Pilgrim: White Safety Relief Valve Problems

August 26, 2015
EN 15-024 

OFFICE OF ENFORCEMENT NOTIFICATION OF SIGNIFICANT ENFORCEMENT ACTION

Subject: ISSUANCE OF FINAL SIGNIFICANCE DETERMINATION AND NOTICE OF VIOLATION

This is to inform the Commission that a Notice of Violation (NOV) associated with a White Significance Determination Process finding will be issued on or about September 1, 2015, to Entergy Nuclear Operations, Inc. (licensee) as a result of an inspection at Pilgrim Nuclear Power Station (Pilgrim).

The White finding involved the licensee’s failure to establish measures to promptly identify and correct a significant condition adverse to quality, or take corrective actions to preclude repetition, relating to a component that is essential to perform the Automatic Depressurization System (ADS) safety-related functions. Specifically, the licensee failed to identify that the ADS ‘A’ safety/relief valve (SRV) did not open upon manual actuation on February 9, 2013. The licensee therefore did not take action to preclude repetition, which resulted in the failure of the ADS ‘C’ SRV to operate upon manual actuation on January 27, 2015. Also, because the licensee was not aware of the ‘A’ SRV’s inoperability from February 9, 2013, until January 27, 2015, a period greater than the allowed Technical Specification (TS) outage time, the required actions of the TS were not followed.

A NOV is included based on the licensee’s failure to establish measures to assure that
conditions adverse to quality are promptly identified and corrected, and failure to assure that the cause of the condition is determined and corrective action taken to preclude repetition in accordance with 10 CFR 50, Appendix B, Criterion XVI, “Corrective Action”.

Oyster Creek: Yellow Safety Relief Valve Problems.

This is probably the model for end-of-life plants in the future. The plant is operating with obscenely obsolete equipment and in its closing years it is just not worth wasting money on a dying plant.

You catch here with these severe safety relief valve (electromatic relief valve) problems it seems to only occur in plants who are severely troubled and many other component have been implicated in degradations.  

What I never got, it was a initial design defect...why did it only show up at end of life?

OFFICE OF ENFORCEMENT NOTIFICATION OF SIGNIFICANT ENFORCEMENT ACTION Subject: ISSUANCE OF FINAL SIGNIFICANCE DETERMINATION AND NOTICE OF VIOLATION

This is to inform the Commission that two separate Notices of Violation (NOV), one associated with a Yellow Significance Determination Process (SDP) finding and one associated with a White SDP finding, will be issued on or about April 27, 2015, to Exelon Generation Company, LLC (Exelon) as a result of separate inspections at its Oyster Creek Nuclear Power Station. The Yellow finding represents an issue of substantial safety significance. The White finding represents an issue of low to moderate safety significance. These findings will result in additional NRC inspection and potentially other NRC action.

The Yellow finding involved the failure by Exelon to establish adequate measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the electromatic relief valves (EMRVs). Specifically, since original installation of the EMRVs in 1969, until the valves were redesigned and reinstalled during the 2014 refueling outage, the EMRV actuators were inadequate because when they were placed in an environment where the actuator was subject to vibration associated with plant operation, the mechanical tolerance between posts and guides created a condition where the springs could wedge between the guides and the posts, jamming the actuator plunger assembly. In addition, given the original design of the valve, the maintenance refurbishing processes were not adequate to maintain the required internal tolerances to prevent excessive fretting and wear of the internal components. As a result, the staff determined that two EMRVs were inoperable for greater than the allowed Technical Specification outage time of 24 hours.

The White finding involved the failure by Exelon to review the suitability of a new emergency diesel generator (EDG) belt maintenance process that was essential to a safety-related function of the EDGs and to verify the acceptance criteria of that process. Specifically, from May 13, 2005, to September 9, 2014, Exelon changed the method for tensioning the cooling fan belt on the EDG from measuring belt deflection to measuring belt frequency and did not verify the adequacy of the acceptance criteria stated for the new method. As a result, the specified belt frequency imposed a stress above the fatigue endurance limit of the shaft material, making the EDG cooling fan shaft susceptible to fatigue and failure which occurred on July 28, 2014. As a result, the staff determined that EDG No. 2 was inoperable for greater than the allowed Technical Specification outage time of 7 days...

Dresden: More White Safety Relief Valve Problems?

This was a prolong event with uncontrollable quality in the safety relief valves (electromatic relief valve). I don't think the white finding is significant enough to create a organizational behavior change across Pilgrim, Dresden or Oyster Creek. 

September 16, 2015
EA-15-115

SUBJECT: FINAL SIGNIFICANCE DETERMINATION OF WHITE FINDING AND NOTICE OFVIOLATION; NRC INSPECTION REPORT NO. 05000237/2015010; DRESDEN NUCLEAR POWER STATION

This letter provides you the final significance determination of the preliminary White finding discussed in our previous communication dated July 1, 2015, which included U.S. Nuclear Regulatory Commission (NRC) Inspection Report No. 05000237/2015002; 05000249/2015002; 07200037/2015001. This report is available in the NRC’s Agencywide Documents Access and Management System (ADAMS) at Accession Number ML15219A500. The finding involved the failure of the Unit 2 “C” electromatic relief valve (ERV) to perform its intended safety function. 

