More Safety Breaker Problems at Junk Plant River Bend.

So where is the reset button to make a plant, fleet, employees and management all new. Where is your reset?  We all deserve a second, third, forth and fifth chance. So where is our next second chance? 

I like riding my mountain bike in the small mountains around my house. I spend a tremendous amounts of time on my bike seat. Being outdoor and in nature reminds me how close we all are to the infinite and god. It is just right there without any deniability. I am heading out to the infinite right now.

It is like coming to a swampy area on my bike. Five or six mosquitoes are on my arm and many more attached to my body. I am getting extremely uncomfortable. I can imagine itching for hours. It's like swatting one mosquito to death on my arm and wondering why I am still being attacked by a cloud of bugs. I should be not thinking stupid thoughts at all, but moving my bike(me)bike out of the swamp.

Honestly, I would never give up on the human spirit and our intelligence. The ability of people to dig a hole nobody can imagine them getting out of. The ability of person at any point, to transform their lives. We are all one higher choice away from a world we dream of. Maybe it not dream at all, but a world god provides for us. It is our choice.  

I don't think this event went down this way they said it did. I think Entergy was worried the NRC would request a shutdown. Or the NRC hinted they might do that. 

Basically by March of 2015 the NRC and Entergy realized they have systemic breaker quality and reliability issues at River Bend. It is unbelievable after all this time and the special inspection, they would discover new breaker problems at River Bend. 

What would a plant reset look like to you?            

Power Reactor	Event Number: 51754
Facility: RIVER BEND Region: 4 State: LA Unit: [1] [ ] [ ] RX Type: [1] GE-6 NRC Notified By: JACK McCOY HQ OPS Officer: STEVE SANDIN	Notification Date: 02/24/2016 Notification Time: 18:16 [ET] Event Date: 02/24/2016 Event Time: 11:00 [CST] Last Update Date: 02/24/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD	Person (Organization): JACK WHITTEN (R4DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	N	0	Cold Shutdown	0	Cold Shutdown

Event Text

ELECTRICAL BREAKER ISSUE IDENTIFIED DURING AN ENGINEERING REVIEW "At 1100 CST on February 24, 2016, with the plant in cold shutdown (Mode 4), the shift manager was notified of a condition that could potentially prevent the automatic closure of the circuit breakers powering the emergency ventilation fans in the both the Division 1 and 2 emergency diesel generator rooms. These fans are not in Technical Specifications, however, they provide a support function to the emergency diesel generators, requiring that both diesel generators to be declared inoperable. This inoperability constitutes a condition that could potentially prevent fulfillment of the safety function of onsite AC power sources, and is being reported pursuant to 10 CFR 50.72(b)(3)(v). "Four additional breakers are affected by the same condition. These breakers supply power to Division 1 and 2 containment unit coolers and the Division 1 and 2 auxiliary building 141 ft. elevation general area unit coolers. The auxiliary building unit coolers are not in Technical Specifications, however, they provide a support function to the electrical distribution system. The Technical Specification required action is to declare both trains of the residual heat removal system (shutdown cooling mode) inoperable. This inoperability constitutes a condition that could potentially prevent the fulfillment of the decay heat removal safety function, and is being reported pursuant to 10 CFR 50.72(b)(3)(v). "Division 2 residual heat removal is operating in shutdown cooling, satisfactorily maintaining reactor coolant temperature. The affected breakers can be manually operated to start/stop their associated equipment, if necessary for operation." This condition was identified during an Engineering review. The licensee has compensatory measures in place. Long term corrective actions are under review. The licensee informed the NRC Resident Inspector.

Special Inspection at River Bend: The Chiller Thriller

(Feb 24, 2016@ 5pm) "Yea well, right after the 2014 Christmas plant trip I noticed the troubles River Bend was having controlling vessel water level. Don’t even get me talking about all the plant trips nationwide in the industry, with them having feedwater control system problems. My favorite recently is Callaway. They partially reverse-engineered an obsolete aux feed water safety controller cards. It is cheaper to replace a card than replace the system (“Failure to assure that the design of the replacement reverse engineered Modutronics controller cards for the auxiliary feedwater control valves were suitable for their application”). 

So I looked at a bunch of LERs about plant trips at River Bend. In most of the trips, the level was unprofessionally banging from the high feed pump trip to the low level scram. I doubt you could call it control. The feed pumps were constantly tripping also. I just couldn’t believe how they were getting away with acting so unprofessional scram after scram. So I made out notes on my blog…I talked to the senior resident explaining the problem. He flipped it into an allegation, I hate the NRC’s allegation
department. They in turn called a special inspection. They discovered known massively leaking feed regulating valves and the simulator had serious fidelity problems. No wonder they didn’t know how to control reactor vessel level. Here is the NRC’s response to me. 

