New Peach Bottom 2.206 Notes

Originally published on 11/8/2012

My first VY 2.206 on SRV's was dated March 17, 2011...

Nov 18: And then this miserable and unprecedented SRV testing failure with unit 2 this sept? The NRC last year just before before Unit 3's refueling whacked them in the head with that terrible Inspection Report...I bet you with the NRC breathing down their throats with last year's inspection report about SRVs the vendor and Exelon did the testing without any falsification. I bet you this year's bad leak rate testing results indicated for many years in the past Target Rock and Exelon colluded and conspired to downplay SRV testing defects...falsification...to not disclose problems with the SRVs and SVs in the self interest of all parties.

Peach Bottom 3 begins refueling: 9/10/2011 end Oct 17
Peach Bottom 2 begins refueling: 9/12/2012

So the below 2011 Inspection Report came just before the unit 3 2011 outage...the SRVs then the next outage on the other unit 2 failed so miserably in  LER2012001.

Peach Bottom IR 2011010 
July 25, 2011 through August 12,2011...inspection letter dated Sept 16

The inspectors identified a finding of very low safety significance (Green) involving a NCV of 10 CFR 50 Appendix B, Criterion XVl, "Corrective Action," because Exelon staff did not implement corrective actions in a timely manner to correct safety relief valve (SRV)/safety valve (SV) lift setpoint drift in excess of Technical Specification 3.4.3, "Safety Relief Valves and Safety Valves" requirements. Specifically, Exelon staff did not implement timely or adequate actions to correct SRV lift setpoint drift that, on four occasions since 2004 and as recently as 2010, has exceeded TS surveillance acceptance criteria and resulted in TS non-compliances.
Since 2003, six of the last eight outages at Peach Bottom have had as-found SRV/SV lift test failures outside the TS SR 3.4.3.1 acceptance criteria of +l-1o/o. On four of those occasions there were greater than two SRV/SV setpoint failures which resulted in noncompliance with TS 3.4.3Each time Exelon staff initiated lRs to document the as-found conditions in the corrective action program. In general, since 2003 Exelon staff has determined that the SRV/SV setpoint drift experienced at Peach Bottom is due to overly restrictive TS setpoint criteria (10lo vs. typical industry standard of 3o/otolerance) and have not identified the condition to be a result of equipment reliability or maintenancerelated aspects. Exelon statf has.consistently determined that a TS amendment to increase the setpoint tolerance to 3%, consistent with other Exelon sites, was the appropriate corrective action to address the TS noncompliance condition that existed at both units. Exelon staff, except for the action to evaluate and submit a TS revision. have not recommended interim or long-term corrective actions to address the SRV/SV setpoint drift TS compliance issue. 

The inspectors' corrective action review noted that as early as 2003 Exelon staff had discussed the option of submitting a TS revision to increase the SRV/SV setpoint tolerance. ln2007 (lR 559430), Exelon authorized a vendor to conduct a SRV/SV tolerance study to evaluate the feasibility and potential impacts of an increase in SRV/SV setpoint tolerance to 3o/o. Based on the results of that study, in early 2009, Exelon authorized a more comprehensive evaluation by a vendor whicn was completed in March 2010 and indicated a 3% tolerance would likely be acceptable with some additional site specific areas of evaluation. However, in May 2010, Exelon deferred the revision since an extended power up-rate project was being considered and the impacts of that power up-rate on the SRV/SV setpoint tolerance, at that time, was not fully known. Subsequently, Exelon staff identified during its most recent outage on Unit 2 in 2010 that two SRVs and one SVs failed to meet TS allowable tolerance and therefore were in violation of TS 3.4.3 as documented and submitted by Exelon in LER 4500027712010003. Exelon staff's evaluation (lR 121662811120516) determined that the non-compliance issue was the result of less than aggressive implementation of a TS revision for the SRV/SV setpoint tolerance. 

The inspectors' review determined that Exelon staff has not implemented timely corrective actions consistent with expectations outlined in LS-AA-125, "Corrective Action Program Procedure," in that actions have not been timely or effective to correct a longstanding condition adverse to quality (sRV lift setpoint rs non-compliances). Specifically, the inspectors determined that the action identified by the station to correct the SRV/SV setpoint drift and associated TS non-compliance aspects has not been implemented. Exelon has deferred or delayed implementation of the TS revision on several occasions. Additionally, the inspectors determined that Exelon has had several  to revisit the timeliness aspects of the long term TS revision action and has not identified interim or compensatory corrective actions to mitigate future TS noncompliances with regard to SRV/SV lift setpoints. The inspectors noted that Exelon staff has implemented several SRV/SV reliability actions over the last five years to improve overall SRV reliability; however, based on interviews with engineering staff and review of corrective action documents, those actions are not expected to directly mitigate or address 
the TS non-compliance vulnerability that still exists regarding the SRV/SV lift setpoint As documented in lR 112051611216628, Exelon staff has actions scheduled in2012 to conduct site specific evaluations required for the TS revision. However, the inspectors also noted that the actual date of the TS revision submittal, based on interviews with Exelon staff, is not affirmed and may continue to be delayed due to continuing conflicts with power up-rate considerations. The inspectors determined that corrective actions resultant from lR 112051611216628 have not resulted in corrective actions to mitigate or address the potential for continued TS setpoint non-compliances going forward. Exelon staff initiated lR 1250472tor disposition of this issue in the station's CAP. 

The Buna -n LER event date was 9/25/2011-11/18/2011.
The inspection report with SRV set point inaccuracy and unreliability was 7/ 25/2011 through 8/12/2011.
My first 2.206 was submitted1/24/2012 and cleared out on Sept 4, 2012.
New SRV setpoint LER 9/25/2012-11/2/2012.

My second was submitted on Oct 3, 2012.

This is code language for: According to risk perspective, it is perfectly permissible for a plant to install grossly defective safety components contrary to federal regulations and engineering codes...in order to support profitable plant operation because the risk of a bad outcome is insignificant. The societal benefits of a plant with a high capacity factor and  supporting the stock price far outweighs the risk of an improbable plant accident caused or made worse by knowingly defective components through intentionally violating engineering codes and federal regulations.  Defective parts and violating federal regulations on safety component in safety systems designed for improbable accidents causes insignificant public risk...    

The problem I have with the NRC is they are accurate on the ridiculous hyer specifics not related to plant condition, but corruptly and  obscenely inaccurate from a plant centric or license operator centric point of view. Just facts that support their agenda! You just pick any old facts to make the PRB's Mars generic point of view,  that supports the point of view Mike Mulligan is a dick...   