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems, and components. Technical Specification 3.4.3, Safety and Relief Valves, Limiting Condition for Operation requires, in part, that in Modes 1, 2, and 3, the relief function of five relief valves shall be OPERABLE. Required Action A states that if one relief valve is inoperable, then restore the valve to operable status within 14 days. Required Action B states, in part, that if the Required Action and associated Completion Time are not met, then (1) be in Mode 3 within 12 hours and (2) be in Mode 4 within 36 hours.

Technical Specification 3.5.1, ECCS Operating, Limiting Condition for Operation requires, in part, that in Modes 1, 2, and 3, with pressure above 150 pounds per square inch gauge (psig), the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE. Required Action H, states that, if one ADS valve is inoperable, then restore the valve to operable status within 14 days. Required Action I states, in part, that if the Required Action H and associated Completion Time are not met, then (1) be in Mode 3 within 12 hours, and (2) reduce reactor steam dome pressure to less than 150 psig within 36 hours.

Contrary to the above, from December 1, 2009, to February 7, 2015, the licensee failed to establish measures for the review of suitability of application for the ADS electromatic relief valve (ERV) actuators, which are essential to perform the safety-related reactor vessel depressurization and overpressure protection functions. This resulted in a failure of the 2C ERV, and an indeterminate period of inoperability and unavailability greater than allowed by Technical Specifications 3.4.3 and 3.5.1 during operating cycle D2C24. The 2C ERV inoperability during the operating cycle was identified after the failure of the valve during its first operational test in mid-cycle outage D2F56. Additionally, because the licensee was not aware of the valve’s inoperability between 2013 and 2015 during operating cycle D2C24, the required actions in Actions 3.4.3 A and B, and 3.5.1 H and were not followed.

What the Hell is Wrong with the McGuire Plant?

9/16/15 11 AM

05000369/370

I never get anything bad NRC reports about these guys.  They seem to be able to stay up at power. No scram problems. They are the kid in the class you never hear a peep out of? He got to be sick. A two unit site, we should be hearing about something all the time?

How the hell do these guys stay out trouble?

Sitting on a beautiful Lake Norman, a few miles from Charlotte and modern civilization and shopping malls. It must be sweet living. The only drawback is it is in the south.

Not a big deal yet...cool coincidence?

Update 9/18

Not a big deal yet...cool coincidence though?

Power Reactor	Event Number: 51406
Facility: MCGUIRE Region: 2 State: NC Unit: [1] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: RYAN WHISNANT HQ OPS Officer: VINCE KLCO	Notification Date: 09/17/2015 Notification Time: 21:32 [ET] Event Date: 09/17/2015 Event Time: 16:10 [EDT] Last Update Date: 09/17/2015
Emergency Class: NON EMERGENCY 10 CFR Section: 26.719 - FITNESS FOR DUTY	Person (Organization): MARVIN SYKES (R2DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	100	Power Operation	100	Power Operation
2	N	N	0	Refueling	0	Refueling

Event Text

MINIATURE ALCOHOL BOTTLE DISCOVERED INSIDE THE PROTECTED AREA "A miniature alcohol bottle, containing trace amounts of liquid, was discovered inside the protected area. Site security took possession of the bottle and removed it from the protected area." The licensee notified the NRC Resident Inspector.

Palo Verde: Electric Breaker Rapid Combustion : )

Can you trust the officials at Palo Verde nuclear Power plant Ever Again?

This was a electrical explosion plain and clear.

Prettifying equipment failures: I now don't trust these guys...

So it looked and sounded and they entered a explosion classification procedure....but its actually rapid combustion. Only in a nuclear plant. They are they too timid to put the truth down on paper. How widespread is prettifying documents at Palo Verde?

I think it demeans the  people who had to respond to this and the potential they all face in a power plant 

As a result, an Emergency Classification of HU2.2, EXPLOSION was declared due to the Load Center breaker failure and noise and visible indication observed in the field. 

Power Reactor	Event Number: 51403
Facility: PALO VERDE Region: 4 State: AZ Unit: [ ] [2] [ ] RX Type: [1] CE,[2] CE,[3] CE NRC Notified By: ROBERT PIERCE HQ OPS Officer: JEFF HERRERA	Notification Date: 09/17/2015 Notification Time: 02:53 [ET] Event Date: 09/16/2015 Event Time: 23:01 [MST] Last Update Date: 09/17/2015
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(a) (1) (i) - EMERGENCY DECLARED	Person (Organization): THOMAS FARNHOLTZ (R4DO) WILLIAM GOTT (IRD) MARC DAPAS (RA) SCOTT MORRIS (NRR) WILLIAM DEAN (NRR)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
2	N	Y	100	Power Operation	100	Power Operation

