Junk Plant Susquehanna Leak Violations: At Least A Red finding?

Is this Susquehanna’s version of Indian Point’s baffle bolt cracks problem???

Basically the NRC is implicated in this outcome. They weren't assertive enough to call for a immediate showdown. We don't know the role of the agency play with their complicity in not making Susquehanna detect the flaw before the leak showed up. It is now a classic conflict of interest with them setting up the violation because they were complicit with their non participation. They need to get a outsider (outside the NRC)to inspect this violation and then set the violation level. All they are doing is basically grading themselves.



May 9, 2016 @ 11 am : Updated and now with a picture of the LPRM tubes.

A month from the big maintenance refueling outage and they got two reactor water leaks???  

• BERWICK, Pa., March 12, 2016 /PRNewswire/ -- Operators at Talen Energy's Susquehanna nuclear power plant disconnected the Unit 1 reactor from the electrical grid late Friday, March 11 into early Saturday, March 12  to begin a scheduled refueling and maintenance outage.

It is really hot under the reactor. More so if they have a lot of fuel failure. This will be the way most US plant expire...basically very high and preventable very high radiation levels will make it too expensive to keep the plants on the grid.

The area under the core is extraordinarily radioactive and contaminated. We had to dress up in triple anti C clothing, a rain suit and respirators. With all the tubes, CDRMs and wires directly under the bottom reactor head, it looked like a up-side-down rain forest, if you had a imagination. It was constantly leaking astonishing highly radioactive water down on us. If they gave us a job down there, they'd say mockingly, "buddy, you are going down to our rain forest" with a huge smile.      

Susquehanna is a very busy two plant. They got problems with too many plant scrams. Here is another blatant example of dangerous certainty/uncertainty gaming to override safety rules. It’s basically Davis Besse redux. It’s as if nobody in the industry has learned a thing about pressure barrier leaks and Davis Besse.

REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH REGULATORY GUIDE 1.45 (Draft was issued as DG-1173, dated June 2007) 

GUIDANCE ON MONITORING AND RESPONDING TO REACTOR COOLANT SYSTEM LEAKAGE

A. INTRODUCTION 

This revision to Regulatory Guide 1.45 (Ref. 1) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for use in implementing the regulatory requirements specified below with regard to selecting reactor coolant leakage detection systems, monitoring for leakage, and responding to leakage. This guide applies to light-water-cooled reactors.

General Design Criterion (GDC) 14, “Reactor Coolant Pressure Boundary,” as set forth in Appendix A, “General Design Criteria for Nuclear Power Plants,” to Title 10, Part 50, “Domestic Licensing of Production and Utilization Facilities,” of the Code of Federal Regulations (10 CFR Part 50), (Ref. 2), requires that licensees or applicants design, fabricate, erect, and test the reactor coolant pressure boundary (RCPB) so as to ensure an extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture. As a result, the design of these nuclear components normally follows the criteria established in Section III of the Boiler and Pressure Vessel Code (Ref. 3) promulgated by the American Society of Mechanical Engineers (ASME). 

During the design phase, degradation-resistant materials are normally specified for reactor coolant system (RCS) components. However, materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and the normal and upset operating environments they experience. Such material degradation could lead to leakage of the reactor coolant. Consequently, GDC 30, “Quality of Reactor Coolant Pressure Boundary” (Ref. 2), requires that plants provide the means for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. 

(ASME: Basically a private regulator writing up corporate sponsored engineering codes from money.)

Basically the containment has rudimentary instrumentation with detecting leaks in the containment. You get it, basically the first leak obscured the control room indications of the second leak? There is no way to detect the second leak. Do you know what a ghost containment leak is? Its having a history of a bunch of non pressure barrier or a prolonged non pressure barrier leaks in containment. This imprints in the minds of the control operators. Months or years later, a pressure barrier leak emerges. The control room operators then assume the new leak is coming from the same component that has past leakage problems. This is Davis Besse and TMI.

Power Reactor	Event Number: 51987
Facility: SUSQUEHANNA Region: 1 State: PA Unit: [1] [ ] [ ] RX Type: [1] GE-4,[2] GE-4 NRC Notified By: LONNIE CRAWFORD HQ OPS Officer: HOWIE CROUCH	Notification Date: 06/08/2016 Notification Time: 07:01 [ET] Event Date: 06/08/2016 Event Time: 02:26 [EDT] Last Update Date: 06/08/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): FRANK ARNER (R1DO)

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

DISCOVERY OF UNISOLABLE REACTOR PRESSURE BOUNDARY LEAKAGE "Susquehanna Unit 1 identified RPV [reactor pressure vessel] pressure boundary leakage from [local power range monitor] LPRM 24-09 housing above the flange during an under vessel leak inspection on 06/08/2016 at 0226 EDT. The leakage point is a through wall indication on the ASME Class 1 LPRM stub tube. The leakage is not isolable from the reactor vessel. The reactor was in Mode 4 at the time of discovery. "This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A)." A repair plan is being formulated. The licensee has notified the NRC Resident Inspector.

They usually have two sumps in containment. One called identified leakage and the other called unidentified leakage and they measure the leakage on a hourly basis. The non identified leakage sump collects from mostly water from the non primary water cooling water systems like the service water system or component cooling water system. Say for the reactor coolant pump motor bearing cooling. A little leakage from non pressure barrier leakage is not risky and you can delay shutting down. The identified leakage can only come from pressure barrier leakage. 

Is something wrong with NRC inspector training? Why didn't they immediately order a plant shutdown?  

But the deal here is, you positively have to know it’s not coming from a primary system pressure barriers like the seal water system or potentially like leakage from the reactor vessel. The idea you have two pressure barrier leaks is chilling.

