Why Are Nuclear Plants Having So Many Safety Related HVAC Problem?

July 23

Is excessive HVAC NRC reporting a symptom of too large plant work back log???

July 22:

Bet you most of the HVAC systems were rushed add on systems during troublesome construction and post TMI...

I think the NRC and industry would come back and say...this is where prescriptive regulations got it wrong. There is a insignificant chance these HVAC problems would lead to core damage and a offsite release.

I would say this is where this perspective incentivizes broadly setting up and tolerating safety component degradation and obsolesces.

If we broadly tolerate component degradations in these amazingly complex machines and organizations...then we become much closer to the day when one of these plants runs away from us.

Is there such a thing as boiling frog with priorities metaphor...normalization of deviance?

Is risk perspective numbing us to threats and real risk...
 

The truth is, if the NRC demanded just a few of this plants to shut down over HVAC problems, the industry would update all their problematic HVAC systems. It is common for these guys to fix these problems in a isolated way…instead of thinking holistic it is more efficient with our resources just to buy a new car or buy a new HAVC system.   

The question is if this is an essential safety system, then why is there only one and doing maintenance on it makes it inop.

If this is essential for accident mitigation, why is it taken off line at 100% full power operation, unless the plant has carefully schedule a DBA not to occur while the essential HVAC system is down for repair?"

 July 21:

All 2015 events.

Is it my VY leaking roofs deal…didn’t realize the system had come to end-of-life.

I bet you HVAC problems would be at the bottom of the barrel of maintenance and budgets priority systems.

What do you make big picture and why this is happening...are the issues trending up?

Would having four HVAC systems powered from two independent buses fix it...

Not a complete list: 

###Power Reactor    Event Number: 51103
Facility: CATAWBA
Region: 2 State: SC
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: THOMAS GARRISON
HQ OPS Officer: JOHN SHOEMAKER    Notification Date: 05/29/2015
Notification Time: 20:17 [ET]
Event Date: 05/29/2015
Event Time: 12:30 [EDT]
Last Update Date: 06/12/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
KATHLEEN O'DONOHUE (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    Y    100    Power Operation    100    Power Operation
Event Text
TECHNICAL SUPPORT CENTER VENTILATION SYSTEM OUT OF SERVICE DUE TO DISCOVERED CONDITION

"This is non-emergency eight hour notification for a loss of Emergency Assessment Capability.

"This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as the discovered condition affects the functionality of an emergency response facility.

"A condition impacting functionality of the TSC [Technical Support Center] Ventilation system was discovered on 05/29/2015 at 1230 [EDT]. The issue involves a loss of cooling capability of the TSC ventilation system due to a failed relay. Maintenance will begin repairs at 0700 [EDT] on 05/30/2015. Estimated time to repair is unknown at this time.

"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Coordinator will relocate the TSC staff to an alternate location in accordance with applicable site procedures. The Emergency Response Organization team will be notified of the condition and the possible need to relocate during an emergency. This condition does not affect the health and safety of the public or station employees. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector will be notified."

* * * UPDATE FROM AARON MICHALSKI TO DANIEL MILLS AT 1557 EDT ON 6/12/15 * *

The TSC ventilation system has been returned to service. The licensee will notify the NRC Resident Inspector.

Notified R2DO (Guthrie).

###Power Reactor    Event Number: 51154
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BLAKE BAXTER
HQ OPS Officer: DANIEL MILLS    Notification Date: 06/15/2015
Notification Time: 17:42 [ET]
Event Date: 06/16/2015
Event Time: 07:00 [CDT]
Last Update Date: 06/15/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
PATTY PELKE (R3DO)

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    Y    100    Power Operation    100    Power Operation
Event Text
TSC VENTILATION TO BE REMOVED FROM SERVICE FOR PLANNED MAINTENANCE

"On 6/16/2015, planned preventive maintenance activities [will be] performed on the Braidwood Generating Station Technical Support Center (TSC), Ventilation System. The work will be completed within approximately 48 hours. This activity includes preventative maintenance that requires the TSC ventilation system to be out of service which will render the TSC ventilation system non-functional.

"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary.

"This event is reportable per 10 CFR 50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit or ventilation system to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency.

"The licensee has notified the NRC Resident Inspector."

###Power Reactor    Event Number: 51164
Facility: LASALLE
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: TODD CASAGRANDE
HQ OPS Officer: JEFF HERRERA    Notification Date: 06/17/2015
Notification Time: 23:39 [ET]
Event Date: 06/17/2015
Event Time: 18:41 [CDT]
Last Update Date: 06/18/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
PATTY PELKE (R3DO)

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    Y    100    Power Operation    100    Power Operation
Event Text
TECHNICAL SUPPORT CENTER VENTILATION SYSTEM RETURN DAMPER FAILED CLOSED

"On June 17th, 2015 at 1841 CDT, it was determined that the onsite Technical Support Center (TSC) Ventilation System return damper 0VS119Y was failed closed, the failed closed damper affects the TSC Emergency Makeup Train filtration efficiency. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO [Emergency Response Organization] activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures.