“NRC: Proof I instigated The 2014 Christmas River Bend plant Scram Special Inspection” 

In this response, the NRC says in investigating my vessel level problem, the NRC discovered an almost identical issue with ventilation breaker. The NRC said I caused two special inspections. It startles me to death to think the licensee and NRC can’t figure this out without an outsider provoking them. 

“Junk Plant River Bend’s Crazy February 2016 Power History So Far...” 

Here I am explaining River Bend’s disgusting recent capacity factor problems on my blog. I again called the senior resident inspector. Basically she told me the down-powers and shutdowns are caused by a host of equipment problems. She wouldn’t get specific. She won’t tell me why they shutdown on Fed 16. River Bend is in a lot more trouble than the ROP and inspection reports disclose. 

Since the 2014 Christmas trip River Bend has had three special inspections. It’s got to be a record. The three special inspections are: the 2014 Christmas scram, the magniblast breaker issue and the recent lightning scram where they flubbed putting on shutdown cooling. I am dying to see if River Bend controlled vessel level in the lightning scram?

NRC Allegation response to me: “Based on the multiple failures of the feedwater system, the potential generic concern with the Magne Blast circuit breakers, and the issues related to reactor vessel level between the Level 3 (low) and Level 8 (high) setpoints following a reactor scram, the NRC determined that the appropriate level of NRC response was to conduct a special inspection.”"

Junk Plant Hope Creek: PSEG's Frivolous Denial Of NRC Non Sited Violation

works in progress

Originally posted 2/28

I have felt guilty for weeks with not commenting on this inspection report and the licencee not accepting the NRC's minimalist finding. They are wasting the NRC's time and their own. 

Basically the industry has been recently trying to dial down the CDBI inspections in regulatory reduction. It is as wasting everyone's time.

I think the gist of this stupid is Hope Creek violation was preforming a intentional experiment on the plant. The NRC gave Hope Creek a gift...then they wouldn't accept the gift. They were having troubles with this valve for years with leaking and they didn't have a adequate safe replacement. They gambled if they turned the valve around from normal...had the flow holding the valve shut against pressure, it would fix the leak. They valve was never designed for this duty, basic engineer work...so there was much more friction opening and closing this valve.

I think this is a symtom of a attitude problem with Hope Creek, they are trying to disrupt and weaken NRC authority and oversight over frivolous non sited violation regulatory disputes. It like a cop stops you for speeding and he then lets you off with just a verbal warning.  

February 9, 2016

SUBJECT: RESPONSE TO DISPUTED NON-CITED VIOLATION HOPE CREEK GENERATING STATION – COMPONENT DESIGN BASES INSPECTION REPORT 05000354/2015007 

Dear Mr. Davison: 

We received your response [ADAMS No. ML15362A564] to our inspection report

 05000354/2015007 [ADAMS No. ML15329A157] issued on

• Denial of NCV 05000354/2015007-02, Inadequate Work Order Instructions and Drawings Resulting in Improper Installation of a Safety-Related SW Valve

• HOPE CREEK GENERATING STATION – COMPONENT DESIGN BASES INSPECTION REPORT 05000354/2015007

November 25, 2015, concerning activities conducted at your facility. In your response, you denied a Non-Cited Violation (NCV) contained in the inspection report. Specifically, PSEG contends that “the weaknesses identified in the inspection report regarding classification, evaluation and corrective actions are not more than minor in that PSEG’s conclusion on operability and corrective actions were not impacted.” 

The NRC conducted a detailed review of your response and the applicable inspection guidance. Region I staff who were not involved with the initial inspection effort performed this review. After careful consideration of the bases for your denial of the NCV, we determined that the violation and characterization of the finding were properly described in the inspection report. Specifically, the inspection team identified that your staff failed to conduct operability determinations in October 2013 following the improper installation of a service water system valve and a subsequent event where the valve failed to operate due to a high torque condition. In response to the inspection team’s questions, your staff, with support from an external engineering organization, performed an operability review and determined that the valve was able to perform its safety function for a limited number of operating cycles. You subsequently scheduled a maintenance activity to fully restore the valve during the October 2016 refueling outage. Our review determined that this finding was appropriately characterized as “more than minor” as there was a reasonable basis for questioning operability of the valve following the October 2013 events described above. We have provided a summary of our evaluation and conclusions as an enclosure to this letter. 