Event Text

NOTIFICATION OF UNUSUAL EVENT DUE TO RAPID COMBUSTION OF A LOAD CENTER BREAKER "The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event, or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. "Non Class Load Center Breaker, 2ENGN-L04 failed, resulting an a visible observation of rapid combustion and resultant charring (burned area) of the breaker enclosure and housing. No physical deformation to the breaker housing or surrounding area has been identified. The rapid combustion self-extinguished immediately following the audible and visible combustion event. As a result, an Emergency Classification of HU2.2, EXPLOSION was declared due to the Load Center breaker failure and noise and visible indication observed in the field. "The plant was, and continues to operate at 100% full power operations on normal power alignment. The 2ENGN-L04 Non-Class Load Center breaker supplies power to non-essential service loads and has no immediate impact to plant operation or safety mitigating systems. The plant remains stable and the event did not adversely affect the safe operation of the plant or health and safety of the public. "The NRC Resident Inspector has been notified."

A Poor Maintenance Fiasco at Fermi

Losing an air compressor is a nasty accident because there is many air operated valves. 

Three important pumps lost and discovered two valve that are broken? I junk plant.

***It looks like the plant simulator didn't model single LOOP operation...

These guys should have made the decision to conservatively shutdown before doing this runaway maintenance monster. It is a outright mania trying to keep these plants up a power no matter how degraded the plant is.

***A pattern of panic scrams and losing control of the cooling systems. Again, you catch how two poor maintenance problem led to the scram. 

On March 19, 2015 the unit automatically scrammed due to actuation of the Reactor Protection System function of OPRM Upscale. The unit had just transitioned to single loop operation after operators secured a reactor recirculation pump due to the loss of its normal and emergency cooling water supply.

***The LER: On March 19, 2015, at 0647 hours, the Fermi 2 annunciators indicated a cooling water leak in the drywell. The Reactor Building Closed Cooling Water (RBCCW) system [[CC]] was isolated and both divisions of the Emergency Equipment Cooling Water (EECW) system were started. Approximately four minutes later signs of Division 1 EECW pump cavitation were observed indicating that the leak affected the north (A) reactor recirculation pump [[AD]] cooling. The north (A) reactor recirculation pump was tripped at 0652 to prevent motor damage from loss of cooling and the reactor transitioned. 

Sounds like they had a water hammer... 

PRELIMINARY NOTIFICATION

September 15, 2015

PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE - PNO-III-15-009

This preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public interest significance. Some of the information may not yet be fully verified or evaluated and is basically all that is known by the Region III staff on this date.

 Fermi Power Plant (Fermi 2)
DTE Energy Company
Newport, MI

SUBJECT: UNPLANNED SHUTDOWN GREATER THAN 72 HOURS FOLLOWING THE LOSS OF COOLING WATER IN THE TURBINE BUILDING

 At 11:05 p.m. (EDT) on September 13, 2015, operators manually shut down (scrammed) Fermi 2 from 100 percent power after attempts to correct a malfunction of a nonsafety-related cooling water system in the turbine building were unsuccessful and the three pumps in the system shut down. The malfunction arose earlier that evening while operators were working on the cooling water system heat exchangers. 

All control rods inserted and all plant systems responded normally to the scram. The loss of the nonsafety-related cooling water system resulted in the loss of the station compressed air system, which affected valves that are operated by air. Operators took the appropriate actions for those affected valves.

The licensee is investigating the cause of the malfunction and will be repairing two valves that were identified as in need of repair. On September 14, the licensee successfully restarted the nonsafety-related cooling water system. The licensee stated that the reactor would be shutdown for at least 72 hours.

The NRC resident inspector was notified of the scram, promptly responded to the plant, and monitored activities throughout the early morning. The NRC is currently monitoring the troubleshooting, repair, and restart activities.

River Bend: Safety-Conscious Work Environment

RIVER BEND STATION – NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000458/2015008

The licensee maintained a safety-conscious work environment in which personnel were willing to raise nuclear safety concerns without fear of retaliation.

 

Supposedly they allow their employees to raise any safety concern without fear of retaliation…but they just don’t listen to them. The blow the employees off.

So an individual would write up CR or other document…the manger would sign off with some hokey response without the first writer seeing it or wanting to see it. The first writer would think I cleared my conscience, even knowing the problem wasn’t cleared. Entergy is famous for this across the fleet.  

The NRC doesn’t inforce falsification or a intent to be deceptive or falsification.

I like the NRC to inspect, employees raise safety concerns and Entergy immediately fixes it. 

SEQUOYAH Indicates Big Problem with 10 CFR 50:59s

Why do I get the feeling these nuclear don’t fear the NRC at all?

Why do I think the NRC is overwhelmed by 10 CFR 50:59s. Just throw in a bum 50:59s or none all…the NRC will never discover it. If they discover it, it will just be inconsequential ding.

Remember this is just a sample of 18 screening. We never get an idea of the total screening. They do have a list screening on the dockets. 

Evaluations of Changes, Tests, and Experiments: The inspectors reviewed eight safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59, “Changes, tests, and experiments,” to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 18 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

You see the extremely dangerous reactionary stance of the agency…they discovered these defects many years after the change or new gear.

