposure – but much lighter. Tungsten shielding reduces the amount of 
radiation to the parts of the body that are most sensitive to ionizing radiation and 
is flexible enough to move with the worker. In 2010, the North Anna Power 
Station became the first nuclear facility in the Dominion Fleet to test and use 
tungsten shielding vests. Due to the non-uniform radiation field created by the 
vest, the Radiological Protection group utilized Effective Dose Equivalent (EDE) 
monitoring of individual whole body compartments. The tungsten shield vests have 
been used for fuel manipulator roller replacement, pressurizer spray valve 
repairs, and transfer canal upender sheave replacement. Initial test results 
showed up to a 30 percent reduction in personal radiation exposure for workers 
doing outage maintenance tasks within the containment building at North Anna. 
The use of tungsten vests has allowed the benefits of shielding to become mobile 
with the worker and their tasks, thus reducing the overall radiation exposure to 
the worker. 

Fuel Reliability: How it affects the industry, and one fuel vendor's journey to flawless fuel performance
http://www.power-eng.com/articles/npi/print/volume-7/issue-4/nucleus/fuel-reliability-how-it-affects-the-industry-and-one-fuel-vendor-s-journey-to-flawless-fuel-performance.html

https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML14338A739

Trait Talk: A questioning Attitude

For example, ten years after 
removal from a reactor, the surface dose rate for a typical spent fuel 
assembly exceeds 10,000 rem/hour, whereas a fatal whole-body dose for humans is 
about 500 rem (if received all at one time). Furthermore, if constituents of 
these high-level wastes were to get into ground water or rivers, they could 
enter into food chains.

http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/1993/in93082.html

Information Notice No. 93-82: Recent Fuel and Core Performance problems in Operating Reactor

These fuel failures have been attributed to high cross flows,
caused, in part, from mixed fuel designs which induced fuel rod vibration with
fretting wear at the lower grids.  The mixture of standard and VANTAGE 5H fuel
(with debris filter bottom nozzles) resulted in axial mismatches between the
bottom nozzles and the grid spacers of the two fuel types.

Although the staff recognizes that it is impossible
to avoid all fuel rod failures and that cleanup systems are installed to handle a
small number of leaking rods, the review must ensure that fuel does not fail as a
result of specific causes during normal operation and AOOs.

The allowable fretting wear should be stated in
the safety analysis report, and the stress and fatigue limits in items (i)
and (ii) above should presume the existence of this wear.

2.3 Fuel Dispersal
Fuel dispersal is the ejection of fuel fragments or particles through a rupture or opening in the
cladding. For the purpose of this report, fuel dispersal is said to have occurred if any fuel
material is found outside of the fuel rod. Even if the fuel material is small in quantity, the finding
will be noted and qualified by the nature of the dispersal (e.g., “only a small black powder on the
test chamber wall was observed”).


3.14. Fuel fragmentation and fuel dispersal
In this chapter, fuel fragmentation refers to situations for which the fuel
cladding breaks into pieces and fuel dispersal to situations for which fuel particles
escape from the cladding following a rupture.
Although fuel fragmentation is traditionally considered to exist only in
conjunction with highly energetic events such as the reactivity-initiated accidents
(RIA), recent results from the Halden test reactor show that fuel fragmentation can
also occur during the loss-of-coolant accident (LOCA).


1981: US. NUCLEAR REGULATORY COMMISSION
STANDARD REVIEW PLAN
OFFICE OF NUCLEAR REACTOR REGULATION

4.2


"Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has,
therefore, been breached.

Fuel rod failure is defined as the
loss of fuel rod hermeticity.


007

In a “fuel rod failure,” the fuel rod leaks and the first fission product barrier (the cladding) is breached.

Definition

1. (General Physics) sealed so as to be airtight

2. hidden or protected from the outside world

Fuel Vendor
PWR
Fuel Failure Management
Handbook
The tendency of grid-rod fretting
increases with increased cross flow, e.g. due to different pressure drops of
different fuel assembly designs sitting adjacent to each other. Baffle jetting will
also increase the risk of getting grid-rod fretting. To reduce the risk of getting
grid-rod fretting, appropriate loop tests should be performed to verify the
fretting performance of a new grid design.

The ultimate solution is the conversion of down-flow design to up-flow
design.
These above mentioned corrective actions have successfully reduced the
number of plants and fuel assemblies that are affected by this failure
mechanism. Very few fuel failure as the result of baffle jetting has been
observed in recent years.


Brief Description: This 10 CFR 50.59 evaluation is being performed to support the
transition to RFA-2 fuel at North Anna.

