2.206: Request Emergency Vessel Inspection Within Six Months On Most USA Nuclear Plants?

(final version)  

March 25, 2014

Mr. Mark A. Satorius

Executive Director for Operations

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Dear Mr. Satorius:

SUBJECT: 10 CFR 2.206 Petition requesting emergency ultrasonic inspection test or best available flaw detection technology for USA reactor plants similar to the thousands of cracks discovered in Belgium nuclear power plants.  

“The discovery of the cracks in the reactors “could be a problem for the entire global nuclear industry,” says Jan Bens, general director of the Belgian Federal Agency for Nuclear Control (FANC), speaking on.”

This is a 2.206 on all US nuclear plants. Please place on all plant dockets especially Vermont Yankee and Kewaunee Power Station.

I don’t even think a LOCA in a vessel is even a consideration in plant licensing and this could be our Fukushima style accident in the USA.

Doel 3 and Tihange 2

“Cracks appear in the nuclear industry

Cracks discovered in the walls of Belgian nuclear reactors are causing unease among experts. The reason: a previously unknown phenomenon – material fatigue. There are fears that many more reactors could be affected.

Several thousand cracks have been discovered by corrosion experts in the pressure vessels of two reactors at the Belgian nuclear power stations Doel 3 and Tihange 2. Caused by a previously unknown phenomenon, material fatigue, it is feared the finding could have implications outside of Belgium.

The discovery of the cracks in the reactors “could be a problem for the entire global nuclear industry,” says Jan Bens, general director of the Belgian Federal Agency for Nuclear Control (FANC), speaking on.

Examination of nuclear reactors demanded

Of most concern are the cracks that have been found in the walls of pressure vessels, the part of reactors where the highly radioactive chain reaction takes place. During such a process, the vessel is under extreme stress and instability caused by the cracks could cause a potentially catastrophic release of radioactive contamination.

It is already known that pressure vessels can become fatigued as a result of stress caused by pressure, temperature and radioactive materials. But the Belgian Nuclear Research Center in Mol has only just found out “that the material is mechanically weakened through radiation much more than previously thought,” says Heinz Smital, nuclear physicist and expert at Greenpeace.

Nuclear material corrosion expert Walter Bogaerts, of Belgium’s University of Leuven says that corrosion factors have until now been “underestimated”, globally. He adds: “I would be really surprised if it had not also occurred elsewhere.”

Reactors could be shut down

Digby MacDonald, an expert in corrosion at the University of California, Berkley, analyzed the cracks together with Bogaerts and has advised nuclear reactor operators and government regulators that they should use ultrasound equipment to carefully examine reactors for cracks. “All reactor operators should be require under the leadership of regulatory authorities,” says MacDonald. He adds that the results of such detailed investigations “could be insignificant, or so strong that all the reactors must be shut down.”

According to nuclear experts, hydrogen from the reactor can penetrate the reactor wall and there in the steel increase the interior pressure causing small bubble and cracks from just a few millimeters in size “up to seven centimeters”, says Smital.

Using special ultrasound equipment, experts discovered 13,047 cracks in total in the Belgian reactor Doel 3 and 3,149 in Tihange 2. The reactors have been shut down, as a result. Whether they will once again be connected to the network is, as of yet, unclear.

Danger for the nuclear industry

The appearance of the cracks as a result of material fatigue has caused a tide of reaction. Safety checks are being demanded all over world and “could lead to a wave of reactor closures”, says Smital.

Greenpeace successfully sued the Belgian nuclear authority FANC in January following the publication of the detailed investigative documents. “It's a very delicate matter and could indeed have a huge impact on the whole nuclear industry,” says Smital. Greenpeace is demanding that all reactors worldwide are closely examined.

The German Environment Ministry has also reacted and is seeking to have immediate contact with the Belgian authorities to see whether the findings could be applicable to German reactors.

But according to Greenpeace, the Belgian findings confirm the growing threat posed by old nuclear power plants. The world’s reactors now have an average age of 29 years. “That is no longer state-of-the-art, which can be dangerous, even when you upgrade,” says Smital. “What are now needed are scenarios for a shut down of plants. Every country needs a get-out plan.””

1) I request exigent and immediate full scale ultrasonic inspections similar or with better technology (as was done on the Belgian Doel 3 and Tihange 2 nuclear plant discovering 13,047 recent cracks) on the following US nuclear power plant:

BWR: Vermont Yankee (June 1974)

PWR: Kewaunee Power Station (Nov 1972) (preferred to be first)

Both these plants are permanently shut down so it won’t disrupt plant operations. An emergency ultrasonic inspection could occur very quickly based on the plants permanently shut down status and the Kewaunee plant is similar to the Belgium PWRs and its age. The nuclear fuel is out of the core and the surroundings would have lower doses. You could quickly strip off the vessel insulation and you wouldn't need to replace it for lower doses. You wouldn’t have to replace the core interior components either?