The NRC isn’t up these nuke plant asses, don’t you dare make that change without doing a proper 50:59 or accurate comprehensive safety evaluation when the 50:59 is ongoing. 

SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS INSPECTION REPORT 05000327/2015007 AND 05000328/2015007

NRC inspectors documented five findings of very low safety significance (Green) in this report.

Five of these findings involved violations of NRC requirements; one of these violations was determined to be Severity Level IV under the traditional enforcement process.

Indian Point Pressurizer Safety Valve: a 66% Failure Rate

 Safety Valves In Our Nuclear Power Plants Going Wild

 

These are the guys with all the scrams and shutdowns. 

Condition Due to Two Pressurizer Code Safety Valves Discovered Outside Their As-Found Lift Setpoint Test Acceptance Criteria"Indian Point Unit No. 3 Docket No. 50-286 DPR-64

On July 1, 2015, Engineering was notified by Wyle Laboratories that two of three Pressurizer {AB} Code Safety Valves (RC-PCV-464 and RC-PCV-468) {RV} removed during the spring 2015 refueling outage (RO) failed their As-Found lift set point test acceptance criteria (2411 - 2559 psig) . The As-Found set pressure testing acceptance criterion for operability is 2485 +/-3%. The SVs were .removed during the last refueling outage (RO) in the spring of 2015 and sent offsite for testing. Testing was performed within one year of removal as required by the Inservice Testing Program. Testing found SV RCPCV- 464 as-found lift pressure was 2573 (0.5% above the allowable As-Found *upper limit of 2559 psig), and SV RC-PCV-468 as-found lift pressure was 2379 psig (1.2% below allowable AS-Found lower limit of 2411 psig), which is outside their set pressure range acceptance criterion. The remaining SV lift tested satisfactorily. All three SVs were found with zero seat leakage. During the RO all three SVs were removed and replaced with certified pre-tested spare SVs. The SVs installed during the RO were As-Left tested to 2485 +/-1% with zero seat leakage in accordance with procedure 3-PT-R5A. Technical Specification (TS) 3.4.10 (Pressurizer Safety Valves), requires three pressurizer safety valves to be operable with lift settings set at greater than 2460 psig and less than 2510 psig. TS Surveillance Requirement (SR) 3.4.10.1 requires each PSV to be verified operable in accordance with the Inservice Testing Program. The condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) in Condition Report CR-IP3-2015-03710 and CR-1P3-2015-03708.

The pressurizer safety valves (SVs) are totally enclosed pop type, spring loaded, selfactuating 6 inch by 6 inch valves manufactured by Crosby Valve Company {C711}, Model HB-BP-86 Type E. The SVs are designed to prevent the system pressure from exceeding the system Safety Limit (SL) of 2735 psig, which is 110% of the design pressure.

The Cause of Event

The exact cause of failure of valves RC-PCV-464 and RC-PCV-468 is not known at this time. The most probable cause of By RC-PCV-464 lifting greater than 3% of its nominal setpoint was setpoint drift. The most probable cause of RC-PCV-468 lifting within less than 3% of its nominal setpoint was spring relaxation. The two valves that failed As-Found testing criteria (valve RC-PCV-464 and RC-PCV-468) will be disassembled and inspected to determine the cause of the failure.

They can't see the degradation at power. If they seen it, they would quickly have to shutdown per tech specs. No rush worrying about about if it is spring. Remember in Pilgrim, they damage the spring on the test stand. 

I think it is a plant design defect. If there were four safety valves, then they never would have crossed a tech spec shutdown. I guess only some required tech spec conservative shutdowns are applicable.   

TS 3.4.10 Condition B (Required action and associated completion time not met or Two or more pressurizer safety valves inoperable) required action B.l is be in Mode 3 in 6 hours and B.2 be in Mode 4 within 12 hours. This TS action was not performed and the actions of Condition B not implemented, the condition is a TB prohibited condition. In the UFSAR Chapter 14 analysis, the opening setpoint of the three PSRVs is assumed to be at +/-4% of the nominal 2485 psig value for applicable Chapter 14 transients…

Professional Reactor Operator Society

Nucpros by Robert Meyer...man he is a old work horse.

Why did his internet site go black?


It is like they run their plants?

Frank Maciuska

September 9 at 9:19am

Is anyone else having a problem getting to "nucpros.com"?

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Comments

• 

  Professional Reactor Operator Society The website has been having some issues lately! I'll follow up with our office manager and see if there's a projected fix date

 

UCS: Fort Calhoun and River Bend

The way we should should look at these thing; what if we had a hundred 2011 Fort Calhouns running like this in the USA? What if we had a 100 River Bends cira 2014? Fort Calhoun and River Bend are in the same region IV. It is very trouble region. The nrc have a unique solution for this, they slough off work through plants like San Onophe through permanent shutdown. The NRC let San Onophe run to ground allowing them to run like Fort Calhoun for decades. Very quickly we'd come up with a meltdown, partial meltdown or industry ending scandal.  

