Junk Plant Watts Bar 1&2: Double Red Finding on RHR

Update May 7, 

Mark my words, all of out institutions are breaking down, most disappointingly, the democrats. I trust nothing from any one of our institutions...  

I have no confidence these computer generated safety evaluations are accurate. I am a testing guy. The only thing I would believe is a mock up of the RHR system, then put in the highest concentration of air into the system, plus some extra amount extra, then figure out when severe water hammer begins. 

At the worrying bottom of it all, these guys look so confused over a settled issue. 

"Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. 

The issue here is these problems are so complex, they can reanalyze the problem,   then come up with any conclusion they need.

Engineering Computer simulation is highly susceptible to fraud and corruption. It too complicated but for a extremely small group of highly trained and educated engineers. How can the legal system catch this kind of fraud. The only way you catch it is after a big accident and something occurs unexpected.  

!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor	Event Number: 53356
Facility: WATTS BAR Region: 2 State: TN Unit: [ ] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: JUSTIN GALLAGHER HQ OPS Officer: DAVID AIRD	Notification Date: 04/22/2018 Notification Time: 04:28 [ET] Event Date: 04/22/2018 Event Time: 02:22 [EDT] Last Update Date: 05/04/2018
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(v)(B) - POT RHR INOP	Person (Organization): ALAN BLAMEY (R2DO)

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

BOTH TRAINS OF RESIDUAL HEAT REMOVAL INOPERABLE "On April 22, 2018 at 0222 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 entered TS [Technical Specifications] LCO [Limiting Condition for Operation] 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 0227 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions." The NRC Resident Inspector has been notified. * * * RETRACTION FROM TONY PATE TO HOWIE CROUCH ON 5/4/18 AT 1455 EDT * * * "This event is being retracted. The initial report was based on a conservative acceptance criteria for gas accumulation adopted on April 19, 2018 when it was determined that the previously used acceptance criteria for gas accumulation in the ECCS was non-conservative. Additional analysis has subsequently been performed and determined that a higher gas accumulation acceptance criteria does not challenge operability. With a void of less than the acceptance criteria, in the event of ECCS actuation, the system piping support loads will remain within structural limits and the piping system will remain operable. Therefore, both trains of Unit 2 RHRS were operable and the previously reported 10 CFR 50.72(b)(3)(v)(B) event is being retracted. "The NRC Resident Inspector staff has been informed of this event retraction." Notified R2DO (Desai) of this retraction.

Update May 1

The Nuclear Regulatory Commission has launched a special inspection into how an excess amount of gas is in the residual heat removal system used to help shut down the reactors at TVA's Watts Bar Nuclear Power Plant near Spring City, Tenn.

Two weeks ago, TVA informed the NRC that revisions to its initial calculations had reduced the acceptable size of a void due to gases in the system that helps cool down the Watts Bar reactors during their shut down. On April 21, the accumulated gas in the Unit 1 system was found to have exceeded the acceptable value, and on April 22, the same observation was made on Unit 2.


The NRC inspectors will review the sequence of events, drawings, calculations and acceptance criteria, walk down portions of the plant's systems, evaluate TVA's response and assess the adequacy of actions to address the causes of the issues. NRC spokesman Roger Hannah said the issue identified by TVA does not warrant having to shut down either of the Watts Bar units and a report on the onsite visit this week is expected by June

Absolutely no excuse for this: 

NRC GENERIC LETTER 2008-01: MANAGING GAS ACCUMULATION IN EMERGENCY CORE COOLING, DECAY HEAT REMOVAL, AND CONTAINMENT SPRAY SYSTEMS

***This is the Trump NRC remember.  

This indicates how bankrupt the NRC is. Why didn't catch this decades ago and upon first startup of unit 2? 

This should be a double red finding. Remember Brown's Ferry RHR issues with a non functioning RHR injection valve. So basically the RHR was non operable for the life of the both plants. They should have gutted out all piping of the RHR and replaced them at a angle where all the air would have accumulated at the highest point in the system. This is a initial design flaw!!!

Power Reactor	Event Number: 53349
Facility: WATTS BAR Region: 2 State: TN Unit: [1] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: CHARLES BROESCHE HQ OPS Officer: THOMAS KENDZIA	Notification Date: 04/20/2018 Notification Time: 00:55 [ET] Event Date: 04/19/2018 Event Time: 19:44 [EDT] Last Update Date: 04/20/2018
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(B) - UNANALYZED CONDITION	Person (Organization): ALAN BLAMEY (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

UNANALYZED CONDITION RELATED TO EMERGENCY CORE COOLING GAS ACCUMULATION ACCEPTANCE CRITERIA "On April 19, 2018 at 1944 EDT, Watts Bar Nuclear Plant (WBN) determined that a preliminary analysis shows current acceptance criteria for gas accumulation in the WBN Unit 1 and Unit 2 Safety Injection System (SIS) and Residual Heat Removal System (RHRS) discharge piping may be non-conservative. The surveillances that check void values and allow venting of the systems are to be performed utilizing conservative criteria at more frequent intervals to ensure gas void volumes remain under acceptable limits. Additional analysis is being performed to determine final actions. "The NRC Resident Inspector has been notified."

Unit 1 BOTH TRAINS OF RESIDUAL HEAT REMOVAL INOPERABLE

"On April 21, 2018 at 2152 EDT, Watts Bar Nuclear Plant (WBN) Unit 1 entered TS [Technical Specifications] LCO [Limiting Condition for Operation] 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 2222 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions."