2) I request large bore holes samples be cut out of both vessel at the worst place similar to the “Davis Besse hole in the reactor head” event. Then transport the vessel specimens to a respected metallurgic laboratory for comprehensive off site testing.

3) Request an immediate NRC report and public meeting on the vulnerabilities with US reactor cracking and these mysterious weakened vessels. 

4) If distressing and unsafe results are discovered at Vermont Yankee or Kewaunee, I request that within six months all USA plants be ultrasonically tested or better technology.

5) How has the average concentration of hydrogen in the coolant changed over the recent decades? Would an increasing concentration of hydrogen in the coolant lead to more hydrogen ions getting injected into the vessel iron?

6) Does noble chemistry increase or decrease this kind of corrosion? 

7) Are there other chemicals added to the coolant that could make this kind corrosion worst? 

8) What are they talking about here:

  

"However, as Belgian continues to debate the fate of the reactors, prolonged studies on the steel used in the construction of the reactors revealed unprecedented embrittlement – unusual swelling – that can compromise the integrity of the plant and possibly cause ruptures, spewing dangerous radioactive material equivalent to an atomic bomb."

9) I understand all US nuclear plants have coupons and I consider them irrelevant to this problem.

10) Request the NRC coordinate with the Belgian Federal Agency for Nuclear Control (FANC).

11) Request detailed inspection on the condition of the reactor cladding and a explanation of any defects. 

Sincerely,

Mike Mulligan

Hinsdale NH

1-603-336-8320

steamshovel2002@yahoo.com

Storm Juno LOOP and SRV Malfunction...the Special Inspection.

Works in progress...I've be adding to this in the next week. Heading towards a 2.206.

Update: March 20, 2015 

Yesterday i called the Hatch nuclear plant senior resident asking him about all their recent failures of the model 0867F SRVs. Could just leave a recording. You know you are in Siberia when the regional public relations person calls you back instead of the resident inspector.

He didn't have much information about the Hatch SRV events…but he did say something particular. There is no doubt he called region I officials and the higher level officials were coordinating the response. What the message the NRC wanted me to relay to the public:

The failed SRV at Pilgrim had 57 cycles on it while a similar Hatch SRV  has 3 to 4 cycles on it. Hatch operates their SRV's fundamentally different than Pilgrim.

I tried to probe this guy (he was a nice guy) asking him how Hatch operates their SRVs different than Pilgrim. I told him I was a license operator at Vermont Yankee. He responded back to me, Pilgrim uses their SRVs to control pressure while up at power while Hatch does not. You never know when these guys are playing dumb or are really dumb? Usually all NRC employees, no matter what the stripe, are particularly intelligent. I asked him twice what he meant by that, the response never changed. 

If I had to translate what he said, I think he was trying to say during start-ups and particularly shutdowns and hard shutdowns, Pilgrim reverts as a normal operation path, especially when the main stream lines are closed, into using their SRVs as the means of cooling the core and pressure control in the vessel. 

Is he saying all their shutdowns and hard shutdowns are causing damage to Pilgrims SRVs? Why are the SRV designs so delicate? 

Typically in the worst accident possible (DBA), they cycle the SRV valves between 300 and 500 times in one accident…

I asked region I senior public affair official to respond to this in a phone message (Mrs. Srenci)

Did the NRC release this for the public meeting tonight?

PNO

Part 21 Event Number: 50900

Only the 0867F is under investigation.

The root cause of the potential test induced defect has not yet been confirmed as of the date of this report.

One of the four installed MS-SRVs may not have fully opened.

As-found steam testing of the affected MS-SRV did not duplicate this failure; the valve opened on demand.  

We are working with all three (4) sites to identify appropriate precautions.

Rep Org: CURTISS WRIGHT FLOW CONTROL CO.
Licensee: CURTISS WRIGHT FLOW CONTROL CO.
Region: 1
City: EAST FARMINGDALE State: NY
County:
License #:
Agreement: Y
Docket:
NRC Notified By: JOHN DeBONIS
HQ OPS Officer: STEVE SANDIN Notification Date: 03/17/2015
Notification Time: 09:59 [ET]
Event Date: 03/17/2015
Event Time: [EDT]
Last Update Date: 03/17/2015
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21(a)(2) - INTERIM EVAL OF DEVIATION
Person (Organization):
GLENN DENTEL (R1DO)
BINOY DESAI (R2DO)
PART 21/50.55 REACT (EMAI)
Event Text
INTERIM PART 21 REPORT - POTENTIAL TEST INDUCED DEFECT IN A 0867F MAIN STEAM SAFETY RELIEF VALVES

The following report was received from Curtiss - Wright via email:

"This letter provides interim notification of a potential test induced defect in a 0867F Series Main Steam Safety Relief Valves (MS-SRVs) manufactured and supplied by Target Rock (TR). The information required for this notification is provided below:

"(i) Name and address of the individual or individuals informing the Commission.