The new VY inspection report: "This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F."  

The Peach Bottom PRB response to me: The typical service temperature range for Buna-N material is -40F to +250F. 

Kind of sad the falsifying PRB documents...how comes the error always goes on the industry's side. The NRC was negligent with providing safety information to me and it was inaccurate...their research vetting was poor. Do you think the board would ever tolerate a 25 degree inaccuracy on the other side saying the buna-n material was only good to 200 degrees. You would have caught and research that repeatedly to make sure.

Nov 18

...Four SRV modes of operations
1) Normal plant operations and testing
2) Plant transients and accident
3) Shudown
4) During testing and surveillance 

...VY SRV LER: Technical Specification (TS) 3.6.D requires at least three of the four RVs to be operable for overpressure protection of the Reactor Coolant System and TS 3.5.F requires all four RVs to be operable to support the Automatic Depressurization System (ADS) function of the Core Standby Cooling System.

Peach Bottom SRV threaded seal : ADS system is designed to provide depressurization of the reactor coolant system during a small break loss-of-coolant accident if the High Pressure Coolant Injection (HPCI) system is not able to maintain the required water level in the Reactor Pressure Vessel (RPV). ADS SRV operation reduces the RPV pressure to within the operating pressure range of the low-pressure Emergency Core Cooling System (ECCS) subsystems so that the ECCS low-pressure subsystems can provide coolant inventory makeup.

Peach Bottom Atomic Power Station (PBAPS) Unit 3

***Big issue not resolved; is the containment nuclear safety valve air actuators fully environmentally qualified to 400 degree Fahrenheit and are all other safety equipment in containment fully environmentally qualified? 

Oct 17, 2012: 

Sept 4,2012: NRC's response to my 2.206 Petition

Your letter dated January 24,2012, addressed to Mr. William Borchardt, Executive Director for Operations, has been referred to the Nuclear Regulatory Commission's (NRC) Office of Nuclear Reactor Regulation, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.206. In your petition, you requested that the NRC: 

NRC Require the Immediate Shutdown of the Peach Bottom Nuclear Power Plant.

The PRB denied the request for immediate action because there was no immediate safety concern to the plant or to the health and safety of the public. The NRC reviewed the licensee's evaluation and actions related to this matter and concluded that the 3-ADS-SRV 71 B degraded seal condition was not caused by improper maintenance practices. Also, trend data did not indicate a potential degradation in that the same seal material had been used at PBAPS Units 2 and 3 for the last 20 years with no other failures. These facts support the conclusion that the failure of the 3-ADS-SRV 71 B threaded seal was not a common mode failure, or an age-related failure. but was isolated to the particular seal installed in November 2010. The inspectors assessed the risk associated with the issue by using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations SDP [Significance Determination Process]." The 3-ADS-SRV 71 B is one of the five PBAPS Unit 3 ADS reactor vessel relief valves. In order to perform the ADS system safety function, four of the five ADS SRVs are required to function. The four other ADS SRVs passed the leakage test, and would have been capable of de-pressurizing the reactor pressure vessel for design basis events. Therefore. during the period that the 71B SRV was inoperable, the overall ADS safety function was maintained. The NRC staff's evaluation of this issue has been documented in Inspection Report 05000277/20120003 and 05000278/2012003, dated August 14, 2012 (ADAMS Accession No. ML 12227 A323). 

So the below is how the NRC-PRB Peach Bottom frames it.

"The NRC reviewed the licensee's evaluation and actions related to this matter and concluded that the 3-ADS-SRV 71 B degraded seal condition was not caused by improper maintenance practices." 

This is how NRC Vermont Yankee Frames it this Oct.

" VY2012004/Oct 31, 2012: However, in consultation with the manufacturer, Entergy incorrectly concluded that the changes to the actuators were “like for like” replacement of components. Entergy failed to determine that the seal material for the actuator stem nut had been changed from Silicon to Buna-N. This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F. During the 2009 refueling outage, Entergy found nitrogen to be leaking from the actuators and determined the actuator stem nut seals were degraded. However, Entergy’s evaluation of the seal incorrectly concluded that the seal material was defective and a new Buna-N seal was installed. Entergy performed a subsequent evaluation of the seal material and determined that the material was Buna-N, not defective, and the failure of the material was due to exceeding the thermal rating (225 degrees F) of Buna-N. Following identification that the seal material did not meet environmental conditions, Entergy performed an operability determination which concluded that the ADS system was operable, but degraded." 

So the below says the NRC found  no degradation in the actuator seals and diaphragms for 20 years...catch that seal  material certainty/ uncertainty language gaming? The PRB provides no evidence to proves their assertion or the NRC would never look for evidence that contradicts the PRB position.  That is how the college boys play corruption.

"Also, trend data did not indicate a potential degradation in that the same seal material had been used at PBAPS Units 2 and 3 for the last 20 years with no other failures."

This is the first report that contradicts the PRB position.

Licensee Event Report (LER) 2-06-02: Based on a review of testing performed on Safety Relief Valves (SRVs) during the P2R16 Refueling Outage, Site Engineering personnel determined that the 71 B and 71 G SRVs did not meet their allowable leak rate for the pneumatic actuation controls for the Automatic Depressurization System (ADS) feature of the SRVs. The cause of the 71 B and 71 G ADS SRV pneumatic leakage is attributed to leakage from each of the SRV's actuator diaphragm and solenoid valve leaks. 

In the above, we don't the know the nature of the actuator leakage...but certainly it contradicts my assertion with the reliability and the environmental qualifications of these actuators. They make these nuclear plants provide to outsiders so little information.  That was my assertion all along, with the fear the actuators weren't environmentally qualified...the PB PBR turned it into justifying the reliability of the seals in normal operation for 20 years. My issue has always been about how the plant would perform in the worst accident. 

Here is the whopper (05000277/200301) report on a dual Peach Bottom plant trip over a partial electrical LOOP disturbance...it is horrendous how complicated it was with bad maintenance of equipment.

During the Peach Bottom dual unit reactor scram event on September 15, 2003, several main steam relief valves experienced problems. Specifically, the more significant malfunctions occurred with Unit 3 main steam relief valves RV-3-02-071D and RV-3-02- 071G, which are three-stage Target Rock relief valves. Sounds like more problems that they didn't document than these two.

071D had its pilot valve misaligned to the seat and it stuck open to 400 degree...a stuck open SRV is a direct precursor to a core melt according to their new SOARCA. So at "unit 3" one SRV stuck open when manually opened and the other one stayed closed when the switch was placed in manual open. And their flaws weren't self evident just before the accident and same flaws weren't evident during bench testing  and pre-operation testing of the valves.