So far I haven't found any LERs on this kind of leakage. Is this kind of leak a industry first? It looks like that according to the NRC documents. Is this Susquehanna's Indian Point baffle bolt crack problem? You have to assume there is more flaws or leaks in the tube. Has the tubes ever been UT'd. What if the Uting of the tubes they discovered many more flaws and cracks throughout the tube. You have to assume all the tubes are degraded. You would have to UT all the tubes. I believe these guys just came out of a outage on May 4, 2016. Why didn't pressure testing of the primary system at end of outage detect the leakage. These leaks can't develop in a month. Did they falsify the testing paperwork?

This is going to be a prolonged outage for weeks and maybe months.  

LPRM Housing

I'd like to know what the pressure of the reactor vessel when the employees measured the leakage. The assumption at power is the water was at approximately 500 degrees and at least 500 psig. The leaking water immediately turned into steam. The steam turns into water when in contact with cooler components and maybe on the surface of the containment. Plus from the bottom of the air cooler in the containmen. This water will then drain into the non identified leakage sump. If the reactor is cooled down then the pressure is about zero. There is a big difference in measured leakage between a zero pressure plant and one at 500 psig and degrees. Besides the water leakage rate, it's an additional large artificial heat load into containment.  

I might got the identified and non identified leakage switched around. But it is the same concept.  

It will give you a false indication its not pressure barrier water. This is a well-known phenomenon. Like I said, employees can’t enter the containment at power because the radiation levels are too high. It is very difficult to tell whether its identified leakage or unidentified leakage. Regulations state, you have to be 120% sure you got non pressure barrier water leaking into containment or you must assume it is pressure barrier water. Then it is a mandatory emergency shutdown.

• Did they have or detect "containment" abnormal pressure or temperature symptoms indicating a high temperature leak?

• This is Davis Besse head leak. These plants monitor the containment air radiation levels. They sample the air radiation levels at least daily. Primary system water is very radioactive. So if pressure barrier water is leaking into the containment, it tremendously spikes up the air radiation level (particulate and gaseous). Davis Besse ignored the spiking containment air radiation levels. The air radiation level is really a sophisticated water leak detection system, it was put in the plant for that purpose. It amplifies the indications of detecting a small primary system leaks.        

If you got pressure barrier leakage, we can’t predict reliably how a crack would grow in a pipe or vessel. If it’s in the reactor vessel and the crack grows, it potentially overrides all of the safety designs of any plant. We are completely powerless to prevent a meltdown.  The NRC and licencees assume a reactor vessel crack and leak is a impossibility. So this accident is not designed into the plant. That is why it is really important to shutdown the plant immediately based on incomplete information.

General Technical Specification requirements and explanations 

PWR not BW: but same concept without steam generators. This is tech specs and constitutes a direct violation of regualtion.

B 3.4 REACTOR COOLANT SYSTEM (RCS) 

B 3.4.13 RCS Operational LEAKAGE BASES 

BACKGROUND 

Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE. 10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. 

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public. 

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leak tight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection. This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA). Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analyses for events resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is [one gallon per minute] or increases to [1 gallon per minute] as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than 150 gallons per day is significantly less than the conditions assumed in the safety analysis. 

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.
 RCS operational LEAKAGE shall be limited to: 

a. Pressure Boundary LEAKAGE 

No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. 

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid -is only briefly released via safety valves and the majority is steamed to the condenser. The [1 gpm] primary to secondary LEAKAGE assumption in the safety analysis is relatively inconsequential.
The [SLB] is more limiting for site radiation releases. The safety analysis for the [SLB] accident assumes [1 gpm] primary to secondary LEAKAGE in one generator as an initial condition. The dose consequences resulting from the [SLB] accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits). 

The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement. Unidentified LEAKAGE (continued) 

One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary. 

c. Identified LEAKAGE 

Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system. 

d. Primary to Secondary LEAKAGE througqh Any One SG
The limit of 150 gallons per day per steam generator (SG) is based on the Operational LEAKAGE Performance Criterion in the Steam Generator Program. The Steam Generator Program criterion states: 

"The RCS operational primary-to-secondarv leakage through any one steam generator shall be limited to 150 gallons per day.' 

The RCS Operational primary to secondary LEAKAGE is measured at standard temperature and pressure.
The operational LEAKAGE rate limit applies to LEAKAGE in any one steam generator. If it is not practical to assign the LEAKAGE to an individual steam generator, all the LEAKAGE should be conservatively assumed to be from one steam generator.

This is really bad professionally. They made a dangerous assumption for a week or more the leak was non pressure barrier water. With the facts known today, they were immediately required to shut down.  The is not a hard call, if you know your limitations on differentiating non identified leakage from identified leakage, you just shut down the plant to fix the leakage.

This should be a red finding at least. They made a safety call on maliciously incorrect information.    

Junk Plant Susquehanna: Crack in Reactor Vessel?

05000387/388...its a 34 year old plant. This is a BWR-4.  

So this the first nuclear plant owned by a private equity firm in the USA. These guys are already cutting employee jobs at Susquehanna. 

Talen Energy to be Acquired by Riverstone

Stockholders to receive $14.00 per share in cash

Transaction has total enterprise value of approximately $5.2 billion

ALLENTOWN, Pa., June 3, 2016 /PRNewswire/ -- Talen Energy Corporation (NYSE: TLN), a competitive energy and power generation company that owns or controls 16,000 megawatts of capacity in eight states, announced today that it has entered into a definitive merger agreement with affiliates of Riverstone Holdings LLC ("Riverstone"), a private investment firm. 