"The licensee has notified the [NRC] Senior Resident Inspector of the issue."

* * * UPDATE AT 1700 EDT ON 06/18/15 FROM TODD CASAGRANDE TO S. SANDIN * * *

"After repairs were completed, the TSC Ventilation was restored to service at 1650 EDT on 06/18/2015.

"The licensee has notified the NRC Resident Inspector."

Notified R3DO (Pelke).

###Power Reactor    Event Number: 51154
Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BLAKE BAXTER
HQ OPS Officer: DANIEL MILLS    Notification Date: 06/15/2015
Notification Time: 17:42 [ET]
Event Date: 06/15/2015
Event Time: 07:00 [CDT]
Last Update Date: 06/18/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
PATTY PELKE (R3DO)

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    Y    100    Power Operation    100    Power Operation
Event Text
TSC VENTILATION TO BE REMOVED FROM SERVICE FOR PLANNED MAINTENANCE

"On 6/16/2015, planned preventive maintenance activities [will be] performed on the Braidwood Generating Station Technical Support Center (TSC), Ventilation System. The work will be completed within approximately 48 hours. This activity includes preventative maintenance that requires the TSC ventilation system to be out of service which will render the TSC ventilation system non-functional.

"If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary.

"This event is reportable per 10 CFR 50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit or ventilation system to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency.

"The licensee has notified the NRC Resident Inspector."

* * * UPDATE AT 1102 EDT ON 6/18/15 FROM DAVID KORTGE TO JEFF HERRERA * * *

"Braidwood Generating Station TSC ventilation was restored to available status at 0700 CDT on June 18, 2015.

"The previously reported system preventative maintenance has been completed."

The licensee notified the NRC Resident Inspector.

Notified the R3DO (Pelke).

###Power Reactor    Event Number: 51212
Facility: HARRIS
Region: 2 State: NC
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP
NRC Notified By: RAYMOND MOORE
HQ OPS Officer: MARK ABRAMOVITZ    Notification Date: 07/08/2015
Notification Time: 17:43 [ET]
Event Date: 07/07/2015
Event Time: 11:05 [EDT]
Last Update Date: 07/08/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
GERALD MCCOY (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
Event Text
TECHNICAL SUPPORT CENTER VENTILATION OUT OF SERVICE

"This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as the discovered condition affects the functionality of an emergency response facility.

"A condition impacting functionality of the TSC Ventilation system was discovered on July 7, 2015 at 11:05 EDT. The issue involved a loss of cooling capability of the TSC Ventilation system due to failed ventilation system components. Maintenance started repairs immediately following the discovery of the component failures and completed repairs to restore functionality of the TSC Ventilation system on July 8, 2015 at 17:07 EDT. On July 8, 2015, at approximately 15:30 EDT, further review of the impact of this equipment failure determined that this condition was reportable as a loss of emergency assessment capability.

"If an emergency were declared requiring TSC activation during the non-functional period, the TSC would have been staffed and activated using existing emergency planning procedures unless the TSC became uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC became necessary, the Emergency Director would have relocated the TSC staff to an alternate location in accordance with applicable site procedures. The Emergency Response Organization team was notified of the maintenance and the possible need to relocate during an emergency. This condition did not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified."

###Power Reactor    Event Number: 51213
Facility: LASALLE
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: BRADLEY BRUMUND
HQ OPS Officer: VINCE KLCO    Notification Date: 07/08/2015
Notification Time: 21:53 [ET]
Event Date: 07/08/2015
Event Time: 18:37 [CDT]
Last Update Date: 07/08/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
ANN MARIE STONE (R3DO)

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    Y    100    Power Operation    100    Power Operation
Event Text
TECHNICAL SUPPORT CENTER VENTILATION OUT OF SERVICE

"This telephone notification is provided in accordance with Exelon Reportability manual SAF 1.10, 'Major Loss of Emergency Preparedness Capabilities', and 10CFR50.72(b)(3)(xiii).

"On July 8th 2015 at 1837 [CDT], it was determined that the onsite Technical Support Center (TSC) Ventilation System Supply Fan belts had failed, resulting in loss of ventilation for the facility. Repairs were not completed within the time required had the TSC needed to be staffed. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO [Emergency Response Organization] activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures.

"The licensee has notified the [NRC] Senior Resident Inspector of the issue."

###Power Reactor    Event Number: 51213
Facility: LASALLE
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] GE-5,[2] GE-5
NRC Notified By: BRADLEY BRUMUND
HQ OPS Officer: VINCE KLCO    Notification Date: 07/08/2015
Notification Time: 21:53 [ET]
Event Date: 07/08/2015
Event Time: 18:37 [CDT]
Last Update Date: 07/11/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization):
ANN MARIE STONE (R3DO)

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    Y    100    Power Operation    100    Power Operation
Event Text
TECHNICAL SUPPORT CENTER VENTILATION OUT OF SERVICE

"This telephone notification is provided in accordance with Exelon Reportability manual SAF 1.10, 'Major Loss of Emergency Preparedness Capabilities', and 10CFR50.72(b)(3)(xiii).