SUMMARY  

Non-Cited Violation (NCV) 05000354/2015007-02

Restatement of the Violation:

The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, “Instructions, Procedures, and Drawings,” because PSEG did not provide adequate work order instructions for the installation of service water (SW) pump discharge isolation valve 2198C following planned valve maintenance in October 2013. Specifically, the inadequate work order instructions contributed directly to maintenance technicians installing the valve in the opposite orientation compared to the intended orientation.

Licensee Response (Summary): 

PSEG denies that the NRC identified any new information that impacted the licensee’s conclusions regarding operability or corrective actions. The improper re-installation of the valve EAHV-2198C was promptly identified by the licensee and entered in CAP, and the additional analyses performed in response to NRC questions supported the licensee’s initial conclusions. 

In addition, PSEG contends the identified weaknesses associated with the classification, evaluation, and corrective actions of EAHV-2198C do not meet the threshold for more than minor.

NRC Evaluation: 

The NRC Region I staff performed an independent review of the documented NCV in Inspection Report 05000354/2015007, using PSEG’s basis for denial for comparison, and made the following observations:

1) Notification (NOTF) 20626219 was initiated by the PSEG staff on October 22, 2013, identifying the installation of valve 2198C 180 degrees different than when removed. This NOTF was subsequently updated by the PSEG maintenance staff to reflect inadequacies in the applicable Work Order (60112463-410) as compared to the valve drawing (M-10-1) and the associated vendor manual. The NCV properly refers to 10 CFR 50, Appendix B, Criterion V, as the appropriate violation of regulatory requirements.

2) The Component Design Bases Inspection (CDBI) report states (page 6) that “there was no documented evaluation of the impact of this misalignment and configuration error prior to operations declaring the C SW pump operable following the 2198C maintenance on October 23.” As stated in the report and further clarified by interview with the inspection team leader, PSEG did not complete an operability evaluation prior to restoration of the C SW pump to service on October 23, 2013, in spite of the improper installation of the outlet valve. Further, PSEG acknowledged this deficiency by initiating NOTF 20705874 (also documented in the report). The inspection report (page 7) identified a second instance where PSEG failed to properly assess a valve operational anomaly, an unexpected high opening torque (compared to the valve’s weak link analysis and Limitorque limits) and its potential adverse impact on system operability.This condition was identified during troubleshooting of valve 2198C on October 27, 2013, but no corresponding operability evaluation was documented.

3) The team’s observations documented in the report and highlighted in 2) above, form the basis for the inspection team’s conclusion that the NRC added value to a licensee-identified finding or violation. The absence of an operability evaluation, for either of the above referenced conditions, was documented as a “weakness in the licensee’s classification, evaluation, or corrective action” (page 9) and was the basis for the team to conclude there was a reasonable doubt of operability, with respect to the valve being able to function under all design basis conditions. In order to conclude that the valve was operable and to answer questions from the inspection team, PSEG performed an operability determination for the issue of the valve being installed in the wrong orientation (NOTF 20705874), performed a technical evaluation to determine that the valve actuator was capable of opening the valve under all required design basis conditions based upon actual measured data (NOTF 20704783), and contracted with Kalsi Engineering to perform an H4BC gear box torque analysis for the valve actuator. This team conclusion was documented in the Analysis Section of the inspection report (page 9-10) in reference to the basis for the observations (and underlying performance deficiency) being more than minor. The team cites Example 3.j of IMC 0612, Appendix E, as justification of the more than minor determination. 

4) PSEG’s Basis for Denial did not address the Appendix E more than minor example referenced in the CDBI inspection report. Rather, PSEG stated that the weaknesses identified by the team regarding classification, evaluation and corrective action were not more than minor. Further, PSEG contends that the additional analyses performed in response to NRC team questions did not change the original operability determination outcome. Neither the subjective contention of the weaknesses being minor nor the final determination of operability being maintained provide a sufficient basis for denial. The NRC determined that the Appendix E example established the basis for determining that this performance deficiency was of more than minor significance.

5) In the Basis for Denial, PSEG opines that the CDBI team’s challenge of the operability impact of the above conditions was unfounded without knowledge of the actual operating conditions of the system (system alignment, flowrates, and valve differential pressure were not recorded and were unknown). PSEG contends that without knowing actual system operating conditions, data cannot be extrapolated with any certainty. The NRC considers that the uncertainties associated with the valve operating parameters highlight the reason why PSEG was required to perform an operability determination following improper installation of the valve and its failure to operate due to a high torque condition.