RV-3-02-071G, an Automatic Depressurization System (ADS) valve, opened on high reactor pressure and initially closed when control room operators placed the control switch in auto. After initial successful manual openings of RV-3-02-071G during the event on September 15, 2003, the SRV failed to open manually.

'During the disassembly of the RV-3-02-071G air operator, technicians noted that the torque on the actuator bolts was low. The eight diaphragm bolts and diaphragm plate looked “blue” which is indicative of a heating effect. The diaphragm was severely damaged, most likely due to exposure to heat. The air operator had internal steam damage and condensation residue. Further inspection revealed corrosion and condensation residue in the air operator stem and packing area. Areas where the packing had lost contact with the stuffing box were noted. The packing that is installed in the Target Rock SRV air operator is asbestos Teflon which is no longer manufactured. Exelon is working with Target Rock to select suitable replacement packing material for the SRV air operator"

Imagine that; this is the big shift from 400 degree F asbestos actuator threaded seals to the 225 degree buna n crap. Can you imagine what the steam leak would do to the 225 degree F buna n instead of the 400 degree asbestos? We don't know what the environmental qualification is with the actuator diaphragm? This steam leak ate the asbestos threaded seal.      

 "The air operator had internal steam damage and condensation residue. Further inspection revealed corrosion and condensation residue in the air operator stem and packing area. Areas where the packing had lost contact with the stuffing box were noted. Condensation residue left a clear trail of where the high energy steam leaked by the stem packing and up through the air operator damaging the diaphragm."

Then more troubles with the SRV actuator leakage in LER 2-06-02 questioning the PRB's "the sky has been all clear for 20 years" with SRV actuator troubles. Is the actuator's reliable and durable enough for nuclear safety?

Based on a review of testing performed on Safety Relief Valves (SRVs) during the P2R16 Refueling Outage, Site Engineering personnel determined that the 71 B and 71 G SRVs did not meet their allowable leak rate for the pneumatic actuation controls for the Automatic Depressurization System (ADS) feature of the SRVs. The cause of the 71 B and 71 G ADS SRV pneumatic leakage is attributed to leakage from each of the SRV's actuator diaphragm and solenoid valve. 

So the 2003 AIT  duel  plant trip SRV operational problems and Licensee Event Report (LER) 2-06-02 actuator and solenoid  leakage prior to testing thoroughly questions  the PBR's statement of : "Also, trend data did not indicate a potential degradation in that the same seal material had been used at PBAPS Units 2 and 3 for the last 20 years with no other failures." We have no idea of the magnitude of the abnormal indications with the  internals and seals, but didn't rise to the level of abnormal leakage and didn't trip a leak rate failure.

***Remember,  Entergy Vermont Yankee and their NRC inspectors say the buna material that Peach Bottom has is environmentally grossly improper SRV actuator threaded seal material, while Exelon Peach Bottom and their NRC inspectors say it is fully environmentally qualified material...

Licensee Event Report (LER) 2-06-02

November 15, 2006This LER reports a condition involving Automatic Depressurization Valve deficiencies that was discovered during a recent Refueling Outage.  

Based on a review of testing performed on Safety Relief Valves (SRVs) during the P2R16 Refueling Outage, Site Engineering personnel determined that the 71 B and 71 G SRVs did not meet their allowable leak rate for the pneumatic actuation controls for the Automatic Depressurization System (ADS) feature of the SRVs. Additionally, the 7 IC SRV, Serial Number (S/N) 83, did not properly re-close on the fourth actuation during laboratory testing. The cause of the 71 B and 71 G ADS SRV pneumatic leakage is attributed to leakage from each of the SRV's actuator diaphragm and solenoid valve. These leaks only occurred when the SRV solenoid valves were energized. The diaphragms and solenoid valves associated with the 71B and 71G ADS SRVs were replaced. As-left leak testing was performed and the valves were restored to an operable condition prior to plant startup from the P2R16 Refueling Outage. A refurbished SRV was installed in the 71C SRV location to replace the S/N 83 SRV. There were no actual safety consequences associated with this event. This event  was determined to not be risk significant.

The 71 B and 71 G SRVs did not meet their allowable leak rate for the pneumatic actuation controls for the Automatic Depressurization System (ADS) feature of the SRVs. The as-found leak rates for the 71 B and 71 G SRVs were documented as off-scale (only when the SRV was actuated) and therefore, exceeded the leak rate limit of 100 cc/min.
The cause of the 71B and 71G ADS SRV pneumatic leaks was attributed to a failure of the associated diaphragm (EIIS: PC) and solenoid valve (EIIS: FSV) for each of the SRVs. When energized, the solenoid valve switches ports resulting in pneumatic pressure being applied to the SRV diaphragm. This results in the opening of the ADS SRV. The diaphragm and solenoid valve leakage only occurred when the particular SRV was actuated and therefore, the leakage would not have been detectable during normal plant operations.

****

Units 2 and 3 Automatic Scrams Resulting from an Off-Site Electrical Grid Disturbance

Licensee Event Report (LER) 2-03-04
Nov 7 2003.  

SRVs on both Units 2 and 3 properly relieved pressure with the exception of the Unit 3 71 D SRV. This SRV did not re-close promptly as designed. The 71D SRV re-closed approximately 15 minutes after its actuation at approximately 400 psig reactor pressure. Also, at approximately 0600 hours, the Unit 3 71G SRV did not open when manually actuated from the Main Control Room during reactor pressure control operations. The 71G SRV did initially open and perform its over-pressure protection function when the event initially occurred and was manually actuated prior to 0600 hours for reactor pressure control.

(September 4, 2012 -"The PRB denied the request for immediate action because there was no immediate safety concern to the plant. or to the health and safety of the public. The NRC reviewed the licensee's evaluation and actions related to this matter and concluded that the 3-ADS-SRV 71 B degraded seal condition was not caused by improper maintenance practices 

The below from the PRB is grossly inaccurate.... I guess if the NRC is into "certainty gaming" the statement is accurate.

Also, trend data did not indicate a potential degradation in that the same seal material had been used at PBAPS Units 2 and 3 for the last 20 years with no other failures.?   

The 71 D failed open to 400 psig due to foriegn matter in the pilot valve seat... 

The failure of the 71G SRV to be subsequently opened was due to degradation of the air operator diaphragm for the 'SRV' The degradation was due to accelerated aging caused by exposure to high temperatures.-The high temperature condition was apparently caused by leaking packing material that isolates the air actuator from the second stage steam space. Further cause evaluation analyses are in-progress in accordance with the site Corrective Action Program.