"On July 8th 2015 at 1837 [CDT], it was determined that the onsite Technical Support Center (TSC) Ventilation System Supply Fan belts had failed, resulting in loss of ventilation for the facility. Repairs were not completed within the time required had the TSC needed to be staffed. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO [Emergency Response Organization] activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures.

"The licensee has notified the [NRC] Senior Resident Inspector of the issue."

* * * UPDATE FROM TODD CASAGRANDE TO DANIEL MILLS AT 1510 EDT ON 7/11/15 * * *

"After repairs were completed, the TSC Ventilation was restarted on 7/9/15 at 0625 EDT for a maintenance run, the TSC Ventilation was restored to operable status at 1500 EDT on 07/11/2015.

"The licensee has notified the NRC Resident Inspector."

Notified R3DO (Stone).

###Power Reactor    Event Number: 51232
Facility: DRESDEN
Region: 3 State: IL
Unit: [ ] [2] [3]
RX Type: [1] GE-1,[2] GE-3,[3] GE-3
NRC Notified By: PATRICK HAARHOSS
HQ OPS Officer: VINCE KLCO    Notification Date: 07/15/2015
Notification Time: 01:04 [ET]
Event Date: 07/15/2015
Event Time: 00:04 [CDT]
Last Update Date: 07/17/2015
Emergency Class: NON EMERGENCY
10 CFR Section: 
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
Person (Organization): 
STEVE ORTH (R3DO)

Unit    SCRAM Code    RX CRIT    Initial PWR    Initial RX Mode    Current PWR    Current RX Mode
2    N    Y    100    Power Operation    100    Power Operation
3    N    Y    100    Power Operation    100    Power Operation
Event Text
TECHNICAL SUPPORT CENTER OUT OF SERVICE DUE TO PLANNED MAINTENANCE 

"At 0004 [CDT] on Wednesday, July 15, 2015, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) emergency ventilation system will be removed from service for planned maintenance activities. During the maintenance, the TSC Ventilation will be shut down. The TSC air filtration fan and dampers will be non-functional, rendering the TSC HVAC accident mode non-functional. This maintenance is scheduled to minimize out of service time. The planned TSC ventilation outage is scheduled to be completed in approximately 24 hours. 

"Contingency plans are in place so that if an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning (EP) procedures and checklists. If radiological or environmental conditions require TSC facility evacuation during ventilation system restoration; the Station Emergency Director will relocate the TSC staff in accordance with station procedures." 

"The NRC Resident Inspector has been notified." 

* * * UPDATE FROM TRAVIS PRELLWITZ TO DONALD NORWOOD AT 1733 EDT ON 7/17/2015 * * * 

"At 1347 CDT on July 17, 2015, Dresden TSC Ventilation was restored. The Dresden TSC Ventilation is Functional at this time. 

"The NRC Resident Inspector has been notified." 

Notified R3DO (Orth).

Safety Valves In Our Nuclear Power Plants Going Wild

Licensee Event Report # 2015-002-01, "Technical Specification Prohibited Condition Caused by Four Main Steam Safety Valves Outside Their As- Found Lift Set Point Test Acceptance Criteria"

Indian Point Unit No. 3

Docket No. 50-286
 

Reference: 1. LER-2015-002-00 submitted by letter NL-15-065 dated April 27, 2015

An extent of condition (EOC) was performed to determine where potential conditions with similar valves and environments could occur. The review determined that EOC round in the failure of MSSV MS-45-4, MS-46-2 and MS-47-4 is restricted to the other 17 MSSVs at unit 3 and the 20 MSSVs at unit 2 due to the valve design. All MSSVs are exposed to high vibrations during their operating cycle during which wear can occur. Previous failures of MSSVs have included wear due to spring skewing and set point…

Why is Entergy having so many problems with Safety Relief Valves and now the Main Steam Safety Valves? The MSSV provide overpressure protection in the main steam lines. 

If I hear fretting or normal main steam line vibrations damaging these valve again I am going to vomit.

Why can’t the engineers design this valves for the duty and conditions of the plant? They just sit there doing nothing for 99.99% of the time.
 

For only three years below, this is shocking. So how have the failure changed in the last decade... is less testing and maintenance behind this.

A review was performed of Licensee Event Reports (LERs) for the past three years for any events reporting TS prohibited conditions due to multiple valve test failures. LER-2011-004 reported two MSSV's outside their as-found lift set point acceptance criteria due to spindle wear and spring skew. LER-2013-001 reported two MSSVs (MS-46-3 and MS-48-3) outside their as-found set point acceptance criteria. The cause of MS-46-3 failure was galling around the circumference of the spindle rod as a result of vibration. MS-48-3 evidenced similar fretting on one side of the spindle consistent with what was found on valve MS-46-3. Failure cause was determined to be due to internal friction caused by foreign material between the guide bearing and spindle. The causes for the previous events reported in LER-2011-004 are similar to this event.

Recent issues with fretting and normal vibration issues with SRVs and MNSSV include the following plants:

Hatch

Oyster Creek (yellow finding)

Dresden
 

Quad Cities
 

Pilgrim

These valves are failing because the vendor can’t control the component dimensional or material conditions in the valves. These are equivalent to the valves that gave us TMI.
  