6) Additionally, PSEG took exception to the observation in the inspection report that the maintenance work instructions lacked sufficient detail. On page 6 of the inspection report, the inspection team noted that PSEG maintenance personnel identified and documented in NOTF 20626219, that the desired orientation of the 2198 valve was not specified in valve drawing M-10-1 or in the vendor manual. The inspection team also noted that the work order contained several diagrams which depicted the wrong valve orientation. The NRC determined that the maintenance work instructions given to the maintenance staff lacked sufficient detail to ensure that the valve was installed in the proper orientation.

For the above reasons, the staff concludes that the violation occurred as described in Inspection Report 05000354/2015007.

Junk Plant Hope Creek: Paul Krohn and Safety Related "SW Pump Discharge Isolation Valves"

Updated 2/24

The NRC says now, the three valves are in the normal reverse orientation direction and the offending one is in the non-reverse orientation. All four valves after this 2016 refueling outage will be in the reverse engineering direction. The NRC is convinced all the valves will meet their intended safety function with all the new engineering evaluations. The NRC says they gave Hope Creek much value added with sw discharge valve CDBI inspection results. The NRC was always concerned with the level of engineering detail over these valves.   

The NRC seems to be saying prior to 2013 with all of the valves in the normal reverse orientation…they were always operable. I have questions on that. Looking in the reverse mirror might indicate that, but what proof did they have in 2012?

I asked when these SW discharge valves with reverse orientation were installed. They didn’t know for sure. Where they installed on new construction or did they put them in post construction…could you give me the date? Did the NRC give permission with the reverse orientation? They said it would have to go past their boss.      

Updated 2/24

Talk to NRC today about this.


Update 11:44

Cowan, Grace <Grace.Cowan@nrc.gov> Mr. Mulligan, I received your e-mail, and forwarded it to Paul. He’s working in HQs until June, so I wasn’t sure if he was in the office today or not. Sincerely, Sincerely, Grace Grace Cowan US NRC-Region I Division of Reactor Safety 610-337-5070

To: Michael Mulligan

Today at 11:44 AMMr. Mulligan,  I received your e-mail, and forwarded it to Paul.  He’s working in HQs until June, so I wasn’t sure if he was in the office today or not.  Sincerely,  Sincerely,Grace  Grace Cowan

I had a hard time finding NRC's Chief Mr. Krohn today. He is a region I official. He is temporally assigned to duties in Washington. Ms. Cowan seems to be a secretary of some NRC group in Washington. She gave me her e-mail address and told me to write a note to Paul...she'd pass the message along to him.  

Engineering Branch 2 Chief: Paul Krohn

'Hope Creek and Chief Paul Krohn'

From Michael Mulligan <steamshovel2002@yahoo.com>

To: grace.cowan@nrc.gov

Today at 11:26 AM

 

grace.cowan@nrc.gov

paul krohn

Mr. Krohn,

I am a safety advocate. I had recent issues with Hope Creek’s SRV setpoint lift inaccuracies and my issues were placed on the docket. We are watching Hope Creek very carefully. You signed off on the below IR.

Entry on my blog:

“Junk Plant Hope Creek: PSEG's Frivolous Denial Of NRC Non Sited Violation”

http://steamshovel2002.blogspot.com/2016/02/junk-plant-hope-creek-frivolous-psegs.html

 

My friends and I were debating this issue this morning. We think this issue deserves a much higher violation level. 

So why wasn’t Hope Creek required to ask NRC permission to put all the safety related “sw pump discharge isolation valves (4 of them)” in the “intentionally reverse direction”?  Why didn’t they fulfill the requirements of 50:59? Why wasn’t this in the inspection report? Why wasn't the public immediately informed of this licensing deviation through a safety evaluation?  

We feel if Hope Creek was required to ask NRC permission, they would have taken the easy way out and then put in the proper quality valves for the intended duty. They would have properly fixed these leaking valves when the symptoms first was seen?  It’s pretty pathetic the NRC didn’t immediately flip this 2013 event up into a regular inspection report.  

We think collectively Hope Creek and the NRC has systemic “Normalization of Deviance” on steroids big time???

I believe I have spoken to you in the past and you are one of the good guys. Did we talk about the structure of the CDBIs? The public meeting? Do I remember it right, with all the nuke guys bitching about how useless the CDBIs were, and a diversion from safety?   

Could we have a discussion just about this particular issue?

I never have any confidentiality or anonymity needs what-so-ever.