I contend with 10 to 20 bad actor plants like River Bend and  Fort Calhoun running at the same time, the NRC gets over burdened and become effectively blinded...  

Hence, with San Onophe, Fort Calhoun, River Bend, Waterford and the rest...region IV was over overburdened and they missed a lot...

The Saturday Night Live Approach to Nuclear Safety: More Cowbell!

Dave Lochbaum, director, Nuclear Safety Project

September 15, 2015

  

Fission Stories #197

The April 8, 2000, Saturday Night Live broadcast featured a skit with cast members pretending to be the rock group Blue Oyster Cult in the recording studio with a famous music producer, played by actor Christopher Walken. The skit is remembered for Walken’s character stating “I gotta have more cowbell.”

The NRC’s Reactor Oversight Process (ROP) needs more cowbell, too.

The Fort Calhoun nuclear plant shut down in April 2011 for a refueling outage. The outage was planned to last a handful of weeks while workers replaced spent fuel assemblies with new assemblies and performed routine maintenance and testing activities. The plan went awry when the ROP identified safety problems that needed to be corrected before the reactor could be restarted.

The operators restarted Fort Calhoun in December 2013 after a short refueling outage morphed into a 32-month safety restoration outage. On March 30, 2015, the NRC announced that it was returning Fort Calhoun to normal handling under the ROP. The NRC also reported expending over 60,000 hours since December 2011 on inspection, assessment and licensing tasks at Fort Calhoun.

60,000 hours is a number without context. To help put this value in context, the NRC reported having expended 6,652 hours, 6,612 hours, and 6,782 hours of total oversight effort at the average nuclear plant in 2011, 2012, and 2013, respectively. So the average nuclear plant received an average of 6,682 hours of oversight from the NRC annually.

Between 2012 and 2014, Fort Calhoun received an average of 18,462 hours of oversight effort each year from the NRC.

Thus, Fort Calhoun received the equivalent of 2.76 nuclear plants’ worth of regulatory oversight attention from the NRC between 2012 and 2014.

Too Little Much, Too Late

Figure 1 shows where the NRC placed Fort Calhoun within the ROP’s Action Matrix each quarter from the inception in the fourth quarter of 2000 until late 2014. 

Fig. 1 (click to enlarge) (Source: UCS)

When performance fell within expected ranges, Fort Calhoun went into Column 1. When performance levels dropped, Fort Calhoun moved into Column 2, Column 3, Column 4, and Column 5 (Column 5 marking when the NRC placed Fort Calhoun under its Manual Chapter 0350 process for especially troubled reactors.)

On two occasions (3^rd quarter 2007 and 2^nd quarter 2008), the NRC returned Fort Calhoun to Column 2 from Column 3 after determining that safety performance had improved sufficiently.  In the 3^rd quarter 2008, the NRC returned Fort Calhoun to Column 1 and routine oversight activities.

What’s wrong with this picture?

As UCS documented in  No More Fukushimas; No More Fort Calhouns fact sheet, many of the safety problems had existed at the Fort Calhoun plant since 1996. Several dated back to the original construction of the plant in the late1960s and early 1970s. In other words, both the NRC’s baseline inspections (those applied to Fort Calhoun when it resided in Column 1) and its supplemental inspections (those applied when the plant was in Columns 2 and 3) failed to detect ALL of these safety problems.

The problems that kept Fort Calhoun shut down for 32 months were not introduced in 2009 and 2010 after the NRC returned Fort Calhoun to Column 1—they existed all along. Yet the NRC’s ROP missed them all. The ROP missed every single one of them, until after the first quarter of 2011. After that time, finding safety problems was like shooting fish in a barrel—NRC inspectors could hardly turn around without finding yet another safety problem that had to be fixed prior to restart.

So how could more cowbell improve nuclear plant safety?

Rather than expending so much time and effort ensuring that the barn door has been closed, safety would be better served by noticing that it’s open sooner. Cowbells should have sounded long before the first quarter of 2011.

UCS’s fact sheet documented many safety problems that existed at Fort Calhoun for years before the ROP’s inception in 2000. Two of the safety problems involved the emergency diesel generators (EDGs).

EDGs are among the most safety significant components at the plant. Consequently, they receive considerable oversight attention by the NRC. Yet that attention failed to identify either of these two problems that had existed since at least 1990.

And it was not just one miss or even two misses by one NRC inspector—it was a lot of misses by a lot of NRC inspectors over a lot of years. A search of ADAMS, the NRC’s online digital library, identified 39 inspections conducted at Fort Calhoun by the NRC between 2000 and 2010 inclusive that included some oversight of the EDGs.

Something is fundamentally wrong with safety inspections of highly safety significant components that fail to notice safety problems. Finding safety problems isn’t one of the reasons for conducting the safety inspections—it’s the only reason for doing them.

And yet many safety problems remained undetected until 2011 when it took an army of workers more than two years to correct them all.

Our Takeaway

Fort Calhoun is not an isolated case. It marked the 52^nd time that a U.S. reactor had to remain shut down longer than a year while safety problems were corrected. The majority of these year-plus outages involved a myriad of safety problems that had existed for months and sometimes years before being noticed.