I am amazed pre-plant testing haven’t picked up the weaknesses of these valves. Pre-plant testing either can't pick up the defect or the testing damages the fragile valves:) 

I am shocked the manufacturers don’t provide completely durable and bullet proof components and valve to these nuclear power plants? Just to be clear, the MSSV come from PWRs and the SRVs comes from BWRs.

You get it, Indian Point don’t trust the safety of their valves, don’t understand the degradation mechanism, talking about “another” valve design…they are testing these valves twice as frequently now because they don't understand the degradation mechanisms.

LER 2011-004-00 plus two other LERs in 10 years...

During the Preventive Maintenance (PM) of both valves, run-out and wear along the radius of the spindles were noted. In a high flow system, the result would be increased wear along the spindle in the form of steps which were found with MS-47-4 and MS-48-4.

It is a runaway train, in defective component these identical problems happen over and over without fixing the problem. These plants are great at churning paperwork...poor a fixing problems so they never show up again.

LER 2009-002-000: Over and over again, half ass fixes for at least 6 years...the runaway degradation occurring unabated at least since 2009. 

Spindle problems, who cares indeterminate and this seems to be the beginning of this problem. What changed before 2009?   

Cause of Event

The apparent cause of the two MSSVs lifting greater than 3% of their nominal setpoint is indeterminate but most likely caused by setpoint drift. MS-45-1 and MS-48-3 were disassembled and inspected and identified to have some scoring on their valve spindles. Assessment with Original Equipment Manufacturer (OEM) could not directly relate the indications discovered on the valves' spindles to the As-Found test results

   

Fort Calhoun Shutdown Again

Sounds like a reactor coolant pump seal...

Three Stage Safety Valves No Longer In Hatch or Pilgrim

Three stage Safety Valves No Longer in Hatch or Pilgrim

As I said, I call the three stage INOP because there was an active defect in the valve and there was no understanding with their failure mechanism. This is called being conservative.   

* * * UPDATE FROM JOHN DeBONIS (VIA EMAIL) TO STEVEN VITTO AT 1256 EDT ON 6/30/15 ***

Curtiss-Wright provided an update to state their root cause analysis findings and corrective actions. Corrective actions are estimated to be completed within 12 months.

"The following plants were supplied 0867F MS-SRVs:

Pilgrim (Model 09J-001) Quantity Shipped = 8

FitzPatrick (Model 09H-001) Quantity Shipped = 4, Quantity on order= 8

Hatch 1 and 2 (Model 09G-001) Quantity Shipped = 24, Quantity on order= 12

"The following plants will be supplied 0867F MS-SRVs:

Hope Creek (Models 14J-001, 14J-002) Quantity on order = 7

"Valves Currently Installed

 

"Target Rock recommends valves currently installed be inspected to ensure the main piston shoulder has contact with the main disc stem shoulder. These inspections should be scheduled based on plant-specific indications of the potential for fretting. These inspections can be performed by removing the base assembly from the main body and physically measuring for shoulder-to-shoulder contact.

"Should you have any questions regarding this matter, please contact Michael Cinque, Director of Program Management at (631 ) 293-3800."

Notified NRR Part 21 Group (via email), R1 DO (Dimitriadis), and R2DO (Suggs).

Junk Hatch plant Target Rock Three Stage Safety Valves

Three take homes.

1) they changed out the 2 stage for three stage...the problems only got worst.

• 2011: LER 2-2011-002, identified multiple SRV setpoint drift for 8 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs during the Unit 2 Spring 2011 refueling outage which was considered at that time to be the long term fix for this corrosion bonding issue.

It sounds like they yanked out the three stage with this below eight failure. So why didn't Pilgrim take out their reliefs early like Hatch???

• 2012 with three stage: LER 1-2012-004, identified multiple SRV setpoint drift for 8 of the 11 SRVs

• 2013:  LER 1-2014-003, identified multiple SRV setpoint drifts for 5 of the 11 two-stage SRVs installed on Unit 1.

• LER 1-2014-003, identified multiple SRV setpoint drifts for 5 of the 11 two-stage SRVs installed on Unit 1. The two-stage SRVs with platinum-coated pilot discs were

• 2015: 2 of 11

2) We have no idea of the magnitude of trhe leakage.

3) Hatch has problems with steam line vibration damaging these valves. 

4) We know this state is totally inaccurate: "3-stage SRVs typically do not exhibit set point drift and the modified pilot reduces instances of vibration induced spurious openings and leak-by." 

5) I don't think you ever can trust the public disclosure dance between a vendor and a licensee.  

I believe for each of these outside tech spec lift inaccuracies, if the plant would have known it...they would have been required to shutdown.
  
JUL 10 2015

Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec 

Surveillance Criteria

On May 11, 2015 at approximately 0923, Unit 2 was at 1 00 percent rated thermal power (RTP) when the "as found" testing results of the 2-stage main steam safety relief valves (SRVs) were received which indicated that two of eleven of the Unit 2 SRVs had experienced a setpoint drift during the previous operating cycle which resulted in their failure to meet the Technical Specification (TS) opening setpoint of 1150 +1- 34.5 psig percent as required by TS Surveillance Requirement (SR) 3.4.3.1.