Mike Mulligan

Hinsdale, NH

16032094206

Hope Creek Generating Station - Component Design Bases Inspection Report 05000354/2015007

Description. 1 EAHV-2198C is the 'C' SW pump discharge isolation valve. The valve is a 28- inch Weir Tricentric butterfly valve with a SMB-1/HBC-4 (60-1) Limitorque motor operator. The valve has an active safety function in the open position to provide normal SW flow to the safetyrelated safety auxiliaries cooling system (SACS) heat exchangers (HXs) and non-1 E reactor auxiliaries cooling system (RAGS) HXs, and emergency SW flow to other systems. PSEG had originally intentionally installed all four 1 EAHV-2198 valves in the reverse flow direction to permit the downstream header pressure to seat the valve tighter to minimize seat leakage during SW pump and strainer on-line maintenance.

Junk Plant FitzPatrick Purchasing Junk New Components.

"The previously installed Temperature Transmitters were Fisher Controls transmitters; model PM511, an original electronic transmitter from the 1970s. Model PM511 is obsolete and replacement parts were running out so an Engineering Equivalent Change evaluated using an alternative model. The new Moore Industries
temperature transmitters, model RBT/3W20-40/4-20mA/117 AC/-EZ84.06-LNP-VTD[EX], were procured from Moore Industries International and Nutherm International performed the Commercial Grade Dedication."

The industry would say this equipment would be needed so infrequently that not shutting down when required by tech Specs is, the risk is insignificant. I am saying if the plant doesn’t pay a steep price for incompetence with purchasing poor quality new safety components( shutdown per tech specs), then the industry will keep calling events like this to their doorstep.

Junk Plant FitzPatrick Purchasing Junk New Component.  

LER: 2015-008 

‘Containment Atmosphere Dilution System Reliability Degraded due to Manufacturer Defect in Temperature Transmitters’

Dec 18, 2015: Part 21notification

July 31, 2015: First Failure

On 11/11/2015, a second failure

On December 18, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when a 10 CFR 21.21(d)(3)(ii) Notification was received from Nutherm International. It identified a defect in Moore Industries temperature transmitters. Specifically, insulation was damaged in the T2 transformer during assembly which could result in premature failure.

These components were installed starting in June 2015 at 27TT-113A and 27TT-113B in the Containment Atmosphere Dilution (CAD) system. The defect caused failures in July and November which resulted in either the “A” or “B” CAD subsystem isolating. Corrective actions included replacing both temperature transmitters with ones that were confirmed to not contain this defect.

Even though these defective temperature transmitters function appropriately until they fail, this defect reduced the reliability of the CAD system to perform its function for its entire mission time. Therefore, this deficiency resulted in a loss of safety function to mitigate the consequences of an accident, reportable per 10 CFR 50.73(a)(2)(v)(D). Also, a single cause affected the safety function of independent CAD trains, reportable per 10 CFR 50.73(a)(2)(vii)(D); and, this condition existed longer then allowed by Technical Specifications 3.6.3.2, reportable per 10 CFR 50.73(a)(2)(i)(B).

Background

The Containment Atmosphere Dilution (CAD) system [EIIS designation: BB] functions to maintain combustible gas concentrations within the Drywell and Torus [BT] at or below the flammability limits following a postulated loss of coolant accident (LOCA) by diluting hydrogen and oxygen with nitrogen. Also, the CAD system provides the pneumatic supply to instruments and controls inside the Drywell; including the long term (100 day) pneumatic supply to the Automatic Depressurization System (ADS) valves [SB] and accumulators following a LOCA…

Event Description 

On December 18, 2015, James A. FitzPatrick Nuclear Power Plant (JAF) was operating at 100 percent power when a 10 CFR 21.21(d)(3)(ii) Notification was received from Nutherm International. It identified a defect in Moore Industries Resistance Temperature Detector (RTD) temperature transmitters supplied to JAF. The transmitters were installed at 27TT-113A and 27TT-113B in the CAD System. A failure of 27TT-113A initiates a closure of CAD subsystem “A” isolation valves 27AOV-128A and 27AOV-129A. A failure of 27TT-113B initiates a closure of CAD subsystem “B” isolation valves 27AOV-128B and 27AOV-129B. Isolating either CAD subsystem impacts the containment makeup capability and the Instrument Nitrogen Header. An Operability Evaluation determined that this defect reduced the reliability of the CAD system such that it may not be able to meet its full 100 day mission time.

This condition was reported to the NRC per 10 CFR 50.72(b)(3)(v)(D), ENS 51613, as a condition which could have prevented the fulfillment of a safety function to mitigate the consequence of an accident.