Reason for Change: ETE-NAF-2011-0173 implements the analyses supporting the
RFA-2 fuel transition at North Anna. UCR-2010-007 is a UFSAR change request that
encompasses the changes related to the RFA-2 fuel transition at North Anna. Three
License Amendment Requests (LARs) were submitted to the NRC for approval in
support of the RFA-2 transition at North Anna. The LAR for Optimized ZIRLO has been
approved and the LARs for VIPRE-D and Westinghouse's Best Estimate Large Break
LOCA evaluation were being tracked and were subsequently approved by the NRC.

http://pbadupws.nrc.gov/docs/ML0631/ML063110095.pdf

BAFFLE JETTING RESULTS IN FUEL ROD DEGRADATION AT VIRGIL C. SUMMER

Examination of two previous cycle fuel assemblies from the same location showed indications of baffle jetting (shiny areas) on the same fuel rod location as P1.



https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML14316A338

November 10, 2014
Mr. David A. Heacock
President and Chief Nuclear Officer
Virginia Electric and Power Company
Innsbrook Technical Center
5000 Dominion Boulevard
Glen Allen, VA 23060
SUBJECT: NORTH ANNA POWER STATION – NRC INTEGRATED INSPECTION
REPORT 05000338/2014004 and 05000339/2014004

The inspectors reviewed the Outage Safety Review (OSR) and contingency plans for the
Unit 2 refueling outage, which began September 7, 2014, to confirm that the licensee
had appropriately considered risk, industry experience, and previous site-specific
problems in developing and implementing a plan that assured maintenance of defense in-
depth.




1998:
North Anna Unit 1 Fuel Failures
During the present refueling outage, it was determined that there were 19 failed (leaking) fuel rods in 9 assemblies in the completed cycle at North Anna
Unit 1. Coolant activity had indicated fuel failures, but the number was unknown until the outage. These failures were all in third-cycle, ZIRLO clad fuel,
located near the baffle. The root cause investigation is underway, with the most likely cause being grid to rod fretting at the mid grids. These failures
appear to be similar to vibration related failures that were experienced in the 1993 time frame at two other plants. The North Anna fuel had incorporated
the rotated grids which was the fix for the previous problem. All affected assemblies were scheduled for discharge during this outage. Virginia Power has
installed vibration suppression devices on all peripheral locations for the upcoming cycle. The Reactor Systems Branch (SRXB) will continue to follow this
issue.


https://www.nukeworker.com/forum/index.php?action=printpage%3Btopic=38257.0

Title: Re: North Anna Fuel Failure
Post by: cheme09 on Sep 18, 2014, 05:39

---

Has been Areva, but we're in the process of going to W fuel. I think the last Areva bundles come out next outage.

https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML060200275
G45 Fuel Assembly Clad Damage

The G45 assembly has been ultrasonically and visually inspected and the failure is in a high power IFBA rod and is believed to have occurred during the initial power ascension for Cycle 6. This is evident by the secondary hydriding of the failed rod and the high Iodine concentrations experienced in the Reactor Coolant System (RCS) during R Cycle 6.

The existence of a cladding leak was initially established during Cycle 6 operation through sampling of the Reactor Coolant System (RCS) that identified elevated levels of Iodine 131 (1-131) and Xenon 133 (Xe-133).

This condition was documented in TVA's corrective action program as Problem Evaluation Report (PER) 9174. A limit for the concentration of 1-131 is defined in Limiting Condition for Operation (LCO) 3.4.16, "Reactor Coolant System (RCS) Specific Activity." In order to ensure this limit was closely monitored during Cycle 6, the RCS was sampled three times a week and reviewed by site management.

At the time a fuel leak was initially identified in October 2003, Operations personnel notified appropriate site
management of the problem and ensured the problem was documented in TVA's corrective action program.
When the cladding defect in rod G45 was identified in November 2005, the Operations staff ensured the
required notifications were made to NRC in accordance with 10 CFR 50.72.

http://www.orau.org/ptp/PTP%20Library/library/NRC/Info/in87039.pdf

August 21, 1987
NRC INFORMATION NOTICE NO. 87-39: CONTROL OF HOT PARTICLE CONTAMINATION
AT NUCLEAR POWER PLANTS
In a recent study for the NRC (Reference 3), it was reported that a plant
operating with 0.125 percent pin-hole fuel cladding defects showed a
general five-fold increase in whole-body radiation exposure rates in some
IN 87-39
August 21, 1987
Page 4 of 5
areas of the plant when compared to a sister plant with high-integrity
fuel (<0 .01="" around="" certain="" degraded="" leakers="" p="" percent="" plant="" systems="" the="">fuel may elevate radiation exposure rates even more.