The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and seating surfaces. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Hatch and within the BWR industry. All 2-stage SRVs with platinum coated pilot seats were removed from Unit 2 during the 2015 refueling outage and replaced with 3-stage SRVs with amodified pilot. 3-stage SRVs typically do not exhibit set point drift and the modified pilot reduces instances of vibration induced spurious openings and leak-by.

A 3-stage SRV with a similar modified pilot was installed on Unit 2 during the ·2013 refueling outage. Based upon "as-found" testing results, it was seen that pressure lift setpoints were maintained during plant operation.

DESCRIPTION OF EVENT

On May 11 2015, at approximately 0923, with Unit 2 at 100 percent rated thermal power (RTP), "as-found" testing of the 2-stage main steam safety relief valves (SRVs) (EllS Code RV) showed that two of the ten main steam SRVs that were tested had experienced a drift in pressure lift setpoint during the previous operating cycle such that the allowable technical specification {TS) surveillance requirement (SA) 3.4.3.1 limit of 1150 +1- 34.5 (+/- 3%) psig had been exceeded. Below is a table illustrating the as found testing results of Unit 2 SRVs that were removed from service during the Spring 2015 refueling outage and replaced with 3-stage SRVs

.

MPL Pilot Serial No. Lift Pressure Percent Drift

2B21-F013B 1006 1155 0.40%

2B21-F013C 1231 1172 1.90%

2B21-F013D 303 1184 3.00%

2B21-F013E 315 1210 5.20%

2B21-F013F 1189 1179 2.50%

2B21-F013G 302 1174 2.10%

2B21-F013H 1230 1190 3.50%

2B21-F013K 1229 1164 1.20%

2B21-F013L 1228 1163 1.10%

2B21-F013M 1008 1179 2.50%

The 2-stage SRVs that were installed on Unit 2 during the previous cycle (Cycle 23) utilized platinum coated pilot discs. The 3-stage SRVs currently installed on Unit 2 have a modified pilot that helps reduce the possibility of inadvertent lift and leak by due to system vibration. The one 3-stage SRV that was installed on Unit 2 during Cycle 23 was recently successfully tested and found to be within the allowable TS SA pressure lift setpoint limit of 1150 +1- 34.5 (+/- 3%) psig.

CAUSE OF EVENT 

The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and its seating surface. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Plant Hatch and in the industry. In General Electric (GE) Service Information Letter (SIL) 196, Supplement 16, GE determined that condensation of steam in the pilot chamber of Target Rock 2-stage SRVs can cause oxygen and hydrogen dissolved in the steam to accumulate. As steam condenses in the relatively stagnant pilot chamber, the dissolved gases are released. In a volume such as the pilot chamber which is normally at approximately a 1000 psig pressure and a temperature of 545 degrees F, the total pressure consists primarily of water vapor partial pressure because 544.6 degrees F is the saturation temperature at 1000 psi g. This wet, hot, high-oxygen atmosphere can be very corrosive and can increase the likelihood of corrosion-induced bonding of the pilot disk to its seat. It was also noted that proper insulation minimizes the accumulation rate of non-condensable gases and the steady-state oxygen partial pressure. Despite improvements made in maintaining the integrity of insulation for the previously installed 2-stage SRVs and installing new platinum coated pilots, the corrosion-induced bonding continued to occur as evidenced by the test results from this most recent outage.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT 

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) because a condition occurred that is prohibited by TS Surveillance Requirement (SR) 3.4.3.1. Specifically, an example of multiple test failures is given in NUREG-1022, Revision 3, "Event Reporting Guidelines 10 CFR 50.72 and 50.73" which describes the sequential testing of safety valves. This example notes that "Sometimes multiple valves are found to lift with set points outside of technical specification limits."

NUREG-1022 further states in the example that "discrepancies found in TS surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on a review of relevant information (e.g., the equipment history and the cause of failure), to indicate that the discrepancy occurred earlier. However, the existence of similar discrepancies in multiple valves is an indication that the discrepancies may well have arisen over a period of time and the failure mode should be evaluated to make this determination." Based on this guidance and the fact that the development of the corrosion occurred over a period of time of plant operation, the determination was made that this "as found" condition is reportable under the reporting requirements of 10 CFR 50.73(a)(2)(i)(B). There are eleven SRVs located on the four main steam lines within the drywell in between the reactor pressure vessel (RPV) (EllS Code RPV) and the inboard main steam isolation valves (MSIVs) (EllS Code ISV). These SRVs are required to be operable during Modes 1, 2, and 3 to limit the peak pressure in the nuclear system such that it will not exceed the applicable ASME Boiler and Pressure Vessel Code Limits for the reactor coolant pressure boundary. The SRVs are tested in accordance with TS surveillance requirement 3.4.3.1 in which the valves are tested as directed by the In-Service Testing Program to verify lift set points are within their specified limits to confirm they would perform their required safety function of overpressure protection. The SRVs must accommodate the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code Limit of 1375 psig peak vessel pressure, has been defined by an event involving the closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches with the reactor ultimately shutting down as the result of a high neutron flux trip (a scenario designated as MSIVF).