William Brunet
Director of Quality Assurance
James White
General Manager
Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation
1966E Broadhollow Road
East Farmingdale, NY 11735

"(ii) Identification of the basic component supplied for such facility or such activity within the United States which may fail to comply or contains a potential defect.

Target Rock 0867F Series of Main Steam-Safety Relief Valves Manufactured by Target Rock. This is a 3-stage piloted valve consisting of a main valve (the 'Main') with an actuator mounted to it (the 'Topworks'). The 0867F is the latest generation of the 67F line of MS-SRVs, including the original 3-Stage and 2-Stage designs, and this product line has over 40 years of plant operational experience. Only the 0867F is under investigation. This is due to the differences between the 0867F design and the other designs.

"(iii) Identification of the firm supplying the basic component which fails to comply or contains a defect.

Target Rock, Business Unit of Curtiss-Wright Flow Control Corporation
1966E Broadhollow Road
East Farmingdale, NY 11735

"(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.

As we understand it, the Pilgrim Station recently manually opened the Target Rock Main Steam Safety Relief Valves (MS-SRVs) as part of cooling down the reactor following a loss of offsite power. One of the four installed MS-SRVs may not have fully opened. As-found steam testing of the affected MS-SRV did not duplicate this failure; the valve opened on demand. However, the valve did not re-close as expected. Internal inspections found damaged parts in the main stage subassembly that could potentially affect the ability of the MS-SRV to operate as designed.

We are investigating potential root causes for this damage. However, we are still unable to determine if a specific defect exists. GE SIL-196, Supplement 17 determined Main Spring relaxation was caused by 'extreme dynamics encountered during limited flow testing . Valve dynamics under full flow conditions (i.e. discharge not gagged) are much less severe than those under limited flow conditions.' These extreme dynamics, under limited flow test conditions, are the focus of our investigation. Specific areas of investigation include;

a) Testing of materials to verify they are consistent with our material specifications,
b) evaluation of differences between the 0867F and earlier designs, and
c) evaluation of the differences between different limited flow test loop configurations and test procedures

"(v) The date on which the information of such defect or failure to comply was obtained.

The Pilgrim event occurred on January 27, 2015. As-found testing occurred on February 2, 2015.

"(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured, or being manufactured for one or more facilities or activities subject to the regulations in this part.

While we have yet to determine if a specific defect exists, 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

"(vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

The root cause of the potential test induced defect has not yet been confirmed as of the date of this report. Therefore, no specific corrective actions have been initiated. Target Rock Problem Report 080 will document the corrective actions when they are determined and complete the 10 CFR Part 21 evaluation of the potential test induced defect. This determination will be based on further mechanical and material evaluations. TR anticipates completing these evaluations within 45 days; however, in the event the evaluations are not completed, TR will forward another interim report within 45 days.

"(viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees.

We are working with all three (4) sites to identify appropriate precautions.

"(ix) In the case of an early site permit, the entities to whom an early site permit was transferred.
Not applicable.

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

SRV-3B Safety Relief Valve Declared Inoperable Dueto Leakage and Setpoint Drift

Licensee Event Report 2013-002-01

Event date 01 20 2013

LER: 2013 002 01

Report Date 1 31 2014

On Sunday January 20, 2013, at 2050 hours with the reactor at 100% core thermal power (RMSS in RUN), PNPS declared SRV-3B inoperable and entered Technical Specification (TS) 3.6.D.2 requiring an orderly reactor shutdown such that reactor coolant pressure is less than 104 psig within 24 hours. On Monday January 21, 2013, at 1300 hours (16 hrs and 10 minutes) reactor coolant pressure was lowered to less than 104 psig. SRV-3B had been declared inoperable consistent with PNPS procedures that state an SRV is inoperable if the first stage pilot thermocouple temperature is 350 F below its baseline temperature. This LER Supplement provides the determination of cause for the leakage. The cause of the SRV leakage was that the natural frequency of the pilot assembly was close to a resonant frequency of the valve assembly when installed on the PNPS main steam line, that had failed to be considered in the design of the SRV. A contributing cause was wear and looseness of parts in the main stage of RV-203-3B.

The reactor was depressurized and a new pilot valve assembly was installed on SRV-3B. On January 22, 2013, at 1015 hours reactor restart was commenced. On January 24, 2013 at 0312 hours 100% core thermal power was achieved.

This LER also reports the as-found setpoint of one SRV pilot valve tested was less than the minimum pressure required by TS 3.6.D.1.