Event Analysis

The previously installed Temperature Transmitters were Fisher Controls transmitters; model PM511, an original electronic transmitter from the 1970s. Model PM511 is obsolete and replacement parts were running out so an Engineering Equivalent Change evaluated using an alternative model. The new Moore Industries temperature transmitters, model RBT/3W20-40/4-20mA/117 AC/-EZ84.06-LNP-VTD[EX], were procured from Moore Industries International and Nutherm International performed the Commercial Grade Dedication.

The new Moore temperature transmitter contained a defect which first failed on 7/31/2015 in 27TT-113A.

When 27TT-113A failed the “A” CAD subsystem automatically isolated by closing the supply valve 27AOV 128A and backup valve 27AOV-129A. This failure was initially identified as component infant mortality and the component was replaced with a spare component.

On 11/11/2015, a second failure of a new Moore temperature transmitter occurred at 27TT-113B. In a similar fashion to the 7/31/2015 failure, “B” CAD subsystem automatically isolated by closing the supply valve 27AOV-128B and backup valve 27AOV-129B. Based on this now being the second failure of newly installed components, there was increased scrutiny placed on this failure. The components were returned to Nutherm to perform a failure analysis.

The December 18, 2015, Part 21 notification from Nutherm informed JAF that the wire insulation in T2 transformer on the Moore Industries RTD temperature transmitter was damaged during assembly. This damage reduced the insulation resistance and dielectric breakdown between the windings of the transformer; resulting in premature failure of the temperature transmitter.

It has been determined that no visual inspection of the transformer or testing after the transformer is installed will discover this defect. This defect can only be found by performing testing on the transformer prior to installation.

The investigation into the two failures on 7/31 and 11/11 became the driving force which led to the Part 21 Notification on 12/18. The replacement spares installed after the two failures were the two serial numbered components included in the Part 21 Notification made to JAF. The removed failed components were also determined to contain the same defect as the Part 21 Notification.

With no method available to inspect the potentially degraded subcomponent after installation, there is not a reasonable assurance the two components would have been able to meet their required 100 day mission time. This condition could have prevented the fulfillment of a safety function, reportable per 10 CFR 50.73(a)(2)(v)(D). This condition caused two independent trains to be Inoperable in a single system designed to mitigate the consequence of an accident, reportable per 10 CFR 50.73(a)(2)(vii)(D). Technical Specification (TS) 3.6.3.2 requires the CAD system to have two Operable subsystems. On June 17, 2015, when the first defective temperature transmitter was installed, “A” subsystem was Inoperable. Required Action A.1 of the TS requires that it be restored within 30 days. On July 2, 2015, a second CAD subsystem became Inoperable when a second defective temperature transmitter was installed. Required Action B.2 requires that one CAD subsystem be restored within 7 days. When this was not accomplished, Required Action C.1 requires that the plant be in Mode 3 within 12 hours. Since JAF was not in Mode 3 within the required completion time, this event was a condition prohibited by Technical Specifications, reportable per 10 CFR 50.73(a)(2)(i)(B).

Cause

The cause of this event is a defect introduced during the manufacturing of temperature transmitters 27TT- 113A and 27TT-113B installed in the Containment Atmosphere Dilution system which resulted in decreased system reliability.

  

Junk Plant June 2012 Announcement of Special Unit Investigation

NRC special unit launches investigation into potential wrongdoing at Palisades nuclear plant related to leaking tank

By Fritz Klug | fklug@mlive.com

on June 27, 2012 at 6:35 PM, updated June 27, 2012 at 8:58 PM
 COVERT, MI — A special unit of the Nuclear Regulatory Commission is investigating whether there was any wrongdoing related to how Entergy Energy handled a leaking cooling tank at the Palisades nuclear power plant.
 The investigation was launched Tuesday by the NRC's Office of Investigations, according to NRC spokeswoman Viktoria Mitlyng. Entergy shut Palisades down two weeks ago after a leak in a cooling tank surpassed a minimum level of 31 gallons a day. Mitlyng said Wednesday that the NRC cannot disclose details of the investigation because it is ongoing. Entergy spokesman Mark Savage would not comment on the investigation. Instead of looking for performance deficiencies at the

Why is the normal inspectors only relegated to only performance issues?  

plant as is typical of NRC investigations, the Office of Investigations looks at whether there was any "potential wrongdoing" in how the company handled a situation and the its participation in an inspection…

52 shares

Junk Plant Palisades Safety Culture

So basically on top of the control room sat the SIRWT. On top of of the SIRWT was the leaking roof. The roof and the SIRWT were both leaking. It looked to me the confusion with the leaking roof was intentional in covering up the Leaking SIRWT. They were certainly trying to disrupt oversight and delaying the repair of the SIRWT until proper planning and service were brought to the site.