The NRC Resident Inspector has been notified.

Unit 2 BOTH TRAINS OF RESIDUAL HEAT REMOVAL INOPERABLE

"On April 22, 2018 at 0222 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 entered TS [Technical Specifications] LCO [Limiting Condition for Operation] 3.0.3 due to both trains of the Residual Heat Removal System (RHRS) becoming inoperable. During surveillance testing, the gas void values on Emergency Core Cooling System (ECCS) piping common to both trains did not meet acceptance criteria. This caused both RHRS trains to become inoperable. Operations subsequently vented the RHRS to meet the acceptance criteria and exited TS LCO 3.0.3 at 0227 EDT. More frequent surveillances will be conducted to monitor gas void volumes while additional analysis is being performed to determine corrective actions."

The NRC Resident Inspector has been notified.

Junk Plant Braidwood: What Is Going On?

Two independent components failing on the diesel generators is not good. Sounds like a not spending enough money on maintenance problem. 

Basically the second one is a safety injection test. It simulates a LOCA up to the injection valves including the pumps. It is a very important test and indication the reliable of these systems since the last test.  

Power Reactor	Event Number: 53353
Facility: BRAIDWOOD Region: 3 State: IL Unit: [1] [ ] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: ALEX TRESPALACIOS HQ OPS Officer: DONG HWA PARK	Notification Date: 04/20/2018 Notification Time: 17:57 [ET] Event Date: 04/20/2018 Event Time: 10:42 [CDT] Last Update Date: 04/20/2018
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(v)(B) - POT RHR INOP	Person (Organization): KENNETH RIEMER (R3DO)

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

Event Text

BOTH DIESEL GENERATORS INOPERABLE On Friday, April 20, 2018 at 1042 CDT, Braidwood Station Unit 1 was at 0 percent power in Mode 6. The 1A Diesel Generator (DG) was inoperable with troubleshooting in progress. The 1B DG was being run for a normal monthly run in accordance with 1 BwOSR 3.8.1.2-2, 'Unit One 1B Diesel Generator Operability Surveillance,' and subsequently tripped. The trip was due to a failure of the overspeed butterfly valve actuator and springs, and not an actual overspeed condition. The unit entered Technical Specification (TS) 3.8.2, 'AC Sources - Shutdown,' Condition B for required DG inoperable. All required TS actions were met at the time of the 1B DG inoperability. The offsite power source remains available. At no time was residual heat removal lost. "This event is reportable under 10 CFR 50.72(b)(3)(v)(B) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. "The licensee has notified the NRC Resident Inspector."

ower Reactor	Event Number: 53358
Facility: BRAIDWOOD Region: 3 State: IL Unit: [1] [ ] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: RICH ROWE HQ OPS Officer: DONG HWA PARK	Notification Date: 04/22/2018 Notification Time: 22:40 [ET] Event Date: 04/22/2018 Event Time: 16:46 [CDT] Last Update Date: 04/22/2018
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): KENNETH RIEMER (R3DO)

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

Event Text

UNDERVOLTAGE ACTUATION OF THE ENGINEERED SAFETY FEATURE BUS "On Sunday, April 22, 2018 at 1646 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned 1A Diesel Generator (DG) Emergency Core Cooling System (ECCS) Actuation Surveillance, initiating the 1A DG to emergency start and sequence loads on a safety injection signal. Following the 1A DG solely supplying electrical power to Bus 141, the 1A DG lost voltage, resulting in an unplanned UV actuation of ESF Bus 141. The 1A DG output breaker was manually opened and local emergency stop of the 1A DG was attempted. The 1A DG continued to run at idle. Fuel supply was secured to the 1A DG and the engine stopped. Subsequently, operators restored power to ESF Bus 141 from the Unit 1 Offsite Power Source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. "This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. "The licensee notified the NRC resident inspector."

Junk Braidwood's Head: Astonishing Cracks On "Top Head To Upper Center Disc Weld"

I don't remember ever seeming something like this? 

A crack in the head would is considered a impossible accident. It would bypass every safety system in the plant. There is no question there would be a meltdown. They would quickly lose all indications of water level. This is worst than the "hole in the head" at David Besse because it is a head weld. 

Something is funny with this? It doesn't seem probable all these cracks and indications would show up in one testing interval. Did they use a updated UT device? There has been a lot of NRC discussions with the accuracy of UT testing. Did the changes with these discussion lead to better UT testing. The implication are these flaws and cracks for many years, the new UT testing just showed them. What implication nationwide-how many non detected cracks and flaws are active at other plants.     

The UT identified 19 indications, 9 of which are not acceptable per ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWB-3510.

Power Reactor	Event Number: 53354
Facility: BRAIDWOOD Region: 3 State: IL Unit: [1] [ ] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: RICH ROWE HQ OPS Officer: DONG HWA PARK	Notification Date: 04/20/2018 Notification Time: 22:22 [ET] Event Date: 04/20/2018 Event Time: 17:30 [CDT] Last Update Date: 04/20/2018
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): KENNETH RIEMER (R3DO)

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

Event Text

DEGRADED REACTOR VESSEL HEAD "On Friday, April 20, 2018 at 1730 CDT, during the Braidwood Station Unit 1 refueling outage (A1R20), a scheduled ultrasonic test (UT) was performed on the top head to upper center disc weld of the Unit 1 reactor head. The UT identified 19 indications, 9 of which are not acceptable per ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWB-3510. "This event is reportable under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'. "The licensee notified the NRC Resident Inspector."