Under terms of the merger agreement, all outstanding shares of Talen Energy common stock not currently owned by Riverstone affiliated entities will be acquired for $14.00 per share in cash. Affiliates of Riverstone currently own approximately 35 percent of the outstanding shares of Talen Energy common stock, which were issued in the June 2015 transaction that established Talen Energy by combining competitive generation assets that had been owned by affiliates of PPL Corporation and affiliates of Riverstone.
The purchase price represents a 56 percent premium to the closing price of $9.00 per share on March 31, 2016, the last trading day before public reports of a potential sale of Talen Energy, and a 101 percent premium to the 60-day volume-weighted average price of $6.95 per share through March 31.

When did they discover the leak. They are supposed to shutdown quickly with a pressure boundary leakage? 

Power Reactor	Event Number: 51983
Facility: SUSQUEHANNA Region: 1 State: PA Unit: [1] [ ] [ ] RX Type: [1] GE-4,[2] GE-4 NRC Notified By: LONNIE CRAWFORD HQ OPS Officer: STEVE SANDIN	Notification Date: 06/06/2016 Notification Time: 08:13 [ET] Event Date: 06/06/2016 Event Time: 05:56 [EDT] Last Update Date: 06/06/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): DAN SCHROEDER (R1DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	9	Startup	8	Startup

Event Text

ISOLABLE LEAKAGE IDENTIFIED FROM SEAL WATER LINE WELD INSIDE RCS PRESSURE BOUNDARY "Susquehanna Unit 1 commenced a manual shutdown on 06/05/2016 for a maintenance outage. At 2202 hours [EDT] on 06/05/2016, operators began reducing power in accordance with plant procedures. At 0352 hours on 06/06/2016, the Main Turbine was tripped with reactor power at approximately 15%. "The Mode switch was taken to 'STARTUP/HOT STANDBY' (Mode 2) at 0515 hours on 06/06/2016. Manual insertion of control rods was paused as scheduled for entry into the drywell for inspections. There were no ESF actuations.

It is not so bad if it is a water seal leakage?  

"At 0556, the licensee identified leakage from a weld on seal water line piping connected to the 1B reactor recirculation pump seal area. The location is within the reactor recirculation loop isolation valves, therefore is isolable from the reactor vessel. The piping is ASME Class 2 and is reactor coolant pressure boundary. The reactor was in Mode 2 at the time of discovery. "This event is being reported as a plant shutdown required by technical specifications pursuant to 10CFR50.72(b)(2)(i) and degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A). "Activities are continuing to achieve cold shutdown." The licensee informed the Commonwealth of Pennsylvania and the NRC Resident Inspector.

Such incompetence, the pump seal area and the reactor vessel is a large distance apart. 

Susquehanna to complete additional work on Unit 1 

11:18 ET from Susquehanna Nuclear LLCBERWICK, Pennsylvania, June 8, 2016 /PRNewswire/ -- The Susquehanna nuclear power plant in Luzerne County, Pennsylvania made a non-emergency notification to the Nuclear Regulatory Commission (NRC) on Wednesday, June 8, 2016. As was previously reported, Susquehanna Unit 1 was safely shut down on June 6 for planned maintenance to

So this is two seperate leaks? Now it sounds like a instrument nozzle welded inside the reactor vessel leak and it is unisolable.  

identify and repair a small leak inside the containment structure that completely surrounds the reactor. During the investigation, Susquehanna staff identified and reported additional leakage from an instrumentation connection to the reactor vessel. The leak rate is currently measured at two drops per minute. 

Unit 1 remains shut down and the condition is not a safety risk for plant workers or the public. "We have established a dedicated team of nuclear professionals to further

The seal piping fix would be really a easy to fix. A crack in the vessel would be bringing into the plant outside experts to this degree. Brace yourself for the nuclear industry quake if this turns out to be a crack in the reactor vessel.   

evaluate the situation and develop a plan to safely complete repairs, said Robert Franssen, plant general manager. "We will continue to work with the NRC, industry experts and our equipment vendors to understand the issue, evaluate our repair options, complete the necessary work and safely return the unit to service."
Susquehanna Unit 2 continues to operate safely at full rated power. 

The Susquehanna plant, located about seven miles north of Berwick, is jointly owned by Susquehanna Nuclear, LLC, and Allegheny Electric Cooperative Inc., and is operated by Susquehanna Nuclear. For information, visit www.susquehannanuclear.com. 

Susquehanna Nuclear LLC is one of Talen Energy's generating affiliates. Talen Energy (NYSE: TLN) is one of the largest competitive energy and power generation companies in the United States. Our diverse generating fleet operates in well-developed, structured wholesale power markets. To learn more about us, visit www.talenenergy.com.

Another
 

Unit 1 reactor at Susquehanna nuclear plant shuts down

Officials are investigating the cause of the automatic shutdown Thursday of Unit 1 at Talen Energy Corp.'s Susquehanna nuclear power plant.

Talen spokesman Todd Martin said a cause of the shutdown had not been determined. He also said it was premature to speculate when the unit would return to service.

NRC Spokesman Neil Sheehan said a preliminary review indicated that the reactor tripped due to high pressure levels. During testing on a main steam line, it apparently resulted in an "isolation valve closing," Sheehan said.

New Junk Plant Watts Bar 2 Consequential Trip.

Update: June 7

You get it, about a zillion stories about first startup and very few stories explaining what happened during the trip. Something is fishy? 

These guys are the worst in the industry with how they treat safety complaints from employees. I can't even begin to tell how much more worst this will become if the NRC is perceived as pulling their punches concerning the Unit 2's trip.

Say, if risk perspectives indicate this plant needs a special inspection...would the NRC say this is a special case, we need not do the inspection telegraphing Watt Bar 2 is in serious trouble.  

I'll bet you it was chaotic as hell in the control room. How many people do you think was in the control room when this happened.  

Watts Bar received more safety complaints from employees this year than any other US plant

By Dave Flessner, Chattanooga Times Free Press

April 07, 2016 0
Employees at TVA's Watts Bar nuclear plant near Spring City, Tenn., have made more allegations about safety problems to federal regulators this year than any other nuclear plant in the country, according to a new report from the U.S. Nuclear Regulatory Commission.