It would be interesting when the NRC was first aware falsification was ongoing.

On June 25, 2012, the U.S. Nuclear Regulatory Commission’s Office of Investigations initiated an investigation to determine whether personnel at the Palisades Nuclear Power Plant(Palisades) deliberately failed to provide complete and accurate information to the NRC regarding a safety injection and refueling water storage tank (SIRWT) leak. The investigation was completed on March 10, 2015.

Would the outcome be different on my 2.206 if the NRC disclose four individuals were being prosecuted for falsifying documents on the SIRWT instead of three year later?

Mike Mulligan's 2.206 on the Palisades SIRW tank. 

November 20, 2012

Mr. Michael Mulligan

P.O. Box 161

Hinsdale, NH 03451

Dear Mr. Mulligan: 

You recently submitted two petitions addressed to Mr. William Borchardt, Executive Director for Operations at the U.S. Nuclear Regulatory Commission (NRC). The petitions were referred to\ the NRC’s Office of Nuclear Reactor Regulation pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.206.

In your first petition dated June 18, 2012, as revised on June 27, 2012, you requested that the Palisades Nuclear Plant (Palisades) remain shutdown. In the petition, you were critical of Entergy Nuclear, the NRC, and the programmatic aspects of the regulatory program, including the Reactor Oversight Process (ROP). You focused on a leak of the Safety Injection Refueling Water (SIRW) tank at Palisades, but also discussed past events at both Palisades and other

Entergy-owned facilities. Finally, you also discussed the lack of an adequate safety culture environment at Palisades.

You requested that the following actions be taken:

(1) The shutdown resulting from the SIRW tank leak should be categorized as unplanned. 

(2) The NRC should move the Palisades performance indicator from Red to the next level V:Unacceptable Performance. 

(3) An outside authority (not the NRC’s Office of the Inspector General (OIG)) should determine why the NRC did not force Palisades (Entergy) to thoroughly investigate the SIRW leak when the leak first appeared. 

(4) Top Palisades management staff should be fired and replaced before startup. 

(5) Entergy’s corporate nuclear senior staff should be fired and replaced before restart. 

(6) The NRC should assign two additional NRC inspectors to Palisades and to the rest of the Entergy nuclear plants. 

(7) A local public oversight panel should be formed around every plant. 

(8) An emergency NRC senior official oversight panel should be convened to reform the ROP.

(9) A national NRC oversight panel of outsiders (consisting of a mixture of professional and academic people, as well as lay people) should be convened to oversee and report on agency activities.

(10) The NRC should perform an analysis to determine the cause of the numerous findings of problems with Entergy plants during this inspection reporting cycle.

(11) The NRC should evaluate if Region III has enough personnel and resources.

(12) Palisades should remain shutdown until all procedures are fully updated and corrected, all technical and maintenance backlogs are updated and corrected, all training completed, and all reports and safety processes are fully completed and implemented. 

(13) An independent outside investigation should review the insufficient process outcome of the 2008-2009 Palisades security falsification, investigation, safety survey local and fleet-wide training and safety surveys.

(14) President Obama should fire Chairman Jazcko and the four Commissioners.

In the second petition dated June 28, 2012, you requested that Palisades remain shutdown. 

This petition was focused on roof leaks at Palisades, but also discussed past events at both Palisades and other Entergy-owned facilities. You discussed a lack of adequate safety culture environment at Palisades, and were also critical of the NRC staff for “tolerating and covering up” very serious safety problems at Palisades and throughout the Entergy organization. This petition also included specific questions related to roof leaks.

This petition duplicated many of your requests discussed in the previous petition. However, in your second petition there were new requests which are provided below:

(15) Entergy should be prevented from starting up until all the safety problems at the site have been publicly identified and the safety culture repaired. 

(16) Heads need to roll in Region III and at headquarters for tolerating and covering up these very serious safety problems at Palisades and throughout the Entergy organization. This all has the potential to gravely damage our nation.

(17) The NRC should report on why the 2.206 petition process failed, and for the agency to hold officials accountable to the plant employees and me with not doing their jobs in trying to understand what was going on at the site and not repairing the organization at the earliest point.

(18) A meeting with the Palisades inspector and other…

Junk Grand Gulf Capacity factor

Feb 23 0%
22 0%
21 0%
20 0%
19 95%
18 95%
17 95%
16 95%
15 96%
14 96%
13 96%
12 96%

Usually it is in a paper if a plant goes into refueling. I would not spend so much time on this if Entergy reported the  refueling in the paper.