http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/1982/in82027.html

Information Notice No. 82-27: FUEL ROD DEGRADATION RESULTING FROM BAFFLE WATER-JET IMPINGEMENT

Although the baffle water-jetting problem has been experienced in a limited 
number of Westinghouse PWRs, this information notice is being distributed to
all licensees and construction permit holders, including PWRs whose core 
baffle designs may have features which contribute to fuel rod failures as 
previously described.  Such fuel degradation may result in relatively high 
primary coolant activity and thereby impede periodic maintenance-related 
functions and/or pose radiological hazards to personnel. 

2010: CR319099, “North Anna 2 confirmed single fuel rod failure” (subsequently
determined to result from unidentified debris)

Just in Time Training Presentation, Response to Recently Identified Fuel Failures, dated 09/22/2014

09/06/2014

REFUELING OUTAGE

LER 2014-002-00: Failed Fuel Assembly  

Nov 12, 2014

05000338/340

On September 15, 2014, with Unit 2 defueled, debris that had the potential to be fuel fragments was located on the core plate directly below the B131 core location, where fuel assembly 4Z9 resided during Cycle 23. Video inspection of fuel assembly 4Z9 identified that the top springs of two fuel pins were dislodged.

Due to the fact that the fuel damage exceeded expected conditions, at 1454 on September 15, 2014, this event was reported as an eight hour report as per 10 CFR 50.72(b)(3)(ii)(A), any event or condition that results in the condition of the nuclear plant, including its principle safety barriers, being seriously degraded. Detailed video inspections estimated that 15 fuel pellets were dislodged from fuel assembly 4Z9. During efforts to identify and recover the fuel pellets, 7 fuel pellets worth of material were not found and have already or are expected to granulate into fine particles that will dissolve in low flow areas of the primary plant systems, or be removed by normal purification processes. Since the specific location of the 7 fuel pellets is undesignated and because those pellets contain licensed material in a quantity greater than 10 times the quantity specified in App. C of 10 CFR 20, a report was made at 1227 on September 30, 2014, pursuant to 10 CFR 74.11(a) and to 10 CFR 20.2201 (a)(ii). The health and safety of the public were not affected by this event.

At 0900 on September 15, 2014, with Unit 2 defueled, debris that had the potential to be fuel fragments was located on the core plate (EIIS System - AC) directly below the B1 1 core location. Ten pieces of material, approximately 1/8" in diameter, were found. The material was near the edge of the outer flow hole and partially under the gap between the baffle plate and the core plate. Fuel assembly (ElIS System -AC) 4Z9 was located at the B131 location during Cycle 23. Video inspection of fuel assembly 4Z9 identified that the top springs of two fuel pins were dislodged.

Due to the fact that the fuel damage exceeded expected conditions, at 1454 on September 15, 2014, this event was reported as an eight hour report as per 10 CFR 50.72(b)(3)(ii)(A), any event or condition that results in the condition of the nuclear plant, including its principle safety barriers, being seriously degraded.

Detailed video inspections estimated that fifteen (15) fuel pellets were dislodged from fuel assembly 4Z9. For reference, the reactor core contains approximately 15 million fuel pellets. During efforts to identify and recover the fuel pellets, debris fragments estimated to represent five (5) fuel pellets were found in the damaged fuel assembly that is currently in the Spent Fuel Pool (SFP) (EIIS System - DA). In addition, an estimated three (3) pellets worth of material was retrieved by the foreign object search and retrieval (FOSAR) efforts in the reactor vessel and are now located in the SFP. The remaining seven (7) fuel pellets have already or are expected to granulate into fine particles that will dissolve in low flow areas of the primary plant systems or be removed by normal purification processes. However, since the specific location of the seven (7) fuel pellets is undesignated, a report was made at 1227 on September 30, 2014, pursuant to 10 CFR 74.11 (a) for the loss of special nuclear material (SNM). At that same time, a report was made pursuant to 10 CFR 20.2201 (a)(ii) because the seven (7) fuel pellets contain licensed material in a quantity greater than 10 times the quantity specified in Appendix C of 10 CFR 20. 10 CFR 20.2201(b) requires a written report after the initial notification for the occurrence of any lost, stolen, or missing licensed material that was reported under 10 CFR 20.2201 (a)(ii) for licensed material in a quantity greater than 10 times the quantity specified in Appendix C of 10 CFR 20. The following topics are required to be addressed:

(i) A description of the licensed material involved, including kind, quantity, and chemical and physical form:

Fuel Pellet Description - Based on the review of the video of the recovered material, the possibility that these fuel pellets have remained intact is very low.