This event had no impact on the health and/or safety of the public.

BACKGROUND:

As background, the pressure relief system includes four (4) SRVs and two (2) spring safety valves (SSVs). During Refueling Outage (RFO-1 8), in April/May, 2011, the four SRVs were replaced with Target Rock Model 0867F 3-stage SRVs. The SRVs discharge through their individual discharge piping, terminating below the minimum suppression pool (torus) water level. The four SRVs are installed on the main steam piping in containment between the reactor pressure vessel and the flow restrictors.

The 3-stage SRV contains a pilot (also called the first stage), a second stage, a main stage, and an air operator.

To monitor these valves for leakage, Pilgrim installed thermocouples at the pilot (first stage), at the second stage, on the tailpipe near the valve (4.5' to 6' away), on the tailpipe far from the valve (-20' away) and at the pilot bellows. Procedure 2.2.23, "Automatic Depressurization System", provides guidance for interpreting the thermocouple data and determining valve operability based in part on testing performed by Target Rock.

Subsequent to installation in RFO-1 8 and prior to this event, Pilgrim experienced minor second stage pilot valve leakage from SRV RV-203-3C on May 18, 2011 and November 25, 2011. Also, on December 26, 2011, SRV RV- 203-3D first stage pilot valve experienced leakage while operating at full power. The SRV was declared inoperable and the plant was shutdown on December 26, 2011 in accordance with TS 3.6.D.2 and RV-203-3C was replaced entirely, and the RV-203-3D pilot assembly was replaced (LER 2011-007-00).

EVENT DESCRIPTION:

On Sunday January 20, 2013, at 2050 hours with the reactor at 100% core thermal power (RMSS in RUN), PNPS declared SRV-3B inoperable and entered Technical Specification (TS) 3.6.D.2 requiring the initiation of an orderly reactor shutdown such that reactor coolant pressure is less than 104 psig within 24 hours. On Monday January 21, 2013, at 1300 hours (16 hrs and 10 minutes) reactor coolant pressure was less than 104 psig. SRV-3B had been declared inoperable consistent with PNPS procedures that state an SRV is inoperable if the pilot stage thermocouple temperature is 350 F below its baseline temperature.

While at full power, indication of a steam leak across the first stage pilot of RV-203-3B was identified. The leakage was evaluated and in accordance with criteria specified in procedure 2.2.23, specifically, if the pilot stage thermocouple temperature is 35 degrees F below its baseline temperature (with a smaller decrease at the second stage thermocouple) and cannot be explained by a corresponding downpower, the SRV is inoperable. The safety relief valve was subsequently declared inoperable and the Limiting Condition for Operation (LCO) for Technical Specification (TS) 3.6.D.2 was entered. Per TS 3.6.D.2 the plant was shutdown and reactor coolant pressure was below 104 psig within 24 hours.

CAUSE:

The SRVs were purchased new, installed, and tested for the first time in April/May 2011 during RFO-1 8.

 Following an extensive investigation, it was determined that the cause of the SRV leakage was that the natural frequency of the pilot assembly was close to a resonant frequency of the valve assembly when installed on the PNPS main steam line. This was not considered in the Entergy specification or the Target Rock design of the

EXTENT OF CONDITION:

This condition potentially applies to all four three stage SRVs that were installed in RFO 18. During Cycle 19 operation, Pilgrim has observed leakage from RV-203-3B, 3C, and 3D. 

·         On May 18, 2011 and November 25, 2011, SRV RV 203-3C second stage pilot valve minor leakage was observed. This condition did not cause inoperability of the valve. SRV RV-203-3C was replaced during the December 26, 2011 shutdown.

·         On December 26, 2011, SRV, RV-203-3D first stage pilot valve experienced leakage that exceeded the operability criteria while operating at full power. The plant was shut down as required by TS 3.6.D.2, RV 203-3C and 3D were repaired and the plant returned to full power operation. The cause of the pilot leakage was later determined to be a combination of the natural frequency issue and weakening of the pilot bellows spring. This bellows spring had a through wall failure during testing at an offsite test facility in March 2013. This failure was the subject of a Target Rock 10 CFR, Part 21 (Reference 1).

·         On January 20, 2013, Pilgrim experienced the event described in this Licensee Event Report, first stage pilot valve leakage of SRV, RV-203-3B. The plant was shutdown as required by TS 3.6.D.2. The pilot valve was replaced with a refurbished pilot and the plant was returned to full power operation.