The Unit 3 71D and 71G SRVs were removed and replaced with factory refurbished SRVs. Other SRVs on Unit 3 were 'also refurbished. An extent of condition review for other SRVs on both Units 2 and 3 was performed. It was determined that the PBAPS SRVs currently installed are highly reliable.

SUBJECT: NRC AUGMENTED INSPECTION TEAM (AIT) 05000277/2003013 AND 05000278/2003013, AND PRELIMINARY WHITE FINDING - PEACH BOTTOM ATOMIC POWER STATION

December 18, 2003EA-03-224

After the Unit 3 drywell became accessible, maintenance technicians identified air blowing from the air operator of RV-3-02-071G. This provided evidence that the air operator diaphragm had failed. RV-3-02-071G was removed from Peach Bottom Unit 3 and sent to Wyle Labs for failure analysis.

The “as-found” lift setpoint of RV-3-02-071G was 1130 psig, 15 psig below its nominal setpoint of 1145 psig, (-1.3%), which is not within Peach Bottom Technical Specifications specified nominal setpoint of +/- 1%.

During the disassembly of the RV-3-02-071G air operator, technicians noted that the torque on the actuator bolts was low. The eight diaphragm bolts and diaphragm plate looked “blue” which is indicative of a heating effect. The diaphragm was severely damaged, most likely due to exposure to heat. The air operator had internal steam damage and condensation residue. Further inspection revealed corrosion and condensation residue in the air operator stem and packing area. Areas where the packing had lost contact with the stuffing box were noted. Condensation residue left a clear trail of where the high energy steam leaked by the stem packing and up through the air operator damaging the diaphragm. 

Additionally, the packing nuts were identified to be loose. As an extent of condition, examination of the seven other SRV air operators removed in Unit 3 refueling outage 14 (3R14) revealed inconsistent torque of the packing nuts. The packing that is installed in the Target Rock SRV air operator is asbestos Teflon which is no longer manufactured. Exelon is working with Target Rock to select suitable replacement packing material for the SRV air operator. Exelon concluded that the failure of RV-3-02-071G was caused by the failure of the air operator diaphragm, preventing the SRV from opening remotely from the control room on control switch demand.

Additionally, the packing nuts were identified to be loose. As an extent of condition, examination of the seven other SRV air operators removed in Unit 3 refueling outage 14 (3R14) revealed inconsistent torque of the packing nuts. The packing that is installed in the Target Rock SRV air operator is asbestos Teflon which is no longer manufactured. Exelon is working with Target Rock to select suitable replacement packing material for the SRV air operator. 

Exelon concluded that the failure of RV-3-02-071G was caused by the failure of the air operator diaphragm, preventing the SRV from opening remotely from the control room on control switch demand.

On September 15, 2003, RV-3-02-071G was initially opened and closed manually by operators in the control room, during which time high temperature/pressure steam leaked past the packing into the air operator causing the failure of the diaphragm in the air operator.

LER 03-2011-03-00

Event date 9/25/2011
Report date 11/03/2011
Refueling started Sept 11, 2012...

Based on evaluation of the 9/25/11 surveillance testing performed..

...Based on March 2011 vendor technical evaluation report, upgrades to the diaphragm thread seal for ADS SRVs on Units 2 and 3 are planned. There were no actual safety consequences as a
result of this event.

...Based on evaluation of the 9/25/11 surveillance testing performed on Safety Relief Valves (SRVs) during the P3R18 Refueling Outage, site Engineering personnel determined that the 71 B SRV did not meet its allowable leak rate for the pneumatic actuation controls for the Automatic

Depressurization System (ADS) feature of the SRV. This resulted in a degradation of the number of times the 71 B SRV could be used during a design basis event. 

...Due to the leakage associated with the 71 B ADS SRV, it was determined that the ADS feature of the SRV was inoperable and the event was considered as a condition prohibited by Technical Specifications (TS).

*** The SRVs are manufactured by Target Rock Co. and are 'three-stage' relief valves. The thread seals are Target Rock part number 715-0004 containing Buna-N material.

...Each ADS valve is provided with a short-term, safety grade, pneumatic supply by means of its associated accumulator to provide sufficient capacity to cycle the valve open five times at atmosphere pressure, twice at 70% of containment design pressure, or once at containment design pressure, all within a 4-hour time period.

***In the unlikely event of a design basis event (including no credit for the normal nitrogen gas supply), the 71 B ADS SRV would not have been able to stroke multiple times as required by the licensing basis.

***The cause of the 71 B ADS SRV inoperability was attributed to a failure of the associated actuator diaphragm thread seal (EIIS: SEAL).

...When inspected by maintenance personnel, the thread seal had indications of being dry and brittle. Subsequent review by Engineering personnel determined that the apparent cause was thermal degradation of the thread seal material.

...Based on March 2011 vendor technical evaluation report, upgrades to the diaphragm thread
seal for ADS SRVs on Units 2 and 3 are planned.

PEACH BOTTOM - NRC INTEGRATED INSPECTION REPORT 05000277/2012003 AND 05000278/2012003, NRC OFFICE OF INVESTIGATIONS REPORT 1-2012-011, AND
EXERCISE OF ENFORCEMENT DISCRETION

August 14, 2012April 1, 2012 through June 30, 2012
On September 25, 2011, while Peach Bottom Unit 3 was shut down for a scheduled refueling outage, Exelon personnel performed a routine ST on the Unit 3 71B SRV.
(Oct 13, 2012: Entergy failed to determine that the seal material for the actuator stem nut had been changed from Silicon to Buna-N. This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F. During the 2009 refueling outage, Entergy found nitrogen to be leaking from the actuators and determined the actuator stem nut seals were degraded. However, Entergy’s evaluation of the seal incorrectly concluded that the seal material was defective and a new Buna-N seal was installed. Entergy performed a subsequent evaluation of the seal material and determined that the material was Buna-N, not defective, and the failure of the material was due to exceeding the thermal rating (225 degrees F) of Buna-N.) 

I now get it where the phrase "stem nut seal" comes from. This is the first inspection the NRC renames it from a  "diaphragm thread seal" to a "stem nut seal". There is no doubt the Peach Bottom and Vermont Yankee Inspectors were in direct contact with each other over this SRV problem...they knew simultaneously one 225 degree buna seal was grossly inappropriate for one plant and then perfectly safe for the next. Right, I made a big deal this "stem nut seal" phrase was not in seen in  multiple NRC and VY report since 2010, but  first showed up in the most current VY inspection report. So this is the first report for both PBABS and VY where "stem nut seal" first shows up.    