The results from this MSIVF event analysis was performed by the Nuclear Fuels Department in order to bound the "as-found" results of the U2 Cycle 21 2-stage SRVs pressure setpoint drift. The results from this analysis showed a small increase in peak pressures relative to the Hatch-2 Cycle 21 reload licensing analysis (ALA) results. The higher peak pressures were due to the fact that eight of the eleven SRVs opened at pressures higher than that which was assumed in the ALA. It should be noted that in this analysis, the larger actual valve bore size was used in the calculations for nine of the valves rather than the smaller bore size which was conservatively assumed in the ALA. Therefore, higher steam flow capacities than those assumed in the ALA were used in this analysis for those nine valves.

Based on the analysis, the calculated minimum margin to the 1375 psig ASME Boiler and Pressure Vessel Code overpressure limit for peak vessel pressure would have been 27.7 psig and the minimum margin to the 1325 psig Tech Spec Safety Limit for the reactor steam dome pressure would have been 2.9 psig during an MSIVF event during Cycle 21 operation. Therefore, these test results show that in this case, where two of the eleven SRVs would have opened at pressures higher than that which was assumed in the RLA, the peak pressure at the bottom of the vessel would have remained below the ASME Boiler and Pressure Vessel code limit and the peak RPV dome pressure remained within the TS Safety limits.

Additionally, a highly reliable, though non-credited, electrical actuation system serves as a redundant, independent method to actuate the SRVs. During Cycle 23 this redundant electrical logic system was fully functional. Based on the analyses performed, the overpressure protection system would have continued to perform its required safety function if called upon in its "as found" condition. Therefore, this event had no adverse impact on nuclear safety and was of very low safety significance.

CORRECTIVE ACTIONS 

The 2-stage SRVs with platinum-coated pilot discs were removed from Unit 2 during the 2015 refueling outage and replaced with 3-stage SRVs that have a modified pilot. 3-stage SRVs typically do not exhibit set point drift due to their design. The modified pilots will help reduce spurious openings and leak-by due to system vibration.

ADDITIONAL INFORMATION

Other Systems Affected: None

Failed Components Information:

Master Parts List Number: 2B21-F013E, H

Manufacturer: Target Rock

Model Number: 7567F

Type: Relief Valve

Manufacturer Code: T020

EllS System Code: SB

Reportable to EPIX: Yes

Root Cause Code: B

EllS Component Code: RV

Commitment Information: This report does not create any licensing commitments.

PREVIOUS SIMILAR EVENTS:

LER 1-2014-003, identified multiple SRV setpoint drifts for 5 of the 11 two-stage SRVs installed on Unit 1. The two-stage SRVs with platinum-coated pilot discs were removed from Unit 1 during the 2014 refueling outage and replaced with 3-stage SRVs that have a modified pilot. The modified pilots will help reduce spurious openings and leak-by due to system vibration.

LER 1-2012-004, identified multiple SRV setpoint drift for 8 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 2-stage SRVs whose pilot discs had undergone a platinum surface treatment which was considered at that time to be the long term fix for this corrosion bonding issue.

LER 2-2011-002, identified multiple SRV setpoint drift for 8 of the 11 SRVs. Corrective actions included replacement of the 2-stage SRVs with 3-stage SRVs during the Unit 2 Spring 2011 refueling outage which was considered at that time to be the long term fix for this corrosion bonding issue. Subsequent to that outage the 3-stage SRVs exhibited signs of unacceptable leakage which resulted in two separate outages that involved changing out four SRVs during the first outage and the remaining seven SRVs during the subsequent outage in May 2012. The 3-stage SRVs were replaced with 2-stage SRVs containing pilot discs that had undergone the platinum surface treatment.

LER 1-2010-001, identified multiple SRV setpoint drift for 5 of the 11 SRVs. Corrective actions included refurbishment of the pilot valves and included the replacement of the pilot discs with discs made from corrosion-induced bonding. These were the same actions that were taken following similar failures reported in LEA 2-2009-001, since improved results had been seen to some degree in the industry for at least one operating cycle when these actions were implemented.

Is Pilgrim Plant Fixed?

A plant like Pilgrim has somewhere near 5 million parts in it. No doubt Entergy opened up their wallet to fix the component problems identified in last winters trip. They probably diverted funds to fix last winter's problem from other weak and aging components.  

There is a huge reservoir of weak and degrading components still in the Pilgrim plant waiting to bite them on the ass. The defects in the NRC and Entergy's organization are still there. They fixed the easiest symptoms.  

Nothing more exemplifies this than the repeated problems with the main condenser and the including consecutive start-ups and last Friday. 

Remember the main problem here is the lack of horsepower of the NRC...the inability to anticipate problems like last winter and make them fix degrading and broken safety components before a accident happens.