·         On February 3, 2013, RV-203-3B first stage pilot valve leakage was identified while at full power. Reactor power was lowered to 80% and at 1000 psig pressure, the pilot was reseated. An Operability Determination with a compensatory measure was implemented to maintain the reactor power at 80% and reactor pressure at 1000 psig. An Operations Decision Making Issue (ODMI) was implemented to monitor and take corrective actions. During the forced outage on February 8, 2013, caused by a loss of offsite power due to a major winter storm, RV-203-3B first stage pilot valve was replaced with a new pilot valve and the plant was returned to power operation. The cause of the pilot leakage was determined to be a combination of the natural frequency issue and weakening of the pilot bellows spring. This bellows spring had a through wall failure during testing at an offsite test facility in March 2013. This failure was the subject of a Target Rock 10 CFR, Part 21 (Reference 1).

The removed RV-203-3B pilot valve was sent to Wyle Laboratory for testing.

As-found test results for the SRV, RV 203-3B pilot valve were:

Pilot S/N SRV Position As-Found Deviation

23 RV-203-3B 1112 psig (-)3.8%

Technical Specification 3.6.D.1 requires the as-found setpoint to be within 1155±34.6 psig (1120.4 psig to 1189.6 psig). The as-found setpoint was less than the minimum pressure specification required by TS 3.6.D.1. This test result was entered into the corrective action program as a separate event, and is included in this LER since the condition was discovered within 60 days from the initial discovery of pilot leakage. Accordingly, this as-found value being out of Technical Specification setpoint is reported in this LER pursuant to 10 CFR 50.73(a)(2)(i)(B).

·         The third pilot on RV-203-3B began leaking on February 26, 2013. Leakage was controlled by reducing power and pressure per the ODMI. This pilot was replaced during the Spring 2013 RFO. The cause of the pilot leakage was that the pilot assembly had a natural frequency that was close to a resonant frequency of the valve assembly when installed on the PNPS main steam line.

CORRECTIVE ACTIONS:

The following corrective actions were taken to address this event related to leaking RV-203-3B:

·         The SRV-3B pilot was replaced with a refurbished and tested pilot.

·         PNPS Procedure was revised to reduce reactor power and pressure to stop leakage per an ODMI as described in "Extent of Condition."

The following corrective actions are being taken to address the results of review of Extent Conditions:

·         To minimize the possibility of further pilot leaks, all currently installed pilots (and replacements if necessary until the long term corrective action can be taken) have been set at the high end of their allowed set pressure band.

·         The recommendations of the Target Rock 10 CFR, Part 21 are being followed.

·         The only PNPS pilot with a bellows spring from the same material and heat treatment certifications as the failed bellows was removed from the plant. Detailed metallurgical analysis did not identify any intergranular cracks such as those identified in the failed bellows.

·         PNPS has ordered new pilot assemblies with enhancements designed by Target Rock to raise the natural frequency of the pilot and make it more resistant to steam system vibration (References 2 and 3).

These pilots include the bellows replacement recommended by the 10 CFR, Part 21. PNPS plans to install these pilots during the spring of 2015 RFO.

SAFETY CONSEQUENCES:

The leaking SRV pilot valves and the plant shutdown to repair the SRV in accordance with Technical Specification 3.6.D.2 posed no threat to the public health and safety.

All leakage from the valve was collected in plant systems, the suppression pool (torus), and processed in accordance with normal station practices.

Pilgrim has installed temperature monitoring to provide sufficient indication of SRV leakage to ensure that timely actions can be taken to ensure that the plant is maintained in a safe condition. Procedure 2.2.23 provides the instructions and guidance for interpreting and responding to SRV temperature indications. Based on these instructions, the plant was shutdown. The SRV would have been able to respond if needed to meet its core cooling

or reactor pressure vessel over protection functions. As a result, the plant safety was maintained. The risk of operating with a leaking SRV is characterized by an increased chance of having an inadvertently opened SRV with increased chance of that valve failing to reclose.

Assuming the plant operated for 24 hours with this condition, this results in a change in core damage frequency of less than 1.OE-7. The impact of setpoint drift (0.8% below the 3% tolerance) is considered to be bounded by delta change in core damage frequency of less than 1.OE-7.

PREVIOUS EVENTS:

Prior to Cycle 19, there were no leakage or setpoint drifts occurrences with three stage safety relief valves since the new design was installed in April/May, 2011, during Refueling Outage 18 for all four safety relief valves.

During Cycle 19, Pilgrim observed minor leakage from the second stage pilot valve of RV-203-3C. Also, first stage pilot valve leakage was observed from RV-203-3D which was confirmed, plant was shutdown as required by TS 3.6.D.2, and first stage pilot valve was replaced. This event is described in LER 2011-007-00. During the outage for RV-203-3D, the entire RV 203-30 was replaced with a new valve assembly.

The industry has experienced numerous instances where SRV leakage has occurred at other plants with other Target Rock Model three stage safety relief valves.