Consider how safe Fort Calhoun really was on April 10, 2011, or during the preceding years when it operated despite a large and growing number of undetected, uncorrected safety problems. The NRC placed Fort Calhoun in Columns 1, 2, and 3 of the ROP’s Action Matrix. In reality, the presence of the same safety problems that put Fort Calhoun into Column 5 in the third quarter of 2011 should have had it there in the fourth quarter of 2000. The safety problems were there in 2000—it took the NRC another decade to notice them.

Fig. 2 (Source: UCS)

Safety and economics both scream out for the NRC to prevent the 53^rd time. As more and more pre-existing safety problems accumulate at an operating reactor, the path shortens for an initiating event to lead to nuclear disaster. Put another way, defense-in-depth works better when there are fewer and smaller holes in each protective barrier.

Likewise, finding and fixing problems sooner results in better financial performance. UCS estimated the cost of the 51 year-plus reactor outages before Fort Calhoun to be over $80 billion.

The NRC should construct timelines for each major safety problem corrected during the 2011-2013 outage at Fort Calhoun. The timelines should indicate when the safety problems were introduced and the subsequent NRC inspections that examined the associated system or component. Because the safety problems existed for long durations, many NRC inspection procedures will correlate to each safety problem. The NRC should then evaluate changes to the inspection procedures that increase the likelihood of detecting similar problems in the future. The NRC does not inspect everything; instead, the NRC audits samples. Conducting a Fort Calhoun retrospective would allow the NRC to adjust the number of items selected for each sample or revise the choice of items within the samples or change how it evaluates sample items so as to become more capable at finding safety problems.

The safety problems at Fort Calhoun were not invisible—they were easily found after the 1^st quarter of 2011. The NRC must figure out how to make them visible sooner. The NRC must detect safety problems sooner and ring cowbells as the barn doors are opening. 

More cowbell = better nuclear safety.

“Fission Stories” is a weekly feature by Dave Lochbaum. For more information on nuclear power safety, see the nuclear safety section of UCS’s website and our interactive map, the Nuclear Power Information Tracker.

Hey Palisades, Haven't Heard From You For A Long Time

AUTOMATIC REACTOR TRIP DUE TO TURBINE TRIP

"At 0117 [EDT] on 9/16/2015 a reactor trip occurred (4-hr non-emergency). The plant was at approximately 85% power performing a coastdown in preparation for a refueling outage when a Digital Electro-Hydraulic (DEH) alarm was received in the control room. Shortly following receipt of the alarm the turbine tripped. This resulted in an RPS actuation and a reactor trip on Loss of Load. The crew entered EOP-1 Standard Post Trip Actions and completed all required actions. The crew subsequently entered EOP-2 Reactor Trip Recovery.

"All full-length control rods inserted fully. Auxiliary Feedwater System actuated in response to low steam generator water levels (8-hr non-emergency). Steam generator water levels are in progress of being returned to normal operating levels. No known primary to secondary leakage. Atmospheric Steam Dump Valves lifted after the trip and subsequently reseated.

"The plant is currently stable in Mode 3 at NOP/NOT being maintained by the Turbine Bypass Valve.

"Initial investigation into the cause of the turbine trip appears to be from a DEH power supply failure.

"The NRC Resident Inspector was notified of the reactor trip at 0139 on 9/16/2015."

More Than Bad Seals At Indian Point Three

 "out of an abundance of caution,”

If this was true, why didn't they shutdown the very minute they confirmed inner seal leakage?

They probably stalled to get thru summer peak, and have new seals made up, and get W contractor to install them.

And the irony, and the prep and disassemble, and how long will it really take to take out the old seals and install the new one, 1 - 4 hrs.

Penny wise and pound foolish.

Watching the parade of scrams and shutdowns walk pass us from the Indian Point Facility.

It is highly abnormal to have reactor head seal problem...it worst than that when they both go.

Outage planned for Indian Point reactor after water leaks

By Colleen Wilson

September 15, 2015 No Comment

Indian Point Energy Center operators planned a shutdown of the Unit 3 nuclear reactor beginning Sept. 14 to replace lid seals that have been leaking for weeks.

The leakage occured between the inner and outer seals separating the lid and vessel of the reactor. The replacement is “out of an abundance of caution,” according to a statement from Entergy Corp., owners of the Buchanan-based power plant.

Jerry Nappi, a spokesman for Entergy, said the inner seal was observed to be leaking shortly after the unit’s planned refueling outage in March and the outer seal began leaking in July.

Operators responded by setting up a water collection line and a tank, Nappi said, adding plant workers “were able to closely track [the leak] and monitor it continuously from the control room.”

The water leak did not pose any threat to the health and safety of workers or the public and there was no radiation released, according to Entergy.

The work to replace the seals requires help from a specialty vendor and Entergy had to coordinate schedules to plan the outage. Nappi said the work is expected to take two weeks.

Amazing bad number!!! It's money.