The NRC reviewed the licensee’s evaluation and actions related to this matter and concluded that the degraded seal condition was not caused by improper maintenance practices.

Please!!!

****

....Just think about it, Exelon is doing extraordinary backflips to save money over their nuclear plants...

Going back to yearly outages...

Exelon Open to Closing Oyster Creek Reactor Pre-2019
By Julie Johnsson - Nov 13, 2012 8:31 PM ETExelon Corp. (EXC), the largest U.S. nuclear plant operator, would consider closing its Oyster Creek station before the plant’s planned 2019 decommissioning, Chief Executive Officer Christopher Crane said.
Exelon would accelerate plans to close Oyster Creek in Forked River, New Jersey, if it faced unexpected new capital costs at a time when depressed power prices and cheap renewable energy are squeezing nuclear generation margins, Crane said in an interview yesterday.
Exelon already has deferred plans to boost capacity at its LaSalle nuclear station in northern Illinois and Limerick plant in Pennsylvania, saving $1.2 billion in the near-term, the chief executive said.
The Chicago-based company also plans to refuel its Clinton nuclear plant annually instead of every two years in a bid to boost profitability at the hardest-hit plant in Exelon’s 17-reactor system, he said.
“That’s a core expense to the company, but hopefully we can keep the plant open longer as we navigate through this issue,” Crane said.
Exelon hasn’t yet considered decommissioning the 1,043-megawatt Clinton, Illinois, station, which is among the youngest in its portfolio, he said. The plant’s economics have been hurt by a glut of cheap wind energy in Illinois.

...They are dealing with massive amounts of equipment that are obsolete ...many manufacturers have gone out of business and other have stopped making parts on the obsolete production lines.

And don't  forget  low quality cheaply manufactured  foreign parts and not traceable parts have put many domestic lines out of business...

..Nov 12: So there are two regulatory regimes of the NRC.

1) Issues that don't threaten plant operation.  The agency and a utility are more open with these issues.

2) Issues of gave nuclear safety or issues that threaten plant operations...shutdown to fix something generated by outsiders and internal screw-up.  In this mode truth and following the rules is optional.

A cover-up in the old days generally consists of putting everything under a secrecy shield. The trouble with keeping secrets like this, is exposing these secret even many years later can cause a scandal...create changes. The new sophisticated cover-up...utilitarian...you just kept secrets for a particular purpose and  specific time frame.  You can play this game over and over again  without getting caught.

So you engineer disclosures of wrongdoing for months and years after the event...you engineer investigation and inspections, coordinate the inspection and  investigation across the agency and NRC until the fix has occurred many years after the regulatory breakdown. Then  for months  and years you begin to piecemeal the details of what happened to make it very difficult for anyone to be able to connect the dots. Then nobody can discover old secrets...creating a scandal.    

Information Notice 1994006

result of high temperatures that would exist following a LOCA could lead to depressurization of the nitrogen system and consequent inability to provide the 100-day supply of nitrogen to the ADS valves. The short-term operability of the ADS valves following a LOCA would be maintained because the ADS valves also are supplied by nitrogen accumulators and supply lines that are separated from the main nitrogen header by check valves.

In January 1993, during a review of the components that form the pressure boundary of the of the safety-related portion of the nitrogen supply system at the Fitzpatrick Nuclear Power Plant, the licensee found that certain parts of the drywell cooler air-operated dampers and their associated solenoid valves were made of Buna-N, an elastomer material that is not suitable for high temperature conditions. These air dampers and solenoid valves are supplied by the same nitrogen header that provides the 100-day nitrogen supply for the automatic depressurization system (ADS) valves. Consequently, the failure of the Buna-N parts as a 

I was inaccurate...Peach Bottom does identify the valve manufacture:

"The SRVs are manufactured by Target Rock Co. and are 'three-stage' relief valves. The thread seals are Target Rock part number 715-0004 containing Buna-N material."

Sounds like 1978:

B 78-014 Deterioration of Buna-N Components in ASCO Solenoids BWR

October 23, 2012
Mr. John DeBonis, Quality Assurance Manager
Curtiss-Wright Flow Control Company
Target Rock Division
1966E Broadhollow Road
East Farmingdale, NY 11735-1768

SUBJECT: NUCLEAR REGULATORY COMMISSION INSPECTION REPORT
NO. 99900060/2012-201 AND NOTICE OF NONCONFORMANCE

Pg 21:

PR-066, “Problem Report for Embrittled Buna-N Thread Seals,” December 4, 2010 

So it is a generic issue...not a one off like the NRC thinks...

So on Oct 25 2010 VY notified the NRC, I believe it was discovered almost a year before this...with the Curtiss Wright guys reporting it in Dec 4 2010.

Vermont Yankee LER:

This leakage, when combined with the RV accumulator leakage, caused two of the four RVs to not meet design actuation requirements. The thread seals were manufactured in 2002, supplied to Vermont Yankee (VY) in new style actuators in 2008 and were in service for one operating cycle prior to the test. The thread seals in the new style actuators are made of Buna-N material, were manufactured by Parker Hannifin Corporation and dedicated for use in safety class applications by Curtiss-Wright Flow Control Corporation, 

I would die to read the Curtiss Wright report on this?


Hmm, mostly electrical stuff?

I am just saying, a nuclear industry huge fake repair parts and falsified parts certification is ongoing in South Korea...

And the companies and international certification authorities remain secret...

In this instance, however, seven Korean and one US firm that delivered parts made in the United States and Canada were found to have fabricated certificates showing their products to have been approved by United Controls International.

...All "delivered parts made in the USA and Canada"... approved by United Controls International...

United Controls International
5139 S. Royal Atlanta Drive
Tucker, Georgia 30084
(770) 496 1406 tel
(770) 496 1422 fax
February 24, 2010
TI. S.

UCI meets the highest standards of quality control established by the nuclear industry and other standards authorities. UCI maintains a 10CFR50 Appendix B program audited by NUPIC. UCI has designed, manufactured, tested, and installed systems in nuclear power generation plants around the world. UCI maintains a database of thousands of nuclear qualified parts and electrical components that have been tested, dedicated, and qualified. UCI has an extensive array of test equipment and an experienced engineering staff to offer a broad range of testing services.