The NRC has received six safety concerns from employees at Watts Bar already this year, bringing to 54 the number of such complaints made to regulators in the past three years. Only the Millstone Nuclear Power Station, the biggest power plant in Connecticut, has had more such complaints from its employees since 2012.

More allegations have been substantiated by the NRC in 2016 for Watts Bar than for any other plant. In fact, the only allegation substantiated by the NRC in 2016 for the entire fleet of operating nuclear plants was at Watts Bar.

TVA vows to thaw chilled atmosphere, better handle employee safety concerns

Under regulatory pressure to change the way employee concerns are handled, TVA says it is trying to thaw the often chilled atmosphere for employees to raise safety concerns at its Watts Bar nuclear plant.

TVA's top nuclear officials said Tuesday they have revamped management procedures, coached bosses on how to better treat workers and conducted surveys and employee meetings to better encourage and respond to safety issues raised by workers.

"We recognize that the Watts Bar environment, and specifically in operations, had degraded to a point where some operators felt reluctant to raise issues — we get that and we own that," Mike Balduzzzi, senior vice president at TVA, told nuclear regulators Tuesday. "We firmly believe it is necessary for our employees to feel free to raise nuclear safety issues without fear of retaliation and, in fact, they should expect to be complimented for doing so."

Despite such reassurances, however, a nuclear watchdog group questioned why TVA keeps having problems with employee concerns at its nuclear plants. During a hearing before the Nuclear Regulatory Commission's regional staff in Atlanta, the director of the Union of Concerned Scientists' Nuclear Safety Project said the corrections program outlined by TVA "is troubling in that it shows how blind TVA continues to be" in its culture and response to worker concerns about potential safety problems at nuclear plants. 

"It's disappointing for a company that has gone through a number of safety-conscious work environment problems repeatedly over a number of years can't recognize signs of those problems until someone else calls their attention to it," said David Lochbaum, a nuclear engineer who once worked at the NRC and now studies the industry for the Union of Concerned Scientists. "Since they are having trouble figuring out that they have those problems, there's not much confidence that they will be able to solve problems that are invisible to them."

Upon the plants first start-up ever, they got only up to 12% 18%, then scrammed on a turbine control valve slamming open.

It doesn’t like the NRC properly vetted the plant for power operation?

June 2, 2016: WATTS BAR NUCLEAR PLANT UNIT 2 CONSTRUCTION - NRC INTEGRATED INSPECTION REPORT 05000391/2016604

It is a risk full accident at this power level. It reminds me of Indian Point's leaking steam generator where they cooled down too quickly. It is implicated in the baffle bolt problem. Wonder what the cool down rate was?

TVA and the NRC are doing a terrible job at preparing the public for the rocky road with a new plant startup.  

Could get a lot of steam generator water slamming down the main steam lines?

What is the significance of Unit 1's annulus over pressure. Must be associated with the air filtering (SBGT) system startup.

You got a very fragile 1960s technological age reactor vessel here. With the amazing metallurgical technological advancement in recent times, you’d get a much more safer plant with current technology. Imagine the risk this is to the nuclear industry public relations if the plant is abandoned at this point. Here comes yet another nuclear industry's perfect storm?  

Right, basically upon governor valve slamming open, the steam generator level went low.  The safety ejection must have came about with a low pressurizer level. The Plant safety logic must have thought they had vessel leak. Did anyone check the underwear with all employees leaving the control room?

Wonder what the delay time between when the governor started to fail and then the scram? They must have been preparing to put the turbine on the grid.

Check out the historic power history. A full week almost :) It is a brand new 43 year old plant. First startup can be very troublesome. But this looks abnormal. Sounds like they were having trouble with the turbine or its control system. It is a very erratic startup. The power level here  is taken at midnight, once a day. It probably much more erratic than shown

Watts Bar Unit 2 begins power generation

The  newest reactor at the Watts Bar Nuclear Plant began generating electricity over the weekend.

The Tennessee Valley Authority synced Unit 2 reactor to the grid for the first time Friday and licensed operators have begun an initial test run of generation equipment.

The team is using this run to collect data to be sure generating equipment is prepared for continuous full-power operation later this summer.

"This is another major step in fully integrating Watts Bar Unit 2 as the seventh operating unit in TVA's nuclear fleet," TVA Chief Nuclear Officer Joe Grimes said in a statement. "It is rewarding to see TVA taking the lead on delivering the first new nuclear unit of the 21st century and providing safe, affordable and reliable electricity to those we serve."

The next step is full-plant testing of systems and controls at increasing reactor power levels up to 100 percent power by this summer.

Combined with Watts Bar Unit 1, the plant will supply power to roughly 1.3 million homes in the TVA service area.

May 31  0%
June 1  14%
2  14%
3  18%
4  18%
5  10%

Power Reactor	Event Number: 51982
Facility: WATTS BAR Region: 2 State: TN Unit: [ ] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: MICHAEL BOTTORFF HQ OPS Officer: DONG HWA PARK	Notification Date: 06/05/2016 Notification Time: 16:22 [ET] Event Date: 06/05/2016 Event Time: 12:27 [EDT] Last Update Date: 06/05/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): OMAR LOPEZ (R2DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
2	A/R	Y	13	Power Operation	0	Hot Standby