Pilgrim Station gets some good news from the NRC

By: Manomet Current | June 19, 2015 

For the second time in six months, a team of NRC inspectors traveled to the Pilgrim Nuclear Power Station to check on work aimed at preventing a recurrence of unplanned shutdowns at the facility.

But unlike the review completed in December, the more recent inspection has found that Entergy, the owner of the plant, has now satisfactorily addressed the areas of weakness, according to a statement released by Nuclear Regulatory Commission Spokesman Neil Sheehan.

 The inspection team’s report calls for the closure of two low to moderate safety inspection findings. That means that as of the end of this month, the NRC’s heightened oversight in response to a series of unplanned shutdowns would end, Sheehan stated.

The NRC is continuing to review an apparent violation issued on May 27 at Pilgrim. That finding involves the maintenance of the plant’s safety relief valves and stems from a special inspection conducted at the site after an unplanned shutdown in January amid a winter storm. If that violation, preliminarily classified is finalized, the agency will  determine the amount of increased scrutiny that should be applied to the plant. A decision on the finding is not unexpected until sometime this summer, according to Sheehan.

Lauren Burm, spokesman for Pilgrim Station issued a statement that the company is pleased that the inspection ended in a “successful conclusion.” She went on to state that “We recognize there are additional areas to work on and we are committed to identifying and resolving issues in a timely and effective manner.”

 Pilgrim began receiving additional attention from the NRC In late 2013 after it had three unplanned shutdowns, according to Sheehan.

During the supplemental review carried out last fall, NRC inspectors determined that deficiencies existed in fixing the problems that were found and in understanding what caused the issues. Consequently, the NRC decided to maintain its elevated level of oversight pending another inspection.

The MIke Mulligan's River Bend Special Inspection Results

http://pbadupws.nrc.gov/docs/ML1518/ML15188A532.pdf

UNITED STATES  NUCLEAR REGULATORY COMMISSION

REGION IV
1600 E. LAMAR BLVD
ARLINGTON, TX 76011-4511

July 7, 2015

EA-15-043

Mr. Eric W. Olson, Site Vice President

Entergy Operations, Inc.

River Bend Station

5485 U.S. Highway 61N

St. Francisville, LA 70775

SUBJECT: RIVER BEND STATION – NRC SPECIAL INSPECTION REPORT 05000458/2015009; PRELIMINARY WHITE FINDING

 

Entergy and River Bend Notes

extent of cause or condition investigation...?

https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML050410438

February 10, 2005
Paul D. Hinnenkamp
Vice President - Operations
Entergy Operations, Inc.
River Bend Station
5485 US Highway 61N
St. Francisville, Louisiana 70775

SUBJECT: RIVER BEND STATION - SPECIAL INSPECTION REPORT 05000458/2004012

 4OA7 Licensee-Identified Violation

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. C 10 CFR 55.46.c states in part, “A plant-referenced simulator used for the administration of the operating test or to meet experience requirements . . . must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond . . . .” RBS experienced two reactor scrams (August 15 and October 1, 2004) in which actual plant SRV manipulations caused shrink, swell, and level indications that were different than what was modeled in the simulator. After some investigation by the licensee, it was determined that level variations in the simulator were 6-8 inches different than in the actual plant. Considering that RPV level is normally maintained between Level 8 (51 inches) and Level 3 (9.7 inches), 6-8 inches constitutes approximately a 15-20 percent difference than actual plant condition. Coupled with the fact that most of the operators on shift during the events had never actually manipulated SRVs in the plant, this simulator fidelity deficiency could have an impact on operator performance. This issue was documented in the licensee’s corrective action program in Condition Report CR-RBS-2004-2334. This violation is of very low safety significance because it did not involve an exam or operating test, but did involve a simulator fidelity issue which impacted operator actions and resulted in negative training.

http://pbadupws.nrc.gov/docs/ML1504/ML15042A392.pdf

 

Dates: October 1 through December 31, 2014

EA-14-147

Mr. Eric W. Olson, Site Vice President
\Entergy Operations, Inc.
River Bend Station
5485 U.S. Highway 61N
St. Francisville, LA 70775

SUBJECT: RIVER BEND STATION – NRC INTEGRATED INSPECTION REPORT 05000458/2014005

Dates: October 1 through December 31, 2014
Dear Mr. Olson:

Green. The inspectors identified a non-cited violation of 10 CFR 55.46, "Simulation Facilities," for the failure of the licensee to retain the results of required performance tests for four years after completion, or until superseded by updated test results. The licensee could not locate scenario-based testing documentation conducted for the March 2014 initial license exam. The licensee asserted in writing that the testing was performed, but that the electronic test packages had been lost. This issue was entered into the licensee's corrective action program as CR-RBS-2014-04595.