Lab: The United Controls International Lab consists of Level I to III of Electrical / Mechanical Test Inspectors. Our daily goal is to dedicate components by following Specific Test Instructions, Commercial Grade Item Specification Dedication Specifications (CGI), Commercial Grade Dedication Specifications (CGDS), Nuclear Qualification Test Plan (NQTP), Nuclear Quality Reports (NQR) and Quality Control Procedures. The technicians ensure records associated with the Commercial Grade Dedication Process are properly completed. Our lab is capable of Seismic, Environmental, Electrical, Mechanical, Hydrostatic and Pneumatic testing using appropriate calibrated equipment. 

Dedication Services

Qualification

• Seismic Qualification
• Harsh Environment Qualification (Radiation, LOCA, Thermal)
• Equipment Qualification
• EMI/RFI

Motor Services & Dedication

***Reverse Engineering

PC Board Services & Dedication

***Sourcing Obsolete Parts

Engineering Consulting Services

Engineering Design

Custom Control Panels & Systems Integration

Quality Assurance

Products

Engineering

Nuclear professional with Fukushima think the plant melted down all the valve electrical motors, pressure and level detectors and transmitters....all the electrical stuff and wires in containment was destroyed first with a lack of containment cooling...thereby making the future fuel meltdown irreversible. The lack of containment cooling is the weak link...it destroyed the electrical components first before the core melted down.

And if you have a even small primary leak in containment....it happens very quickly...

That is the problem with the blackout and other accident coping strategies, like the fire system and other means to put low pressure water into the core...you just don have the ability to keep the containment and its invaluable electrical equipment cooled. Once you destroy the electrical equipment in containment it makes an approaching core meltdown irreversible.

And believe me, everyone in the NRC in good standing would lie through their teeth...would never admit mike caught us all with our pants down and he caused us to shut down for repair a nuclear plant in order to do the right thing. They would never send that kind of signal with a kook shutting down a nuclear plant for the right reasons to the investment community. Right, if they ever were forced to shutdown Peach Bottom 3 for repairs of the actuator...the NRC would be found guilty of not shutting down PB2 and Vermont Yankee and a cover-up for the same reason. 

The NRC says they can't imagine the use of the manual operation of a SRV with the containment about 225 degrees for safety reasons...I say because these accidents are so complex, you have to have global temperature requirements in the containment...in other words, all the components need to be qualified for 400 degrees or 360 degrees because of the not estimatable complexity of these accidents.

I say I want to see a comprehensive engineering evaluation on how this wrong buna-n material and installed in PB 3, was in PB2 andVY...how the material performs in normal operation, it degradation mechanism and certainly its performance as temperatures rises until its worst designed accident. It is only ethical if they would have had this kind of fair material engineering evaluation before they installed this wrong crap in a nuclear power plant. As it sits now, the NRC is only making unprofessional assertions disconnected from any science and engineering with how this material will preform.

How much of this buna-n and low temperature type 2  actuator or electric motor environmentally unqualified  crap do they got in these nuclear plants?  

The industry and the NRC based on a profit centered  corporatized model is basing safety on god's eye view in a computer model only based on the relative worth calculation of components in a reactor...its high tech witchcraft...while everyone knows it is primitive, inaccurate and mostly blind...

The NRC defense is and they don't admit it....on the risk basis of manual operation of the SRVs are insignificant.

So ignore mike and put in lower quality materials because a worst accident coming out of a inability to use the manual-remote operation of SRVs is implausible and insignificant.

***I wonder what the most limiting gas flow component is in the nitrogen system. Is it the pressure  regulator, piping size or nitrogen vaporizer/ heater....lets say all the seals failed, could the nitrogen system still operate the SRVs? How much excess nitrogen flow/pressure capacity is there for the minimal operating pressure to open/shut the valve? Wonder what side of the nut seal is on the actuator...will the open or shut position of the valve be affected most.


I think the quickest way to get a core melt is to overheat the contaiment...destroy all the all the electrical, seal, and diaphragm components in containment. Make yourself blind and quadriplegic.  And a big part of this is the QA environmental...high temperatures...qualification of susceptible heat sensitive components in the containment. Right, either high temperature or age related past their service life or a combination...

I'd like to get a computer program that could model on a nuclear accident dialing down the temperature ( say from 400 to 225 degrees or past service life) qualifications and age related service life of susceptible components in the containment.

I think our sacred duty as nuclear professionals, as perfectly as possible, is to make the plant ready for the worst accident as a higher priority that regular power operations.

Even the PRW's don't model operator control of their PORVs....why?    

***PB Soarca

Certainly severe fuel damage at any plant would wreck the USA's nuclear industry and destroy 20% of our electricity capacity....I think a lot less fuel damage, or any fuel damage and is some cases a no fuel damage accident could get you to the exact same place as destroying an enormous amount of electricity capacity...

Should you align the nuclear industry so it protects itself against a highly improbable severe fuel melt accident or any kind of accident that would threaten national security...ie a much lesser fuel melt accident or bigger loss of credibility accident? We are talking about solely how the greater public perceives...perceptions...events in the industry, not how the industry predicts public perception.

One asks you, how do you align the nuclear industry materialistically in fantasy land disconnected from human relationships. The other asks you, how do you align the nuclear industry in a modern democracy and in hyper connected human relationships with the materialistic and component aspects playing a much smaller role. A severe core damage accident arises solely from faulty human relationships...it never emerges from throwing the dice or appears in a random manner.          

"The American Society of Mechanical Engineers (ASME) American Nuclear Society (ANS) PRA standard RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," issued March 2009 [ASME/ANS, 2009] defines core damage as "uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage are anticipated and involving enough of the core, if released, to result in offsite public health effects."

....From: Michael Mulligan

To: "Kim, James"  
Sent: Thursday, November 8, 2012 1:15 PM
Subject: Re: Peach Bottom 2 and 3- 2.206 Petition on SRV

Mr Kim,

Thanks for talking with me today...I would like to address the PRB.

The question I got are... for NRC QA requirements and licensee QA or procedures requirement? What are the required containment safety environmental temperature range of the:

Say, the RHR injection valve inboard electric motor?

The whole SRV air actuator including the diaphragm or any other plastic, rubber or seal material or insulation heat sensitive components?

The SRV air actuator electric solenoid valve and its wiring for manual operation in the control room.

Oh what the heck, the SRV itself?

Thanks,
mike

From: Michael Mulligan
To: "Kim, James"  
Sent: Thursday, November 8, 2012 3:56 PM
Subject: Re: Peach Bottom 2 and 3- 2.206 Petition on SRV

To the PRB board...

The NRC in the new VY IR 2012004 characterizes the buna material as unapproved material.
How come either PB or VY hasn't done a extent of cause/condition investigation with this low temperature unapproved material...how come the NRC hasn't forced their hand? Do we have other bad material in containments...