Event Text

REACTOR TRIP AND ECCS ACTUATION CAUSED BY TURBINE GOVERNOR VALVE FAILURE "On June 5, 2016 at 1227 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) Unit 2 was in MODE 1 at approximately 12.5% power when a safety injection actuation occurred, followed by an automatic reactor protection system (RPS) trip. Preliminary data suggests that the #1 high pressure turbine governor valve failed open causing a steam header pressure rate of decrease safety injection [SI] actuation signal. As designed, the safety injection actuation caused both trains of the shared Emergency Gas Treatment System (EGTS) to align to Unit 2, requiring WBN Unit 1 to enter Technical Specification (TS) LCO 3.0.3 at 1227 [EDT]. Also as designed, Unit 1 annulus pressure momentarily increased, causing operation personnel to enter LCO 3.6.15 Condition B at 1240 [EDT]. At 1242 [EDT], after annulus pressure normalized, operations personnel exited LCO 3.6.15. At 1245 [EDT], Operations personnel secured safety injection and Unit 2 was stabilized in MODE 3 at normal operating pressure and temperature. By 1349 [EDT], Unit 1 Operations personnel had restored both trains of EGTS to standby readiness, and exited TS 3.0.3. No primary safety barriers (RCS, containment and fuel clad) were challenged and no primary or secondary safety or relief valves actuated during the event. The Unit 2 plant trip was uncomplicated and safety equipment operated as expected. "Unit 2 is stable in Mode 3 at normal operating temperature and pressure, in normal shutdown power alignment. Unit 1 is stable in Mode 1 at 100% power. "The Senior Resident Inspector has been notified of this event." The MSIVs are shut with the steam generators (SG) discharging steam using the atmospheric dump valves. There is no primary to secondary leakage. Motor driven AFW pumps are running to maintain SG levels. RCS pressure remained above ECCS (Emergency Core Cooling System) discharge pressure

Major limitations of ultrasonic testing

Major limitations of ultrasonic testing are:

• Surface must be accessible
• Skill training is more extensive that with some other methods
• Normally requires couplant to promote sound transfer
• Surface roughness, complex geometries, small parts or exceptionally thin materials are difficult to inspect
• Coarse grained materials i.e. cast iron are difficult to inspect due to low sound transmission and high signal noise
• Linear defects oriented parallel to the sound beam go undetected
• Reference standards are required for equipment calibration
• Some kinds of noise like hissing and flow noice can throw the detector off
• austenitic steel because of large grain size is very troublesome

Junk Plant Indian Point Baffle Bolts: Shocking New News

I'll make the case today the NY grid prices are around 25 bucks per megawatt-hour. They need on the far side of 60 bucks to making any profit. In the last two two years across the board, there has been shocking reduction of grid prices across the board in the Northeast. 

My position is all bolt and baffle panel need to be replace before start-up, plus all the supports for the panals. 

Degradation of Indian Point Reactor Triggers NRC Investigation

06/01/2016 08:42 am ET | Updated 1 day ago

   Roger Witherspoon Journalist

The unprecedented degradation of critical baffle bolts in the Indian Point 2 reactor has triggered an extensive investigation by federal officials seeking to learn why the

Yep, I identified this too. The loose parts monitor is much like the UT detector. You can adjust the sensitivity of this device. If the sensitivity is set to low, then again you get false positives. You enter a multi dollar outage chasing ghost loose parts. Basically in order to save money, they bypass the loose parts monitor. They set the sensitivity so high it basically doesn't work.  

problem was so severe, why systems designed to detect loose metal objects failed, and whether the plant has a specific flaw that could compromise its ageing management program. 

The discovery in March during a routine fuel change that more than 27% of the stainless steel bolts needed to channel cooling water through active nuclear fuel rods were broken, distorted or “missing” has prompted the Nuclear Regulatory Commission to demand answers from Entergy, which owns Indian Point 2&3 in Buchanan, NY. The NRC will first review Entergy’s Licensee Event Report, due within a week, on the degradation problem and regulators insist they will not allow the plant to resume operating until critical questions are answered. 

“The analysis to be performed by Entergy regarding the bolts,” said NRC spokesman Neil Sheehan “will need to demonstrate that the baffle-former plates would remain in place even if a certain number of bolts failed.

This is a change in licencing. They need to do a 50:59.  What does plant design and licencing say as far the acceptable amount of broken baffle bolts. It is a huge reduction of margins from initial licencing. You get it, this will be a paper whipping evaluation without any actually testing.   

And those criteria will include the safety margins. We will review Entergy’s analysis and plans before deciding if the company’s proposed course of action is acceptable. Part of those evaluations will include whether it intends to assess the baffle-former bolts in Indian Point 3 in the near-term or several years from now.” 

An escalating series of problems began with an enhanced inspection of the interior of the Indian Point 2 reactor during a scheduled refueling. New York Attorney General Eric Schneiderman had insisted in a series of legal actions starting in 2007 that Entergy go beyond the visual inspection by a camera lowered into the radioactive coolant and utilize more intensive ultrasonic testing. That revealed that 227 of the 882 bolts, more than one in four, were damaged due to the intensive wear and tear in the interior. That is a failure rate never recorded in any reactor in the world. And loss of the bolts and the directed coolant could lead to a fuel meltdown.

An Expanding Problem 

But as Entergy began the slow process of removing each two-inch long, stainless steel bolt using robotic arms and cameras under 40 feet of water, there were additional developments impacting the NRC’s urgent need to find the reason for the massive failure:

Basically they can dial up or down the sensitivity of the ultrasonic detectors. If they make it too sensitive, then they will have a lot of costly false positives. This will slow down the job.  This is the most shocking news here. We need a clear explanation on the limitations with the UT devices. They have to replace all bolts and baffle panels. We have to know in all inspected plants where all the bolts UT'd. What percentage weren't UT'd I just can't believe it, The UT devices sensitivity were set to low to catch all fatal flaws. This means we need a new UT inspection at all plants whose baffle bolts were inspected. How does the NRC qualify UT devices to catch all fatal flaw and tinny flaws? How do we know the device is accurate.  