OE33766 - Three Stage Safety Relief Valve Pilot Leakage just below Normal Operating Pressure - Plant Hatch. The plant Hatch installed the same model 3-stage SRVs in Unit 2 in April 2011. Hatch experienced numerous pilot leaks during 2011. On some occasions, leakage was reduced by power and/or pressure reductions. Hatch Unit 2 had some success through power and/or pressure reductions and operating for several months after reseating the first stage pilot valve through power and/or pressure reductions.

OE26394 & OE26892 - Planned Shutdown due to a three stage Safety Relief Valve Leak - Peach Bottom Unit 3

OE32805 - Safety Relief Valve Temperature Phenomenon – Fitzpatrick

OE34730 - Target Rock 3 Stage Main Steam SRV Bore to Seat Misalignment - Limerick 2

OE19219 - Plant Shutdown Due to Increasing Tailpipe Temperature - Duane Arnold

REFERENCES:

1. Target Rock Letter NID#13307, "10 CFR Part 21 Report, Notification of a Defect, Bellows Failure," June 17, 2013.

2. Target Rock Technical Evaluation of Replacement Items TERI 075, "Technical Evaluation of Pilot Assembly 304095-1 Replacing Pilot Assembly 303977-1 for 0867F-001," Target Rock, Revision A, January 14, 2013.

3. Target Rock Letter SRP1 3003, "Enhancements to Primary Pilot Design," Target Rock, January 21, 2013.

4. Condition Report CR-PNP-2013-0378, Safety Relief Valve RV-203-3B, Pilot Leakage.

Pilgrim Juno Scram Linked to History of Prior Failures

Five years earlier, a nor’easter in December 2008 caused a loss of offsite power at Pilgrim, which was accompanied by switchyard flashovers and then an unplanned scram. Entergy was supposed to have determined the root cause and corrected the switchyard flashover problem back in 2008 and again in 2013. Obviously they hadn’t. Instead, in 2013 Entergy had just stored the failed insulator in a warehouse, where the NRC inspectors found it 21 months later.

OF NUCLEAR INTEREST: Pilgrim Juno scram linked to history of prior failures 

By Bill Maurer and Meg Sheehan Posted Mar. 5, 2015 at 2:00 PM Updated at 4:02 PM

On Tuesday, Jan. 27, when winter storm Juno hit Entergy’s Pilgrim Nuclear Power Station (PNPS) in Plymouth, it caused an emergency “scram,” also called a “reactor trip” and more simply known as an unplanned shutdown. Unplanned shutdowns present a risk to public safety, especially when efforts to control reactor temperature and pressure during an unplanned shutdown are complicated by multiple critical equipment failures, as was the case at PNPS during Juno. Juno knocked out Pilgrim for 11 days while Entergy was making repairs to failed equipment.
Pilgrim was no sooner coming back online when winter storm Neptune hit on Valentine’s Day. This time, Entergy shut down Pilgrim as a “precautionary” measure – an explicit acknowledgement that public safety would be at risk if there was another emergency at Pilgrim. Pilgrim was offline for three days, taking an additional five days during restart to reach full power. The U.S. Nuclear Regulatory Commission (NRC) spokesman Neil Sheehan reported that during the restart Entergy was “working through some non-safety related, balance-of-plant equipment problems. These are new issues and not problems from the 1/27 storm. Such issues are not unusual following two shutdowns and start-ups in a short period of time.” 

Entergy’s Jan. 27 Pilgrim scram raised alarms not only with the public, but at the NRC too. The agency sent a six-member special inspection team to PNPS for a week to figure out what went wrong this time. Their report is due by the end of March. 

It’s no surprise that Pilgrim’s aging facilities could not handle Juno and were forced into shut down. In 2013, Pilgrim had four emergency scrams, which put it in the “degraded category” under NRC rules. In the fall of 2014, the NRC investigated the degraded conditions at Pilgrim, and one day before Juno, on Jan. 26, issued a report. The NRC found that Pilgrim failed its inspection because Entergy had not fixed all the problems that caused the four scrams in 2013. One thing the NRC found that Entergy had not addressed was how to handle severe weather events like Juno. Juno proved the NRC right – PNPS was forced to shut down.

The NRC’s Jan. 26 report also found that in 2013 Entergy had failed to deal with a recurrent switchyard performance failure called a “flashover,” which is electricity arcing between two points causing a fault. A failed insulator in the Pilgrim switchyard was identified as a contributing cause of the flashover during a storm in 2013. Five years earlier, a nor’easter in December 2008 caused a loss of offsite power at Pilgrim, which was accompanied by switchyard flashovers and then an unplanned scram. Entergy was supposed to have determined the root cause and corrected the switchyard flashover problem back in 2008 and again in 2013. Obviously they hadn’t. Instead, in 2013 Entergy had just stored the failed insulator in a warehouse, where the NRC inspectors found it 21 months later. Entergy had deferred the funding for the investigation of the failed insulator 11 times, causing the NRC to determine that Entergy “failed to investigate a deficient condition.” 