"Entergy modified the actuator system in 2008. However, in consultation with the manufacturer, Entergy incorrectly concluded that the changes to the actuators were “like for like” replacement of components. Entergy failed to determine that the seal material for the actuator stem nut had been changed from Silicon to Buna-N. This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F. During the 2009 refueling outage, Entergy found nitrogen to be leaking from the actuators and determined the actuator stem nut seals were degraded. However, Entergy’s evaluation of the seal incorrectly concluded that the seal material was defective and a new Buna-N seal was installed. Entergy performed a subsequent evaluation of the seal material and determined that the material was Buna-N, not defective, and the failure of the material was due to exceeding the thermal rating (225degrees F) of Buna-N." 

So PB 3 has incorrect material in their actuator. The VY NRC says the normal operating temperatures caused the degradation of the unapproved material, it failed as expected because the material wasn't designed for normal operating temperatures of the containment, which is totally contrary to what the PB PRB's explained to me.

Honestly, did an engineer research this below...can you get your upper limit temperatures straight or correct (225/250). All it would have taken is make a phone call to Peach Bottom. Hint, one came from a research on the internet on general buna-a and the other actually was pulled out of a nuclear engineering department document and actual vendor specs. I worry about the NRC being accurate about anything...

The problem I have with the NRC is they are accurate on the ridiculous hyer specifics not related to plant condition, but corruptly and  obscenely inaccurate from a plant centric or license operator centric point of view. You just pick any old facts to make the PRB's Mars generic point of view,  that supports the point of view Mike Mulligan is a dick...

The new VY inspection report: "This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F."  

The Peach Bottom PRB response to me: The typical service temperature range for Buna-N material is -40F to +250F.

Kind of sad the falsifying PRB documents...how comes the error always goes on the industry's side. The NRC was negligent with providing safety information to me and it was inaccurate...their research vetting was poor. Do you think the board would ever tolerate a 25 degree inaccuracy on the other side saying the buna-n material was only good to 200 degrees. You would have caught and research that repeatedly to make sure.

"In addition, the mechanical overpressure function of the SRV would not be impacted by even a complete failure of the ADS pneumatic thread seals. The mechanical overpressure function actuates the SRV based on a setpoint of reactor pressure." 

This is like being in a desert with the perspective of how a operator would see a SRV failure to operate remotely. What is in the procedures about this and how do they train you on the simulator? Would you are this point call the SRV inoperable, would you fear the failure of the automatic operation of the SRV is right around the corner? Would you call the whole valve inoperable for the EOPs, but behind the containment the automatic function was never threatened?

I hate the Soarca on this because they never consider the actuator as a failure mechanism....how the operator would perceive the lost of remote operation and they don't force training and procedure with how Soarca accepts a loss of SRV remote control room operation.

It would totally be irresponsible for a operating crew in such a serious condition as to use the automatic function of a SRV valve. You never depend on a machine or component to operate in such a risky venture as this. An automatic operation of a SRV without warning has the likely result of tripping HPCI or RCIC, let alone creating another scram at a inopportune time and additional trips of cooling components. There is so much consequence with depending on the random operation of a SRV...you got to have a trained person on operating the switch and communicating to the rest of the crew its operation.

So its an acceptable NRC practice to lose the remote operation of the SRV...how does a plant warn, procedural precautions, a containment temperature precaution explaining at these temperatures you need to expect to lose the capability to remotely operate the SRV. After this point does the EOP explain how to bring a plant under control with the remote operation of a SRV? I hate when the NRC abandons the control room operators with no procedures or training, on a acceptable secret lost of safety equipment.

You know a soarca blackout, these poor guys are trying to deal with this terrible complex accident in the control room. So you are in the deap shit, you are standing there operating the SRVs manually, and then one doesn't open on your command. Then the next one over fails to operate. You have no idea what is going on in containment. What do you do when all of the SRVs don't operate manually because the SRV seals burn up. Does your cognition break down on the complexity of the accident...then you make a big mistake in a panic. Do you based on absolutely no information on trying to troubleshoot what caused the SRVs to not operate remotely manually....do you get it wrong and do the wrong thing. There is no warning light that pops up on the annunciator panel warning the actuator is dead because of a seal failure, but the automatic function is still good.

I mean, this is how the culture tries to so called save the industry...but screws the poor guys up in the control room during a terrible accident. I swear to god, you all answer to those lonely poor guys in the control room to not screw them when we are most at risk...

The PRB's tack with me with the Nov 6 response is the buna material is a approved and acceptable material ....while the VY NRC inspectors and Entergy says it is bad and it was wrong to put in the buna material. This is the approved material in PB 3 now amd PB2 before they put in new stuff. How come there is so much daylight between the PRB's explanation to me and the NRC Vermont Yankee inspector in his new inspection report?

I wished I could get some written explanation to me on the docket explaining how this actuator seal problem of mine positively influenced the NRC and the utility seeing how two NRC official has repeated this same story to me?

Bottom line the bureaucracy's safety processes failed miserably...I don't think you could catch the next one. I think you all operate the power plant through the rear view mirror.
Could these last two e-mails be placed on the docket?

thanks,
mike

***Ethics is situational depending on what side they think you are on. They never make an error that doesn't advantage the NRC or utility. There is just is no enforcement of ethics and truth telling to a petitioner or members of the public with the NRC...it's all spin. It's all "I belong to a special group and we got to protect our family's income stream", and everyone else is a mushroom. I think these guys intentionally went to the internet for Buna-n upper temperature research intending to grab a higher temp operability limit than actual. Buna-n has all sorts of formulation...meaning it can go on a almost infinite amount of lower and upper temperature limits. Think of all the NRC PhD engineers and material scientist this PRB has at its beck and call...and they go to the internet without any idea what Buna-n formulation that drives the service temperature range. They choose what temperature range they wanted totally disconnected from the facts. 

The new VY inspection report: "This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F) to 225 degrees F. 

The Peach Bottom PRB response to me: "The typical service temperature range for Buna-N material is -40F to +250F."

From: Kim, James
Sent: Tuesday, November 06, 2012 3:17 PM
To: 'Michael Mulligan'
Subject: Peach Bottom 2 and 3- 2.206 Petition on SRV

Mr. Mulligan,

Your petition dated October 13, 2012, was assigned to the Office of Nuclear Reactor Regulation (NRR) for review. My name is James Kim, the NRR project manager in the Division of Operating Reactor Licensing (DORL) and I have been assigned as a petition manager.