• At least four additional bolts that were not detected by the ultrasonic examination broke while work was being done on the identified, degraded bolts. Sheehan said “the ultrasonic testing is not considered to be 100% foolproof.” But its rate of inaccuracy is not known and there is no way to tell what other bolts or critical internal structures may be poised to break inside the reactor. Ultimately, Entergy found it necessary to replace an additional 50 baffle bolts, bringing the total to 277, meaning one in three critical bolts was thought to be fatally degraded. 

• Alarmed by conditions in Indian Point 2, New Jersey’s energy utility, PSEG Co., conducted an ultrasonic scan of its baffle bolts during an outage last month for their Salem 1 reactor. Both Salem 1&2 had previously passed the visual inspections via underwater camera and an ultrasonic scan was not scheduled before 2019. But the scan showed 18 baffle bolts were degraded out of 832, or 2.1%. It was the first time degraded baffle bolts had been found in Salem 1, though the percentage of worn-out bolts was in keeping with industry projections for ageing systems. It added pressure on Entergy to show a circumstance specific to Indian Point 2, and not a design flaw affecting both plants. 

I  was the first person to associate the IP baffle issues with the North Anna water jetting issues.   

• The issue of water jetting, stemming from what had been regarded as an isolated incident in 2014, has now been added to the potential problems facing Indian Point 2 which may need to be addressed in the ongoing ageing management program. In the North Anna nuclear reactor in Virginia, one of the baffles became slightly loose - though the bolts proved to be sound - and the pressurized coolant shot through the gap in an intense, cold water jet and damaged two 12-foot-long fuel rods. The water jet forced the rods to spin, and the zirconium cladding burst, sending 15 radioactive uranium fuel pellets careening throughout the reactor ( http://bit.ly/1XTuCBR ). Actual core damage can be a precursor to a meltdown since the loose fuel cannot be controlled and roaming metal parts can block the cooling system for other fuel assemblies. 

• Increasing attention is being paid to a tube break in one of the Indian Point 2 steam generators which occurred in February, 2000. The rupture sent radioactive coolant surging into the normally “clean loop” that turns the electric generating turbine on the non-nuclear side of the plant. The accident forced the first regional nuclear “alert” and activation of the emergency operations center. The record shows that in the confusion, ConEd workers—then rated by the NRC as the worst in the nation - improperly shut down the reactor by inserting too much cold water in violation of rules designed to protect the reactor and critical systems from thermal shock. The official report of the incident shows the temperature differentials during the improper shutdown were not very large ( http://bit.ly/1WROTZD ) and it was thought there would be no long-term damage. But it marked a clear deviation and threat to the integrity of the baffle bolts. There are indications Entergy researchers have sought evidence that the thermal shock was greater than initially reported. If so, Indian Point 2 had the radioactive equivalent of an incipient cancer at the time it was sold to Entergy in 2000, and it may or may not be curable. 

• Pressure is mounting on the NRC to take over the investigation. The environmental group Friends of the Earth filed an emergency petition with the agency asking the Commissioners order an investigation by an agency Augmented Inspection Team which takes over investigations when the danger of a meltdown increases by a factor of 10. Further, the Commissioners are asked to issue a Confirmatory Action Letter prohibiting the plant from reopening until a root cause analysis shows it is safe to do so, and shutting Indian Point 3 for an intensive examination. Such letters are issued if the risk of a meltdown is increased by a factor of 100. 

“This plant was built in an age when people worshipped nuclear power,” said David Freeman, former president of the New York Power Authority which built and operated Indian Point 3 until it was sold to Entergy in 2001. As head of the NYPA Freeman ordered Indian Point 3 shut for two years in the 1990s because of lax standards and frequent mishaps.

“ConEd even proposed building a nuclear reactor in downtown New York,” he added after the Friends of the Earth press conference. “They never dreamed it, in itself, is as dangerous as a bomb. But you can have a radioactive fire the likes of which you’ve never imagined. 

“We’re asking the NRC to do their job and thoroughly examine every aspect of those plants. Even if Entergy replaces all of the bolts at Indian Point 2, what is the basis for not thinking that something else is way out of whack? You just can’t take risks, and it defies common sense not to shut down Unit 3 and examine it.”

Controlling Nuclear Power 

At the core of the dispute is the ability of Entergy to accurately monitor critical systems it cannot directly see or easily access in one of the world’s most forbidding industrial environments.  

Normal air pressure is about 14 pounds per square inch (psi). The pressure in a typical home water tank is about 35 psi, which is more than enough to fight gravity and send water from the bottom to the top of the home. The water does not reach a boiling point, and one could conceivably stir the water by hand, if desired. 

The heart of a nuclear reactor is the 12-foot-long fuel rod, consisting of about 360 uranium pellets encased in a zirconium sleeve. At the North Anna plant, a pressurized water reactor like Indian Point, some 264 fuel rods are grouped into “assemblies”, each containing boron control rods to slow down or speed up the fission process. There are 157 of these assemblies strategically placed inside the huge reactor core, according to Dominion Power Co. spokesman Richard Zuercher. 

If the fission process is uncontrolled, the uranium pellets will heat up to the point where it becomes a molten slag and melts through the steel reactor. So control of the temperature around the assemblies, and the flow through the assemblies is critical to safely operating the system that is, essentially a controlled chemical fire. 

The operating temperature of a reactor is about 700 degrees Fahrenheit, and the pressure is maintained at about 2,250 psi in order to keep the 300,000 gallons of superheated water in a liquid state. It is not a jumbo home water heater, and at that enormous pressure the water in the reactor core most resembles superheated, roiling concrete.
It is critical that there be a constant flow of water to avoid buildups of hot pockets around individual fuel rods. That is the function of the baffle former bolts, which hold a series of 13-foot baffles that line the core of the reactor and surround the fuel assemblies. A series of pumps, operating at even higher pressures, force some 300,000 gallons per minute of cooler water to flow down between the baffles and the wall of the reactor and come up through a series of openings and flow through the fuel assemblies to keep them in the optimal operating temperature. 