Sounds like they got failed fuel pin failures????

Grand Gulf has been saying for almost a year they are coasting down to refueling. I guess you can say there is a slow decline in the power level. They problem is this has been going on for about three months. 

Dec 23, 2015:

Grand Gulf 1

98

HOLD FOR SETUP FOR POWER SUPPRESSION TESTING

Dec 11

97

INCREASING POWER

*

Seems Grand Gulf has been having trouble keeping their plant at power.

GRAND GULF NUCLEAR STATION - NRC 95001 SUPPLEMENTAL INSPECTION REPORT 05000416/2014009 (Aug 2014)  

Dear Mr. Mulligan (not me): On June 20, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed a supplemental inspection at your Grand Gulf Nuclear Station. The enclosed report documents the results of this inspection, which were discussed with you and members of your staff, during an exit meeting on June 20, 2014, as well as during the re-exit meeting on August 6, 2014, with members of your staff. As required by the NRC Reactor Oversight Process Action Matrix, this supplemental inspection was performed in accordance with Inspection Procedure 95001, “Supplemental Inspection for One or Two White Inputs in a Strategic Performance Area.” The purpose of the inspection was to examine the causes for, and actions taken related to, a White Performance Indicator in the Initiating Events Cornerstone at Grand Gulf Nuclear Station. The performance indicator was for Unplanned Reactor Scrams per 7,000 Critical Hours and crossed the Green-to-White threshold during the first quarter of 2013. The performance indicator value was noted as 3.2. This inspection also reviewed the details of all five licensee event reports that were submitted to the NRC for unplanned scram events that occurred between the dates of December 29, 2012 and March 17, 2014. There was an additional unplanned scram event that occurred on March 29, 2014, but due to a vendor review in process, the root cause evaluation was not complete for this inspection period. Thus, the licensee event report for that event will not be addressed in this report.

201501 

Corrective actions associated with the adverse trend are:

• The licensee has taken action already to increase staff allocation in the electrical field so that there are more staff to accomplish the preventative maintenance tasks. Currently, the licensee has identified that there is a shortage of electrical workers and is actively working to increase staff.

November 13, 2015

SUBJECT: GRAND GULF NUCLEAR GENERATING STATION, UNIT 1– NRC

COMPONENT DESIGN BASES INSPECTION REPORT 05000416/2015007


Dear Mr. Mulligan:

On October 1, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Grand Gulf Nuclear Station Unit 1. On August 27, 2015, the NRC inspectors discussed the preliminary results of this inspection with you and other members of your staff. On October 1, 2015, the NRC inspectors discussed the final results of this inspection with you and other members of your staff.

Inspectors documented the results of this inspection in the enclosed inspection report. The NRC inspectors documented seven findings of very low safety significance (Green) in this report. All of these findings involved violations of NRC requirements; one of these violations was determined to be Severity Level IV under the traditional enforcement process. Additionally,

the NRC inspectors documented three Severity Level IV violations with no associated finding.

The NRC is treating these violations as non…

Security related violations and a security related OI investigation are always a symptom a plant is running away from management. 

GRAND GULF NUCLEAR STATION – NRC SECURITY INSPECTION REPORT 05000416/2015404

Dear Mr. Mulligan:

On October 8, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a security inspection at the Grand Gulf Nuclear Station. An NRC inspector discussed the results of this inspection with you and other members of your staff. The inspector documented the results of this inspection in the enclosed inspection report. NRC inspectors documented three findings of very low security significance (Green) in this report. All of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NVCs) consistent with Section 2.3.2.a of the Enforcement

Policy
 

Dear Mr. Mulligan:

This letter refers to the investigation conducted by the U.S. Nuclear Regulatory

Commission's Office of Investigations, Region IV, at Grand Gulf Nuclear Station; Inspection Report 05000416/2015406 enclosed. The purpose of the investigation was to determine if there was a willful security-related violation at the Grand Gulf Nuclear Station. Following notification by Grand Gulf Nuclear Station staff of a potential willful security-related violation the NRC initiated an investigation on February 26, 2014. The investigation was completed on February 25, 2015. Based on the evidence developed during the investigation, the NRC

determined that a willful security-related violation occurred. The enclosed inspection report documents the inspection results, which were discussed on July 16, 2015, with Paul Salgado, Acting Director of Regulatory Assurance and Performance Improvement, and other members of your staff.