The NRR Petition Review Board (PRB) met on November 1, 2012, to consider your request for emergency shutdown of Peach Bottom Unit 2 and 3 based on 50.73(a)(2)(v)(D) for an event or condition that would prevent the fulfillment of a safety function. The PRB denied the request for immediate action because there was no immediate safety concern to plant, or to the health and safety of the public.

The typical service temperature range for Buna-N material is -40F to +250F. Prolonged exposure to higher temperatures could cause permanent hardening of the seal with resultant leakage. The normal operating environment for the Peach Bottom seals is well within the typical service temperature range. While the seals may experience temperature excursions higher than 250F in some post-accident scenarios, these excursions are not prolonged and would not contribute significantly to hardening. Further, the condition of the seals is monitored through routine surveillance leakage testing and the seals are replaced at a frequency specified to offset any age or temperature degradation that may be occurring.

In addition, the mechanical overpressure function of the SRV would not be impacted by even a complete failure of the ADS pneumatic thread seals. The mechanical overpressure function actuates the SRV based on a setpoint of reactor pressure. The pilot stage, or first stage, initiates the mechanical actuation of the 3-stage SRV. Essentially, when reactor pressure overcomes spring pressure, the SRV will mechanically actuate. The ADS function is pneumatically operated, and is physically separate from the mechanical setpoint. The pneumatic operator actuation is based on an electronic signal to a solenoid valve. The electronic signal can come from a remote manual control switch, or from an electronic ECCS actuation signal. Therefore, reactor vessel overpressure protection would remain intact.

In accordance with the 10 CFR 2.206 process, the Petition Review Board (PRB) is offering you an opportunity to address the PRB to provide any additional explanation or support for the petition before the PRB makes an initial recommendation. Please let me know whether you would like to address the PRB.

Thanks

James Kim

Project Manager, DORL

U.S. Nuclear Regulatory Commission

301-415-4125

E-mail: james.kim@nrc.gov

New Nov 8 
October 31, 2012
Mr. Christopher Wamser
Site Vice President
Entergy Nuclear Operations, Inc.
Vermont Yankee Nuclear Power Station
Vernon, VT 05354
SUBJECT: VERMONT YANKEE NUCLEAR POWER STATION – NRC INTEGRATED
INSPECTION REPORT 05000271/2012004

2 Annual Sample: Automatic Depressurization System Actuator Leakage

a. Inspection Scope
The inspectors performed an in-depth review of Entergy’s apparent cause analyses and corrective actions associated with the issue of actuator stem leakage on valves in the automatic depressurization system (ADS). Specifically, Entergy identified repeat occurrences of leakage around actuator stems during the 2009 and 2011 refueling outages. The inspectors determined whether Entergy had taken appropriate corrective actions to prevent recurrence of the leakage. Additionally, the inspectors reviewed an operability determination performed during the previous operating cycle following the discovery by Entergy that the seal installed on the ADS actuator stems did not meet environmental qualification requirements.

The inspectors interviewed plant personnel and reviewed test procedure results, condition reports, engineering evaluations, root cause analyses, and manufacturer data to assess Entergy’s problem identification, evaluation, and corrective action effectiveness with respect to the ADS actuator leakage. Specifically, the inspectors reviewed the documents to determine if the seal material used on the ADS actuator stems from 2008 to 2011 should be attributed as the root cause of the 2009 and 2011 stem leakage and to verify that the replacement seal material now installed was qualified for the expected environmental conditions. Additionally, the inspectors reviewed the TS, the UFSAR, and Vermont Yankee licensing documents to assess adverse impact due to the leakage with respect to design basis requirements. Finally, the inspectors evaluated whether the compensatory actions taken by Entergy following identification of the degraded condition provided reasonable assurance of operation of the ADS system during a design basis event and that Entergy’s conclusion that the system remained operable with the degraded condition was correct.


Findings and Observations
No findings were identified.
Entergy modified the actuator system in 2008. However, in consultation with the manufacturer, Entergy incorrectly concluded that the changes to the actuators were “like for like” replacement of components. Entergy failed to determine that the seal material for the actuator stem nut had been changed from Silicon to Buna-N. This change resulted in the temperature rating of the seal dropping from 400 degrees Fahrenheit (F)
to 225 degrees F. During the 2009 refueling outage, Entergy found nitrogen to be leaking from the actuators and determined the actuator stem nut seals were degraded. However, Entergy’s evaluation of the seal incorrectly concluded that the seal material was defective and a new Buna-N seal was installed. Entergy performed a subsequent evaluation of the seal material and determined that the material was Buna-N, not
defective, and the failure of the material was due to exceeding the thermal rating (225 degrees F) of Buna-N. Following identification that the seal material did not meet environmental conditions, Entergy performed an operability determination which concluded that the ADS system was operable, but degraded. These performance deficiencies were previously evaluated by the NRC in inspection reports 05000271/2011002 and 05000271/2011008.

The ADS system consists of four 3-stage safety relief valves with an actuator attached to the valves so that they can be opened using a nitrogen gas supply. The UFSAR states that nitrogen for the actuation of the valves is stored in accumulators installed in the drywell that are sized to ensure sufficient gas is available for the required number of ADS valve actuations following a design basis accident. This system was credited to
respond to design basis accidents and was required to be operable by TS. Additionally, nitrogen bottles were installed outside the drywell to actuate the ADS system following a design basis seismic event. The bottles were sized to allow operators to control reactor pressure using the ADS system for several days following the event. The inspectors determined that this portion of the system had not been evaluated or licensed for design basis accidents other than seismic events.

The inspectors reviewed the evaluations performed by Entergy that assessed past operability of the system prior to the 2011 refueling outage and the operability determination performed during the operating cycle. By crediting the use of the nitrogen bottles, Entergy determined that an adequate nitrogen supply would be available to respond to design basis accidents and events even with the additional loss of inventory
from the accumulator stem leakage. Entergy concluded that the ADS system had remained operable because there was adequate nitrogen inventory available. The inspectors questioned whether the bottles and piping would be available for all design basis accidents. In response, Entergy performed an evaluation and concluded the bottle system had been designed to survive the required design basis accidents and would be
available. The inspectors reviewed and concurred with the assessment, but noted that the evaluation was not done prior to crediting the system in the 2011 operability determination.

Finally, the inspectors evaluated the corrective action that replaced the Buna-N seal material with Viton®, a flouroelastomer, during the 2011 refueling outage. The inspectors found that this material had the same properties as the previously installed silicon seal,with a temperature rating of 400 degrees F, and met the environmental requirements for the system.