Since the Blizzard of 1978, switchyard flashovers at Pilgrim have been a recurrent equipment performance failure. Now, Pilgrim has had at least eight unplanned scrams all provoked by Nor’easters delivering blizzard conditions. Pilgrim’s switchyard equipment is located outside, totally exposed to wind-driven salt air, spray, rain, ice and snow. The Juno 2015 loss of offsite power, flashovers and unplanned scram was predictably just one more time.

The NRC’s Jan. 26 report speaks directly to the predictability of recurrent failures during nor’easters: “Inspectors determined that the inadequate guidance for pre-storm actions represented a condition adverse to quality that was reasonably within Entergy’s ability to identify and correct by execution of corrective actions identified in the RCE” (Root Cause Evaluation). Additionally the NRC said Pilgrim has “safety culture” issues and faulted Entergy for “overconfidence and complacency” in the face of safety operations. The NRC says that Entergy’s failure to correct problems is a “significant programmatic deficiency that could lead to worse errors if uncorrected.”

Entergy has had enough chances to fix the many problems that plague Pilgrim. It is time to put public safety first and stop playing Russian roulette with a nuclear reactor having a troubled history of recurrent performance failures.

Meg Sheehan is a public interest attorney and native of Plymouth. Bill Maurer is a retired construction project manager.

Netanyahu

I think he is a conservative and he is trying to bolster the conservatives in the USA no matter what it takes. I don’t like him, but it takes a tough bully to survive in that part of the world.

I visited Israel many years ago. I think Israel as our family and brother…do what ever it takes to assure my brother's survival and success. I think we owe that to the dead Jews for our failure before WW II, with standing by and doing nothing. I’ll die for your right of survival today!

Yep, just like us, they are not perfect.

Dam, I think Netanyahu is right. My philosophy is all war or total war, or doing nothing. Knock them out before they can damage our interest or hurt us. I think an injured mad dog running around the world is very dangerous. 

Iran killed or maimed many of servicemen in Iraq or Afghanistan...that is unforgivable!!! 

I think Netanyahu is betting on the right side; Iran is just stringing the USA along and they will never come to any agreement with the USA.

Is he setting up the president for a great fall?

So how is this going to look like in six months with everyone walking away from the negotiating table without no deal.

Netanyahu is playing Obama…

I sometimes think Israel always ends up undermining the current administration…either Bush or Obama…for the perceive greater ends of Israel in the future.

  

Maybe only as a tool for a Israeli politician to get elected.

Magic New NRC Technique Discovers RCIC Flaws At LaSalle Nuclear Plant?

These kinds of postulated shorts have been about the industry for thirty years. The agency has been discovering this often.

See, the NRC’s risk basically comes from a numerical calculation from acceptable core damage or fence line dose. I think this is chocked full with self-interested assumption. Everyone except me thinks if a licensee knows the real risk…this will push the licensee to acceptable behavior. This doesn't work. 

I think these licensees need an incentive, such as once an awhile, a bat upside the head. It will deter other licensees from making the same mistake. 

You catch how they are diluting the feedback...both plants at this site had the same flaws. But the NRC treats it as just a flaw in one plant with a “green very low safety significance”. 

If you believe the assumptions and the efficacy of the NRC’s calculated significance to change behavior…it would be 2 times a green very low safety significance. This is all snake oil…  

Bottom line, the agency doesn't create enough terror in the industry where on the emergence of this kind of accident mitigation flaw discovered, all the industry would be terrified of having the NRC finding one more.

So why did they discover this now, instead of ten or twenty years ago?  

February 27, 2015: LASALLE COUNTY STATION, UNITS 1 AND 2, TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000373/2014008; 05000374/2014008

• Green. The inspectors identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the licensee’s failure to ensure that the alternate shutdown capability was independent of the fire area. 

Specifically, in the event of a fire in the control room, the alternate shutdown capability for 16 motor operated valves (MOVs) associated with the Reactor Core Isolation Cooling (RCIC) may be affected, and may not be available due to lack of breaker fuse coordination. Fire-induced failures could result in tripping valve power supply breakers prior to tripping the control power fuses for several motor operated valves, thereby, potentially imparing the operation of RCIC from the Remote Shutdown Panel (RSP). The licensee entered this issue into their Corrective Action Program and established compensatory measures, and added steps to the safe shutdown procedures to reset the affected breakers if needed. In addition, the licensee intended to perform plant modifications to replace or revise existing breakers settings to correct the issue. 