This is the sixth time the Unit 3 reactor has shut down this year and, of those six, the second time the reactor went offline for planned maintenance. The four other times Unit 3 went offline were unexpected automatic or manual shutdowns for a pipe leak, transformer failure, electrical disturbance and fluctuating water levels.

Beating the hell out of the plant and safety systems...abusing them.

Recent series of Indian Point shutdowns worst in years Ernie Garcia, elgarcia@lohud.com12:08 p.m. EDT August 4, 2015

Besides a transformer failure that spilled oil into the Hudson River, this year's shutdowns were due to a steam leak, a pump motor failure and switch yard breaker failure

BUCHANAN — Four unplanned reactor shutdowns over a two-month period at Indian Point are the most setbacks the nuclear power plant has experienced in years.

A review of unplanned shutdowns from January 2012 to the present showed this year's events happened within a short time frame, between May 7 and July 8, in contrast with events from other years that were more spread out, according to data released by Indian Point.

So many mishaps at the Entergy-owned plant haven't occurred since 2009, when one of two units at the Buchanan site experienced a similar series, said plant spokesman Jerry Nappi.

Besides a May 9 transformer failure that spilled some 3,000 gallons of oil into the Hudson River, this year's shutdowns were prompted by a May 7 steam leak, a July 8 pump motor failure and a June 15 switch yard breaker failure offsite in a Consolidated Edison substation.

If a nuclear plant has more than three unplanned shutdowns in a nine-month period, its performance indicator could be changed by the federal Nuclear Regulatory Commission, which results in additional oversight. That's what happened with Entergy's Pilgrim Nuclear Power Station in Plymouth, Mass., after four unplanned shutdowns in 2013.

So far, Entergy said there doesn't appear to be a pattern to the Indian Point shutdowns.

"You do want to look at these events holistically to see if there is something in common, but you also look individually to see what the causes were," Nappi said. "A plant shutdown in and of itself is not a safety issue."

One of the four recent Buchanan shutdowns triggered a special inspection by the NRC and calls to close the nuclear plant by environmental groups and elected officials. Gov. Andrew Cuomo has said in the past Indian Point should close, but his office did not respond to a request for comment about whether the recent shutdowns have prompted any state scrutiny.

The NRC is expected to release a quarterly report on Indian…

NEISO Grid Electric Price

$0.02 megawatt-hour @ 9:45 pm

President Burnie Sanders and Nuclear Power.

When is the media going to ask Burnie Sanders about his Vermont Yankee adventures? The leaking radioactive pipes and VY’s loss of credibility was one of the largest media stories in Vermont in decades. Burnie was heavy involved in shutting down Vermont Yankee and the nukes hate him. He was always at the local NRC meetings and the large scale demonstrations.

 

What would a Sanders president mean for the nuclear industry?

I'll bet you to the one, except Gov Shumlin, most politicians hate the nuclear issue coopting their wider campaign issues. Gov Shumlin used the anti VY issue to his advantage and that is how he got to served three terms as governor.

MY blog: “Pilgrim Nuclear Plant Following Exactly the Vermont Yankee Model?”

Just do a google search on Vermont Yankee and Burnie Sanders?

No doubt where he stands on nuclear power?

Bernie Sanders Urges Obama to Declare a Nuclear Moratorium

In the aftermath of the nuclear disaster in Japan, U.S. Sen. Bernie Sanders urged the White House to form a presidential commission on nuclear safety in the United States as part of a five-point crisis response.

In a letter to President Barack Obama, Sanders (I-Vt.) also asked for a moratorium on license renewals by the Nuclear Regulatory Commission. He said the White House should withdraw a request for $36 billion to bankroll building new nuclear plants. He questioned why taxpayers - not nuclear plant owners - are on the hook for damages in the event of a meltdown or other accident at a private power plant. And he said states should get more say on plant safety.

Sanders serves on the Senate committee that oversees the NRC, the federal agency that regulates commercial nuclear reactors in this country….

Statement: Sanders on Vermont Yankee

Thursday, January 14, 2010

BURLINGTON, January 14 – Sen. Bernie Sanders (I-Vt.) today issued the following statement in response to reports that Vermont Yankee nuclear power plant officials provided misleading information to state lawmakers about the risk of radioactive leaks from underground piping at the plant.

“It is alarming to me that officials from Vermont Yankee now admit that they misled state legislators about the risk of radioactive leaks from the plant, and that we now know that there are elevated levels of radioactive materials leaking at the plant site.

“This is a very serious situation. I will ask the Nuclear Regulatory Commission to conduct a full investigation, and to take appropriate action.”