Specifically, because of the lack of documentation the licensee was unable to demonstrate that its scenario-based testing would ensure the simulator is capable of producing the expected reference unit

response without significant performance discrepancies, or deviation from an approved scenario sequence, for scenarios used to evaluate licensed operators and applicants. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, and the corresponding Appendix I, "Licensed Operator Requalification Significance Determination Process," the finding was determined to have very low safety significance (Green) because it is a "Simulator Testing, Maintenance, or Modification Deficiency." 

http://pbadupws.nrc.gov/docs/ML1413/ML14133A700.pdf

July 7, 2015

EA-15-043

Mr. Eric W. Olson, Site Vice President
Entergy Operations, Inc.
River Bend Station
5485 U.S. Highway 61N
St. Francisville, LA 70775

SUBJECT: RIVER BEND STATION – NRC SPECIAL INSPECTION REPORT 05000458/2015009; PRELIMINARY WHITE FINDING

 

Analysis. The team determined that the failure of the plant-referenced simulator to demonstrate expected pressure response across the main steam isolation valves during conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response could adversely affect the operating crew’s ability to assess plant conditions and take actions in accordance with approved procedures. In accordance with NRC Inspection Manual Chapter 0609, “Significance Determination

Process,” Phase 1 Worksheets, and the associated Appendix I, “Licensed Operator Requalification Significance Determination Process (SDP),” Block 15, the finding was determined to be of very low safety significance because the deficient simulator performance did not negatively impact operator performance in the actual plant during a reportable event. This modeling deficiency did not have any generic training implications, nor did it have any actual impact on operator performance. Therefore, the inspectors determined it did not have any cross-cutting aspect.  
 

 

The New River Bend Feed Pump Scram?

Originally published on June 3, 2015

Just saying, a well maintained plant scram should go like this, a component fails, then the scram should occur without any other component failures or observed component degradation and the operator should be error free. Multiple equipment failures showing up in one scram is very serious...  

6/02 Event Report: The transient began with a trip of Reactor Feed Pump 'A', followed by a Reactor Scram and a trip of Reactor Feed Pump 'C'.

It is startling after all this NRC attention and all the recent problem in River Bend: it initiated from a feed water trips, then another feed water trip and then issues with the Reactor Feedwater Master Level Controller problems.

Actually is sounds like a out of control Reactor Feedwater Master Level Controller set this whole thing off.

NRC to me: Jan 21, 2015 No: IV-15-003 "during which you expressed concern related to equipment issues leading to scrams and operator performance following scram."

The Pilgrim special inspection turned into being a train wreck, we recently got a special inspection on Indian Point, along with the two special inspection at River Bend. That is four inspection ongoing at Entergy Plants...at least 38% of their sites under special inspection. The double hitter River Bend special Inspection should be released within days.

Imagine that, another set of feed pump trips, pathetic!!!

Unprecedented NRC attention at River Bend with feed pump trips, out of control reactor water lever control and scrams...now we got another one. 

The symptoms of a totally impotent NRC...

The next Pilgrim scram or down power is right around the corner? 

Power Reactor	Event Number: 51112
Facility: RIVER BEND Region: 4 State: LA Unit: [1] [ ] [ ] RX Type: [1] GE-6 NRC Notified By: JACK MCCOY HQ OPS Officer: DANIEL MILLS	Notification Date: 06/02/2015 Notification Time: 02:00 [ET] Event Date: 06/01/2015 Event Time: 21:11 [CDT] Last Update Date: 06/02/2015
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): JACK WHITTEN (R4DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	A/R	Y	90	Power Operation	0	Hot Shutdown

Event Text

REACTOR SCRAM DUE TO LOW REACTOR WATER LEVEL "At 2111 [CDT] River Bend Nuclear Station sustained an Automatic Reactor Scram due to low Reactor Water Level (Level 3). The plant is currently stable, with level being maintained in a normal band of 10 - 51 inches with Condensate and Feedwater. Reactor Pressure is in the prescribed band of 500-1090 psig. The plant is in Mode 3, Hot Shutdown, and will remain in Mode 3 until investigation of the scram is complete. The transient began with a trip of Reactor Feed Pump 'A', followed by a Reactor Scram and a trip of Reactor Feed Pump 'C'. Reactor water level was recovered with Reactor Feed Pump 'B' to a normal post scram level band. There was a problem noted with the Reactor Feedwater Master Level Controller; this was mitigated by the Operator placing the controller to manual. There was no subsequent Level transient. Reactor Pressure was stabilized in normal pressure band with Turbine bypass valves. During the transient, a Reactor Recirculating Flow Control Valve Runback was not received as expected. Reactor Recirculating Pump 'A' responded as expected to transient [switching to low pump speed], Reactor Recirculating Pump 'B' tripped during transient. A Level 3 isolation signal was received, all expected isolations occurred. "The cause of the transient is currently under investigation." The reactor is stable in Mode 3 with decay heat being removed via turbine bypass valves, and a normal electrical line up. The NRC Resident Inspector has been notified. * * * UPDATE FROM JACK MCCOY TO HOWIE CROUCH AT 0712 EDT ON 6/2/15 * * * "At 2231 on 6/1/15, Reactor Water Cleanup System isolated on High Reactor Water Cleanup System Heat Exchanger room temperature due to loss of Turbine Building chill water during the initial transient. All Reactor Water Cleanup System Valves isolated as expected. Reactor Water Cleanup was the only system affected by this isolation signal." The licensee has notified the NRC Resident Inspector