The inspectors determined that the issue was more than minor, because fire induced circuits could impair the operation of RCIC and complicated shutdown of the plant in the event of a fire in the control room. The finding affected the Mitigating Systems Cornerstone. The finding was determined to be of very-low safety significance based on a detailed risk-evaluation. This finding was not associated with a cross-cutting aspect because the finding was not representative of the licensee’s current performance.

Palisades: Heading Back to a Yellow or Red Finding, or Worst?

You get the trend, this has repeated over and over again. The NRC attention(yellow finding) marginally drives the safety culture up for a year or two…then it goes back into steep decline.

Right, this is about holes through their titanium gonad shields and vest?  I wonder if there is boys or girls titanium gonad shields? 

It shows how ineffective the agency is.

The deal with the inaccurate dose calculations,  is they can artificially increase the efficiency of the high dose employees. The lower dose allows these employees to work more at other plants and within the radiation fields. Sometimes with big accumulated doses these employees are forced to work in non radiation fields for days, weeks and months because they are getting close to their legal dose limits.   

NRC Has Increased Oversight of Palisades Nuclear Power Plant  

The Nuclear Regulatory Commission staff issued a white finding of low-to-moderate safety significance to the Palisades nuclear power plant for the failure to accurately calculate radiation doses to workers during an activity last year. The finding will result in increased oversight by the NRC. The plant, operated by Entergy Nuclear Operations Inc., is located in Covert, Mich., five miles south of South Haven. The doses received by the workers were below the NRC’s annual radiation limit and are not expected to have any impact on their health. NRC inspectors reviewed of plant’s methodology for calculating doses to workers involved in replacing control rod drive housings during the 2014 refueling outage. They determined that the methodology did not meet NRC requirements. Specifically, the licensee failed to ensure that radiation dosimeters were placed in the highest exposed location of the body for this activity, which resulted in inaccurate dose calculations. In addition, the licensee failed to establish a procedure to ensure proper placement of dosimeters. This resulted in inaccurate calculation and assignment of dose for numerous workers. “Even though this incident did not result in harm to workers, our action underscores the importance of adhering to NRC requirements to ensure an accurate understanding and adequate monitoring of doses to workers at nuclear plants,” said NRC Region III Administrator Cynthia D.Pederson...

Profound financial problem with Entergy? 

Vt. AG pushes NRC to look into Entergy finances
By RICHIE DAVIS

Recorder Staff
Wednesday, February 25, 2015
(Published in print: Thursday, February 26, 2015)

Vermont’s attorney general has endorsed a petition filed with the Nuclear Regulatory Commission raising concerns about Entergy Nuclear’s ability to pay for decommissioning the shutdown Vermont Yankee nuclear plant, and seeking an investigation into its fiscal health in relation to the Vernon plant as well as Pilgrim Station in Plymouth and the FitzPatrick reactor in Oswego, N.Y. 

Vermont’s filing joins a similar action by Massachusetts Attorney General Martha Coakley last October and New York’s attorney general last April in the petition filed by Citizens Awareness Network in March 2013. The petition, still pending before the NRC, raises concerns about Entergy’s financial viability in anticipation of what are expected to be more than $1 billion in decommissioning costs for the plant, as well as the company’s intention to use $50 million of its decommissioning fund — now totaling about $650 million — to provide security of high-level radioactive waste stored on site. 

“The State of Vermont ... has a direct interest in ensuring that Entergy, when financing certain post-closure activities, abides by applicable NRC regulations and contractual obligations, which limit the circumstances in which Entergy can withdraw funds from its Nuclear Decommissioning Trust Fund. To determine whether Entergy — either through subsidiaries or the parent corporation — has adequate financial means without undue or unauthorized reliance on the ... fund, the NRC should fully investigate the financial qualifications of Entergy and its subsidiaries, including directing Entergy to respond to the issues raised by the Attorneys General of New York and Massachusetts,” wrote Chief Assistant Attorney General William E. Griffin. “Vermont also retains a strong interest in how Entergy intends to finance its obligations.”

CAN President Deborah Katz said, “These states are really concerned about how Entergy is going to be responsible.”
She added, “When the Memorandum of Understanding came through for Entergy to buy Vermont Yankee, the parent corporation said that it would cover expenses, and would be responsible if Entergy couldn’t come through with the money. We believe it’s really important to understand Entergy’s vulnerability, in not just an operating reactor, but for issues of safety in terms of cleanup, since they want to end the emergency planning zone and to substantially diminish what’s in the decommissioning fund.” 

Entergy Vice President Michael Twomey told members of two Vermont legislative committees last week that the company offers no guarantees it will pay to decommission the plant if the job is still not done by the end of a 60-year “SAFSTOR” period. The Associated Press reported him telling the House and Senate Natural Resources committees that he expects there would be litigation, with the state and Entergy taking different positions...