Friday, January 15, 2010

Vermont Congressional Delegation Requests NRC Investigation Into Radioactive Leak Risks At Vermont Yankee

WASHINGTON, January 15 - Vermont's Congressional Delegation - Sen. Patrick Leahy (D), Sen. Bernie Sanders (I), and Rep. Peter Welch (D) - Friday asked the Nuclear Regulatory Commission to conduct a thorough investigation into whether there was any attempt by Vermont Yankee officials to mislead state officials regarding the plant's safety and underground piping.  Press reports have suggested that officials from Entergy, which owns the Vermont Yankee nuclear power plant, may have provided inaccurate information to investigators about the risk of leaks at the facility.  The text of the delegation's letter to Nuclear Regulatory Commission Chairman Gregory B. Jaczko is below: 

January 15, 2010

The Honorable Gregory B. Jaczko
Chairman
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Dear Chairman Jaczko:

We are writing in response to the alarming news that Entergy, owner of the Vermont Yankee nuclear power plant, may have misled state officials regarding the safety of the plant.

Elevated levels of radioactive tritium have recently been found in a groundwater monitoring well on the plant site and we understand from recent news reports that Entergy has confirmed that underground piping is among the possible sources of the contamination.  According to various reports in the media, this comes after Entergy Vermont Yankee officials had told state investigators, on a number of occasions, that there was no underground piping carrying water that could contain radioactivity.   This leak highlights our ongoing concerns about Entergy Vermont Yankee's commitment to safety and to being forthright with the public and state and federal regulatory and safety agencies.

We therefore request that you undertake an immediate and thorough investigation to determine if there was an attempt by Entergy Vermont Yankee to mislead state officials regarding the plant's safety and underground piping.  Please also determine whether information provided by Entergy to the NRC has been accurate, complete, and consistent with that provided to the State of Vermont.  We hope you can pinpoint exactly what Entergy knew about the extent of their underground piping and this leak, and when they knew it.  We would also like to know whether and why state regulatory agencies were not made aware of the extent of underground piping and the risk it posed prior to this incident, and whether communications to the NRC have been complete and timely.  Finally, we would like the NRC to continue to work with the plant to determine the cause of the leak and resolve the situation as quickly as possible to avoid any further release of radioactive materials.

Please continue to keep us thoroughly informed as more information becomes available.  We are committed to assisting Vermont and the NRC to ensure that the Entergy Vermont Yankee plant meets its safety obligations.  We appreciate your timely attention to this issue. 

Sincerely,

Entergy Maliciously Contesting minor NRC Violations at ANO

Look at how much time is being eaten up over this non cited violation of the Fuel Oil piping by the mindlessly complex codes and regulation?  This is how bureaucratic war looks like to Entergy and the NRC. The interpretation of rules and codes as ruthless weapons? You see the excessive rules favors Entergy.

Entergy is sending the message to the NRC, you better be hyper vigilant on the rules and codes future violations...the bureaucracy that nobody can understand. So Entergy is trying to consume limited NRC resources on a non cited violation...the NRC is going to be excessively careful in the future concerning small violations. They are going to have to triple vitrify all the rules and codes. They are going to be buried in the excessive complex bureaucratic rules while not seeing the big violations.

They is how a war looks like between a shameless licensee and a neutered NRC...

0CAN091501

September 3, 2015

U.S. Nuclear Regulatory Commission

ATTN: Document Control Desk

11555 Rockville Pike

Rockville, MD 20852

SUBJECT: Response to Non-cited Violation in NRC Integrated Inspection

Report 05000313/2015002 and 05000368/2015002

Arkansas Nuclear One – Units 1 and 2

Docket Nos. 50-313 and 50-368

License Nos. DPR-51 and NPF-6

REFERENCE: NRC letter to Entergy, Arkansas Nuclear One – NRC Inspection Report 05000313/2015002 and 05000368/2015002, dated August 5, 2015

(0CNA081501) (ML15218A371)

 

Reference 1 provided the results of the Arkansas Nuclear One (ANO) integrated inspection for the second quarter of 2015. Per 10 CFR 50.4 and in accordance with the guidance in the

Enforcement Policy, Entergy Operations Inc. (Entergy) hereby contests one of the non-cited violations (NCVs) identified in the report.

A green NCV of 10 CFR Part 50, Appendix B, Criterion XI, “Test Control,” was identified in the report for failure to establish and maintain an adequate testing program for the fuel oil transfer piping for ANO, Units 1 (ANO-1) and 2 (ANO-2). Specifically, the licensee did not establish inservice inspection (ISI) requirements to detect degradation of the fuel oil piping, above ground and buried, between the fuel oil storage tanks and the emergency diesel generator (EDG) day tanks.

ANO-2 Conclusion

The ANO-2 ASME Section XI Inservice Inspection (ISI) program correctly excludes the diesel fuel oil piping based on the requirements of 10 CFR 50.55a and the ANO-2 licensing basis. There are no requirements of ASME Section XI or 10 CFR 50.55a that require plants that received a construction permit after January 1, 1971, to include piping in the Section XI boundaries not required by regulation or the licensing basis to be designed and constructed to ASME Section III, Class 1, 2, or 3 requirements. For ANO-2 Entergy’s implementation of Safety

Guide 26 (RG 1.26) and the classification of the fuel oil piping as ANSI B31.1 were described in the licensing basis and accepted by the NRC. Also, the position taken by the NRC in the subject NCV appears inconsistent with the regulatory guidance provided in NUREG 1482, Revision 2.