Type of Special Nuclear Material Uranium dioxide pellets initially enriched to 4.45% 

Length of fuel pellet 0.4 inches nominal Pellet diameter 0.3225 inches 

Total Uranium in the 7 fuel pellets 32.3 grams (Sept 2014) 

Total Uranium 235 in the 7 fuel pellets 0.4 grams (Sept 2014) 

Total Plutonium in the 7 fuel pellets 0.4 grams (Sept 2014) 

Total Fissile Plutonium in the 7 fuel pellets 0.3 grams (Sept 2014) 

Activity Level 266 Ci 

Average Burnup of Assembly 4Z9 46733 MWD/MTUEffective Full Power Days (EFPD) of Assembly 4Z9 1160 EFPD

(ii) A description of the circumstances under which the loss or theft occurred:

The fuel pellet loss occurred as a result of baffle jetting on the fuel assembly. The affected fuel rods had their top springs dislodged and fuel pellets were able to escape the fuel rod. Fragments of fuel pellets were found within the associated fuel assembly and on the core plate. However, about seven (7) fuel pellets worth of material were not located and have already or are expected to granulate into fine particles that will remain in low flow areas of the primary plant systems or be removed by normal purification processes. The possibility of theft is not plausible because of the plant's radiation monitoring instrumentation, physical security measures, and the size and type of container required for transporting nuclear material of this nature.

(iii) A statement of disposition, or probable disposition, of the licensed material involved:

During efforts to identify and recover the fuel pellets, debris fragments estimated to represent five (5) fuel pellets were found in the damaged fuel assembly that is currently in the SFP. In addition, an estimated three (3) pellets worth of material was retrieved by the FOSAR efforts in the reactor vessel and are now located in the SFP. The remaining seven (7) fuel pellets have already or are expected to granulate into fine particles that will dissolve in low flow areas of the primary plant systems or be removed by normal purification processes.

(iv) Exposures of individuals to radiation, circumstances under which the exposures occurred, and the possible total effective dose equivalent to persons in unrestricted areas:

No unauthorized exposure to radiation occurred to the plant staff or members of the public because the fuel pellet fragments either remain in the SFP or granulated into fine particles that will dissolve in low flow areas of the primary plant systems or be removed by normal purification processes.

(v) Actions that have been taken, or will be taken, to recover the material: During efforts to identify and recover the fuel pellets, debris fragments estimated to represent five (5) fuel pellets were found in the damaged fuel assembly that is currently in the SFP. In addition, an estimated three (3) pellets worth of material was retrieved by the FOSAR efforts in the reactor vessel and are now located in the SFP. The remaining seven (7) fuel pellets have already or are expected to granulate into fine particles that will dissolve in low flow areas of the primary plant systems, or be removed by normal purification processes.

(vi) Procedures or measures that have been, or will be, adopted to ensure against a recurrence of the loss or theft of licensed material:

Westinghouse fabricated and delivered a low-enrichment RFA-2 fuel assembly armored with seven (7) stainless steel rods in place of fuel rods which could be affected by jets from baffle gaps for core location B1 1 in cycle 24. A similar modification to that of the reactor vessel upflow conversion design change that was performed on Unit 1, DC NA-95-001, will be developed and implemented on Unit 2.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS

No significant safety consequences resulted from this event because the reactor coolant system activity levels during Unit 2 Cycle 23 were well within the requirements of Technical Specification (TS) 3.4.16, Reactor Coolant System Specific Activity. After Cycle 24 startup, the activity remains well within the requirements of TS 3.4.16. The health and safety of the public were not affected by this event.

