US Nuclear Plants Having a Cascade of Failures Mechanical, Electrical, and Control All At Once

Aloha folks, please drop a comment, or add other plant problems in the comments and I will add to main body of the article. 

These old nuke plants, many past their design life, and with for profit corporations skimping on maintenance whilst their industry is under pressure from cheap gas and cheap solar, well, they are getting seriously dangerous.

Each instance below shows 4 major, multiple system failures, in 4 days since Jan 29.   A component fails, and then a backup system or other system also fails, then the outside power trips out.  

Many do not know, but without outside power, nuclear plants are at risk of melting down.    They usually have 2 or 3 EDG Emergency Diesel Generators, but they also have problems, and they are NEVER tested in a full load simulated outage test.  

It' time to shut these plants down, which the NRC could do, if they were not captured by the industry they are meant to regulate.    Without nuclear, the need for the NRC itself would go away.

You can see the power production status of all US reactors here
 http://www.nrc.gov/reading-rm/doc-collections/event-status/reactor-status/ps.html

Here you can see the scheduled outages, so if a plant is less than 100% and it is not in a scheduled outage, then you can assume there is a real problem

https://www.roadtechs.com/nukeout.htm

stock out
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Feb 1: US_NPP Davis Besse 1 (OH) AUTOMATIC REACTOR TRIP DUE TO REACTOR PROTECTION SYSTEM ACTUATION

 http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160201en.html#en51696

"At 1322 EST, with the unit operating at approximately 100% full power, an automatic reactor trip occurred due to actuation of Reactor Protection System (RPS) Channel 4. The cause of the RPS actuation is being investigated at this time. Nuclear Instrumentation calibration for RPS Channel 2 was in progress at the time of the trip, with Channel 2 in bypass and Channel 1 in trip. All control rods fully inserted. Immediately post trip, the Steam Feedwater Rupture Control System actuated due to high Steam Generator 1 level due to unknown causes. The Main Steam Isolation Valves closed and Auxiliary Feedwater started as expected. Secondary side relief valves lifted in response to the trip, with two of the relief valves (one on each header) not properly reseating until operators manually lowered Main Steam Header pressure. The Bayshore 345 kV Offsite Electrical Distribution Circuit automatically isolated at the time of the unit trip. This was unexpected. The remaining offsite circuits remain in service.
"The unit is currently in Mode 3 (Hot Standby) and stable, at approximately 550 degrees F and 2155 psig. Steam is being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage, and all safety systems functioned as expected…


FEB 1: US_NPP Davis Besse UNANTICIPATED SFRCS ACTUATION WHILE RESTORING MAIN FEEDWATER TO STEAM GENERATORS

 http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160201en.html#en51702

"At 0123 EST, with the unit shutdown in Mode 3 (Hot Standby), during the performance of procedure DB-OP-06910, 'Trip Recovery,' while attempting to restore main feedwater to the Steam Generators, Davis-Besse received a Steam Feedwater Rupture Control System (SFRCS) 'reverse delta pressure' signal to the Auxiliary Feedwater System (AFW). The Auxiliary Feedwater System was operating at the time, feeding the Steam Generators.
"The SFRCS signal did result in actuation/closure [of] several valves in the Main Steam System, as the SFRCS signal is designed to do. This SFRCS signal/valve actuation was not anticipated.
"The unit remained in Mode 3 and is stable. This actuation did not have any negative impact to the AFW system and the ability to feed the steam generators

Quote
There is no known primary to secondary leakage, and all safety systems functioned as expected…

stock here
BALD FACED LIE

1. 2 Steam valves didn't close
2. Offsite power tripped out
3. Other valves were activated, and worked as intended, BUT it was due to a false signal from a control system.

Jan 29---US_NPP Riverbend 1 (LA) SPECIFIED SYSTEM ACTUATION AFTER LOSS OF ONE OFFSITE POWER SOURCE

 http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160201en.html#en51701

"On January 29, 2016, at 1518 CST, with the plant in cold shutdown, power was lost on reserve station service (RSS) line no. 1. This is one of two sources of offsite power required by Technical Specifications. The power loss de-energized the Division 1 onsite AC safety-related switchgear, causing an automatic start of the Division 1 emergency diesel generator (EDG). The Division 1 reactor protection system (RPS) bus was also de-energized, causing a half-scram signal. Approximately 8 minutes later, a full actuation of the RPS occurred due to a high water level condition in the control rod drive hydraulic system scram discharge volume header. All reactor control rods were already fully inserted.