The roiling, highly radioactive reactor water is pumped out of the reactor and into the steam generator, a massive heat exchanger featuring 3,260 U-shaped, thin steel tubes (http://bit.ly/RSxaiw ), and then flows back to the reactor. To avoid thermal stress, the NRC operating regulations dictate that the temperature difference between the “hot leg” coming out of the reactor and the “cool leg” returning to the reactor can’t exceed 72 degrees F. And when cooling down the reactor, the temperature can’t drop more than 100 degrees per hour. 

The reactor water provides the first part of a 3-step process of generating electricity. The second step involves a “clean loop” of water which is forced over the steam generator tubes containing the superhot reactor water, and quickly turns to steam. This blows out to an adjacent building and through fans turning the massive, 40-ton turbine which generates electricity. The steam then passes through a reverse heat exchanger featuring cool water from the Hudson River. The steam condenses into hot water and is routed back to the steam generator to be reheated and repeat the process. The hot river water is dumped back into the Hudson, a process that kills more than a billion fish annually. 

It is a system that is intricately designed, infinitely interwoven, and extremely efficient. There are sensors to monitor temperature changes in each fuel bundle; X-ray sensors to detect wear and tear on steam tubes; a Metal Impact Monitoring System to detect loose metal parts which can damage fuel rods or block a coolant stream from properly circulating; and a Reactor Coolant System Activity monitor to detect cracks in fuel rods. 

In 2000 this efficient system went horribly wrong. 

The Nuclear Emergency 

ConEd saved money by ignoring repairs. By Feb. 15,2000, there were more than 15,000 items on its “corrective action list” of major or minor items which had broken or malfunctioned and needed repair or replacement. The NRC”s Augmented Inspection Ream report ( http://bit.ly/1WROTZD ) of April 28, 2000 notes: “the number and duration of the equipment problems reflected weaknesses in engineering, corrective action processes, and operational support at the Station. The Licensee’s response to a number of equipment problems identified during the event reflected an acceptance of ‘working around’ rather than fixing the problem.” 

Indeed, workers in one area had to walk around holding a broom in front of them to detect thin jets of steam leaking from high pressure pipes that were hard to see but could cut through a person like a laser. If the broom suddenly shredded the workers knew where a leak was.

Emails To NY DPS Concerning Junk Plant Indian Point

Showing the DPS my credentials. 

1)

mailto:steamshovel2002@yahoo.com 

To: brandon.goodrich@dps.ny.gov

Mr. Brandon.

It was a nice talk with you today. I was a little nervous talking to you and a bit hyped up. 

Here are my talking points concerning our conversation. I was very please how quickly the DUP responded to my concern. You never know what information they will release from this...how big it will be. This applies to many other plants. We are talking about hundred of millions of dollars. 

"Indian Point: It's Core Components Highly Irradiated Differential Of Expansion, Stupid" 

Feb 04, 2016Victor M. McCreeExecutive Director for OperationsU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001 

SUBJECT: Runaway Main Steam Safety Valve (MSSV) Tech Spec lift Setting Failures at Indian Point since 2009. 

Dear Mr. McCree, 

Why is the NRC sleeping at the switch? I request a 2.206 petition at Indian Point. 

Why has there been an astonishing increase in Main Steam Safety Valve Setpoint drift testing failures beginning in 2009? There seems to be none before 2009. Why did these failure begin in 2009…considering all corrective action and processes at the site. It continues today unabated. 

It is like they caught some mysterious disease and there is no cure. These valves being opened cool the core and control steam generator pressure.

Shearson Harris mysterious began having similar failures. Two valves fail in 2013 and in the most current LER five valves failed. They have a similar clean record before 2013. Why has this mysterious problem popped up from nowhere and the failure rate has been drastically increased as time goes on. Why can’t these two plants get control of this problem? 

There seems to be vibrations issues. An odd assortment of internal parts degradations seems to be the culprit. What new has changed?

What new has changed in both these plants to cause the internal parts to fail? I am particularly irked by pressurizer safety internal coil or spring being stretched. It reminds of the test stand damage seen a Pilgrim before installation. It is known, when most of these guys fails lift setpoint and out of federal tolerance, they just adjust the setting to acceptable Tech Spec limits. They don’t do any further inspection or proactive or preventative maintenance. No doubt this in happens at many plants. 

What is the concern to me the most, is there is no new “Information notices” on SRVs, MSSVs and pressurizer valves degradation or failure? We know the LERs really only capture-report on a small proportion of the valve problems. Your industry wide trends and information is not being kept up to date on these components. Your NRC notices are grossly out of date (most of them a decade out of date or more) on the components and these licensees are constantly referencing these grossly out of date notices in the current LERs.

You would think a nuclear plant having problems like this would go the market to get a better reliable design and up to date components new. Is Indian Point have problems with component replacement and can they buy sufficient valve new internals? 

1) Request an immediate special inspection on the magnitude of the valves involved anthe constant reoccurrence of these failures. These guys have constantly said in each LER the problem is solved. If they can’t keep these valves operable, then both plants shouldn’t be operating. I am sure NY State would agree with me. 

2) Please create a new NRC Information notice on the failures of the valve and the organization. 

3) I request Indian Point immediately eradicate any problems with their MSSV up to and Including a shutdown. They are controlled by technical specifications.
.The involved LERs at Indian Point below. Why did it start in 2009…what has changed? 

LER 2015-002-01LER 2012-003-02LER 2012-005-01LER 2011-004-00LER 2010-002-00LER 2009-002-00 

I suspect they will have more MSSV unreliability issues. 