3.0 CAUSE

The direct cause of the event was due to baffle jetting. Baffle jetting is the process by which water on the outside of the core baffle plate is forced through small openings in the baffle seams and onto the fuel assemblies. During Unit 2 Cycle 23, baffle jetting caused two rods in assembly 4Z9, located in core position B11, to begin rotating and vibrating. This movement resulted in fuel rod wear and eventual mechanical failure and rod separation. Once separated, a maximum of 15 fuel pellets were released from the two affected fuel rods. The Root Cause of the failed fuel assembly was the change in material properties of the baffle plates and bolting due to aging mechanisms resulting in the gap widening at the baffle joint. Stress, temperature, and irradiation since initial plant start-up have resulted in relaxation, creep, and loss of pre-load in the bolting and baffle plates. The changes in material properties allowed the gap in the corner baffle joint, adjacent to location B1 1, to widen when subjected to the relatively high differential pressure, approximately 25 psi, associated with the baffle-barrel downflow configuration in North Anna Unit 2.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

Westinghouse fabricated and delivered a low-enrichment RFA-2 fuel assembly armored with seven (7) stainless steel rods in place of fuel rods which could be affected by jets from baffle gaps for core location B1 1 in cycle 24. Visible debris on the core plate from 4Z9 was found and retrieved. An inspection of the baffle was performed with no anomalies noted. An inspection of the fuel assembly that was previously located at B1i1 for fuel cycle 22 was performed and no indications of baffle jetting were noted. An inspection was performed of other cycle 23 fuel assemblies in other baffle locations for baffle jetting damage. Ten of the other assemblies exhibited some indications at the center injection locations ranging from slight marks on a mid-grid adjacent to rod 15 or rod 3 to some slight surface erosion or buffing of the grid at the same locations. Other than being in the proper location for where center injection would be expected to occur, it was not clear whether the indications were due to baffle jetting or to some other interaction such as fuel handling or wear from a bowed assembly rubbing against the baffle plates. The indications were reviewed by Dominion's Nuclear Analysis and Fuel (NA&F) Fuel Performance Analysis (FPA) group, and it was determined that no further action was required. Both AREVA and Westinghouse reviewed the video of 4Z9 and concluded that the cause was baffle jetting. A revised Reload Safety Evaluation (RSE) incorporating the replacement fuel assembly for location B1 1 was completed and approved. An Operability Determination, OD000600, was completed for baffle jetting.

5.0 ADDITIONAL CORRECTIVE ACTIONS

No additional corrective actions were identified by the Root Cause Team.

6.0 ACTIONS TO PREVENT RECURRENCE

A modification similar to the reactor vessel upflow conversion design change that was performed on Unit 1, DC NA-95-001, will be developed and implemented on Unit 2.

7.0 SIMILAR EVENTS

Unit 2 has operated without indications of baffle jetting for 34 years and Unit 1 has operated without baffle jetting since 1996 when the upflow conversion was performed. While Unit 1 did have baffle jetting issues prior to 1996, the baffle jetting issues were from the center joints. Whereas the Unit 2 baffle jetting was from a corner joint. Additionally, Unit 2 has a different bolting configuration that made it less susceptible to the baffle jetting experienced on Unit 1.

8.0 ADDITIONAL INFORMATION

Unit 1 continued operating in Mode 1, 100 percent power during this event.

Skyrocketing Wholesale Electricity Price this Winter (not)

Dec 3:

The high price of ISO electricity today seems to have made me eat my prior words...

because:

1) The grid crooks and speculators know the authorities and politicians are watching them. 

2) Natural Gas pipelines planning is ongoing.

3) The crooks and speculator would like to see weakening prices so as to disrupt the new pipelines.

4) The benchmark energy source petroleum is in steep decline. 

ISO-New England's Winter Outlook Again Shows Need ForMore NatGas Pipelines

David Bradley 

November 21, 2014

New England should have sufficient resources in place this winter to meet consumer demand for electricity, but "insufficient pipeline capacity to meet power generators' demand for natural gas continues to be a particular concern during the winter months," according to the region's grid operator. 

"New England's dependence on natural gas puts the region in a vulnerable position, especially during cold weather, because the current pipeline infrastructure cannot deliver all the gas required to serve both heating customers and power generators," the Independent System Operator of New England (ISO-NE) said in its 2014-2015 Winter Outlook. "Most gas-fired generators do not have firm contracts for natural gas delivery and instead rely on the release of spare pipeline capacity from gas utilities. 

"With increased residential and business conversions to natural gas for heating, spare pipeline capacity is often not available for power plants." 

The call for more natural gas infrastructure has become a common plea from the region in recent years (see Daily GPI, Oct. 31; Sept. 30; Jan. 23). New England's governors last year vowed cooperation to improve the region's energy infrastructure (see Daily GPI, Dec. 6, 2013).

Last winter, demand for electricity in ISO-NE's footprint peaked at 21,453 MW on Dec. 17, and "periods of sustained cold weather boosted demand for natural gas, causing severe pipeline constraints that led to record-high natural gas prices," ISO-NE said. "As a result, for much of winter 2013/2014, natural gas was often more expensive than oil, which is relatively uncommon."