"The loss of Division 1 RPS also caused the actuation of the Division 1 primary containment isolation logic. The Division 1 isolation valves in the balance-of-plant systems closed as designed. Both trains of the standby gas treatment system actuated.

"The loss of RSS no. 1 occurred during post-modification testing on relays at the local 230kV switchyard. The exact cause of the event is under investigation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A).

"The unit remains in cold shutdown with 1 source of offsite power and all 3 [EDG] available.

Jan 29 ----US_NPP Duane Arnold 1 (IA) OFFSITE POWER DECLARED INOPERABLE DUE TO LOWERING GENERATOR VOLTAGE
 http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160201en.html#en51697

"At 0920 CST on 1/29/16, while performing main generator voltage adjustments at the direction of the transmission operator, Duane Arnold Energy Center (DAEC) switchyard voltage lowered to the calculated point where offsite sources were considered inoperable. During this event, both onsite emergency diesel generators were fully operable and capable of performing their intended safety function.

"While no safety concerns arose, Technical Specification (TS) Limiting Condition for Operability (LCO) 3.8.1.a Condition C was entered due to two offsite electric power circuits being inoperable. Immediate actions were taken to adjust main generator voltage to restore switchyard voltage. In less than 10 minutes, both offsite circuits were declared operable and LCO 3.8.1.a Condition C was exited. This resulted in a reportable event pursuant to 10CFR50.72(b)(3)(v)(D).

"During the time of reduced grid voltage, no bus low-voltage alarms nor any equipment issues occurred. The cause of this event is under investigation, and there are not any current operability concerns with the offsite power circuits."

 TS REQUIRED SHUTDOWN DUE TO UNIDENTIFIED LEAKAGE IN DRYWELL

"At 2100 hours [EST], on January 23, 2016, the Perry Nuclear Power Plant commenced a reactor shutdown due to unidentified leakage in the drywell. At 2122 hours, drywell unidentified leakage exceeded the Technical Specification 3.4.5.d limit of 'less than or equal to 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in Mode 1.' The unidentified leakage increased to approximately 3.8 gpm at 2122 hours. Current unidentified leakage is 3.02 gpm.

"Technical Specification 3.4.5 actions allow 4 hours to reduce the leakage within limits or be in Mode 3 within 12 hours and Mode 4 within 36 hours. The plant is required to be in Mode 3 by 1322 hours on January 24, 2016 and Mode 4 by 1322 hours on January 25, 2016.

"A drywell entry will be made in Mode 3 to identify the leak source.

"This notification is being made due to an expected inability to restore the leakage within limits prior to exceeding the LCO action time.

"Follow up question from NRC: Event times do not match (2100 versus 2122) - explained downpower was commenced at 2100 with leakage less than TS limit. When Reactor Core flow was reduced, un-identified leakage increased above the TS limit."

The Licensee has notified the NRC Resident Inspector.

* * * UPDATE FROM MIKE DOTY TO DANIEL MILLS AT 1123 EST ON 1/24/16 * * *

"At 1007 hours, on January 24, 2016 with the plant at 8% power during a feedwater shift to place the motor feed pump in service, reactor level rose to the level 8 scram set point and the Reactor Protection System (RPS) initiated, scramming the reactor. During the scram, all rods fully inserted into the core. Decay heat is being removed via turbine bypass valves to the main condenser. Reactor level control is currently being maintained via feedwater. The plant is stable with cool down and depressurization to Mode 4 to follow. The cause of the rise in feedwater level is under investigation. This notification is being made under 50.72(b)(2)(iv)(B) for a RPS initiation while critical."

All safety shutdown systems are available. The electric plant is in its normal shutdown alignment being supplied by offsite power.

The licensee has notified the NRC Resident Inspector. Notified R3DO (Cameron). NRR (Morris) and IRD (Gott) were notified via email.

* * * UPDATE FROM DAVID O'DONNELL TO HOWIE CROUCH AT 1915 EST ON 1/24/16 * * *

"Following a shutdown required by plant Technical Specifications a small leak was identified coming from the Reactor Recirculation Loop A Pump Discharge Valve vent line. The Recirculation Loop is part of the reactor coolant system making this reportable under 50.72(b)(3)(ii)(A) as a degraded condition. It was subsequently determined to require a plant cool down in accordance with Technical Specification 3.4.5, Action C which requires the plant to be in MODE 4 within 36 hours. Technical Specification 3.4.5 was previously entered for increased unidentified leakage in the drywell. The plant is required to be in Mode 4 by 1322 hours on January 25, 2016."

The licensee has notified the NRC Resident Inspector. Notified R3DO (Cameron). NRR (Morris) and IRD (Gott) were notified via email.