Will you let me know if there are new LERS? 

The NRC response to me: 

http://steamshovel2002.blogspot.com/2016/03/nrc-2206-response-junk-safety-valves-at.html 

I was heavily involved with the Pilgrim’s unsafe Safety Relief Valves, the 2015 plant trip…the resultant special inspection and Entergy shutting the plant in 2019.  I was on the SRV problem in 2013 and it is in the record. 

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

This is ridiculous. 

The aim of my game is to fix problems early and anticipate then, not fix big costly problems re-activity.

Please read my comments on the NRC blog:  
An Outage Twist: Degraded bolts at New York Nuclear Plant Warrant Attention

An Outage Twist: Degraded bolts at New York Nuclear Plant Warrant Attention Neil Sheehan Public Affairs Officer Region I When the Indian Point Unit 2 nuclear power plant entered a refuelin...

Did I mention I was a licensed operator at Vermont Yankee? 

North Anna had Baffle Jetting in 2014. The edges of the baffle plates widened. Basically caused by the same phenomena as Indian Point. The jetting damaged fuel pins leading to a large release radioactively in the coolant, but no offsite release. Before the fuel pins leaks, they have wear marks on the pins from slowly spinning in place by the water jet. This can be seen under close inspection. 

I have already given you way too much information. 

I apologized  to you for that.


Sincerely,
Mike MulliganHinsdale, NHHome 16033368320Cell 16032094206
Call or email me anytime?

2)

To: brandon.goodrich@dps.ny.gov 

May 14 at 9:57 AM 

I went over the last SG report coming from IP 2. They say they discovered some foreign debris in the SG, but don't accurately (nothing) detail what the new debris is on the steam side. This is highly suspicious. Most plants detail the discovered debris in this report. You need to know if components are spitting off metallic parts from the rest of the plant and threatening the steam generator tubes. With debris, you need to hunt down where the metal came from. It is important to put it in a trend, is the threat to the tubes increasing or decreasing over time. Read up on their last SG tube rupture event at the site. Did baffle debris end up in the steam side of the SG? Was this a cover-up? This is a regulatory failure, in that the NRC should have forced IP to publically document in detail all SG debris in the refueling outage tech spec mandated report.

You need get the NRC to list and describe all SG debris in their upcoming inspection report for the last ten years. I believe its in IP's secret internal reports? The NRC knows where its at.   
Mike MulliganHinsdale, NH

3)

To : brandon.goodrich@dps.ny.gov
May 4 at 5:51 PM
How are these guys at IP? I know IP has intensive relicensing contention core inspections. See attachment. It's the Cook root cause on the degradation. I hope the state is going to pry the RCA on the baffle bolt issue from IP.

Clevis Insert Bolts
" Clevis Dowel Pins
* LRSS Lug Weld
* BMI Nozzles and Welds

Can you even imagine the  plant vulnerability with the worst case degradation on the above components and the worst case degradation of the baffle plates? I like to see the risk perspective numbers on that baby.

It is a accident never conceived by anyone before. Sound familiar? Thank you UCS!  
 

A Huge Intensification of the Nuclear Industry's Historic Financial Problems.

We are waiting for somebody to say boo...where the utilities start to throw (cascade)overboard the regulated plants. These regulated plants are highly susceptible to the politicians. These guys are grossly not profitable. 

I think we are unprecedented times with these financial troubles silencing their employees.   

Basically three plants going dead. 

Exelon Makes Good on Threat—Quad Cities and Clinton Nuclear Plants to Close

06/02/2016 | Aaron Larson

Given the lack of progress on Illinois energy legislation, Exelon Corp. announced on June 2 that it would begin taking steps to permanently shut down its Quad Cities and Clinton nuclear power plants. 

In a statement, the company said Clinton would close on June 1, 2017, and that Quad Cities would follow exactly one year later. The two facilities have long been rumored to be on the chopping block due to economic struggles. 

“This is an extremely difficult day for the 1,500 employees who operate these plants safely and reliably every day, and the communities that depend on them for support,” said Chris Crane, Exelon president and CEO. “We have worked for several years to find a sustainable path forward in consultation with federal regulators, market operators, state policymakers, plant community leaders, labor and business leaders, as well as environmental groups and other stakeholders. Unfortunately, legislation was not passed, and now we are forced to retire the plants.” 

The two stations are said to have lost a combined $800 million during the past seven years, despite being two of Exelon’s best-performing plants. The news comes on the heels of the Clinton station setting a new U.S. record for the shortest refueling outage for a boiling water reactor. It returned to full power on May 30 after only 11 days offline (Browns Ferry Unit 3 had held the record previously at 14.5 days). 

Although the Illinois legislative session has not ended, Exelon said that the path forward for the “Next Generation Energy Plan” legislation—which included a “Zero Emission Standard” that would make Illinois one of the first states to recognize the zero-carbon benefits of nuclear power—is not clear. As a result, Exelon has begun taking necessary steps to shut down the two nuclear plants. The steps include:

• Making permanent shutdown notifications to the Nuclear Regulatory Commission within 30 days.
• Terminating capital investment projects required for long-term operation of Clinton and Quad Cities.
• Immediately taking one-time charges of $150 million to $200 million for 2016, and accelerating approximately $2 billion in depreciation and amortization through the announced shutdown dates.
• Cancelling fuel purchases and outage planning.

The company said it has and will continue to brief the Governor’s office, legislative leaders, the Illinois Commerce Commission, Illinois Environmental Protection Agency, Illinois Power Agency, other relevant state agencies, and host community leaders, on developments as it executes the shutdown plan. 

“We are deeply grateful for the broad support we received from policymakers, plant community leaders, labor officials and business leaders who spoke out to help keep the plants operating,” Crane said. “We wish the outcome would have been different, and we appreciate all the support this effort has received