Millstone Nuke Pant Can't Keep Their Safety Doors Functional?

Date of first incident 12/12/2014

Simple Door Latch Sticking Problem At Millstone, Indicates A Bigger Problem?

Date of second incident  2/19/2015

Licensee Event Report (LER)2015-001-00

1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE

Millstone Power Station Unit 3 05000423 1 OF 3

4. TITLE

Unlatched Dual Train HELB Door Results in Potential Loss of Safety Function

5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVED

SFACILITY NAME DOCKET NUMBER

MO YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR F 05000

MONTH DY YA YER NUMBER NO.05 0

FACILITY NAME DOCKET NUMBER

02 19 2015 2015- 001 00 04 20 2015 05000

9. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

[1 20.2201(b) El 20.2203(a)(3)(i) [E 50.73(a)(2)(i)(C) [I 50.73(a)(2)(vii)

[1 20.2201(d) El 20.2203(a)(3)(ii) El 50.73(a)(2)(ii)(A) El 50.73(a)(2)(viii)(A)

1E 20.2203(a)(1) El 20.2203(a)(4) 0l 50.73(a)(2)(ii)(B) [I 50.73(a)(2)(viii)(B)

[_1 20.2203(a)(2)(i) El 50.36(c)(1)(i)(A) [1 50.73(a)(2)(iii) E- 50.73(a)(2)(ix)(A)

10. POWER LEVEL [E 20.2203(a)(2)(ii) El 50.36(c)(1)(ii)(A) [E 50.73(a)(2)(iv)(A) El 50.73(a)(2)(x)

ID 20.2203(a)(2)(iii) [E 50.36(c)(2) El 50.73(a)(2)(v)(A) El 73.71(a)(4)

10 20.2203(a)(2)(iv) El 50.46(a)(3)(ii) El 50.73(a)(2)(v)(B) El 73.71(a)(5)

1E 20.2203(a)(2)(v) [1 50.73(a)(2)(i)(A) El 50.73(a)(2)(v)(C) El OTHER

El 20.2203(a)(2)(vi) El 50.73(a)(2)(i)(B) [ 50.73(a)(2)(v)(D) Specify in Abstract below or in

___________________ ___________________________ ___________ ______________N__C___oNRCForm66A

12. LICENSEE CONTACT FOR THIS LER

LICENSEE CONTACT TELEPHONE NUMBER (Include Area Code)

William D. Bartron, Supervisor Nuclear Station Licensing 1 (860) 444-4301

13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT

CAUSE SYSTEM COMPONENT MANU- REPORTABLE CAUSE SY MANU- REPORTABLE

FACTURER TO EPIX C FACTURER TO EPIX

14. SUPPLEMENTAL REPORT EXPECTED 15. EXPECTED MONTH DAY YEAR

On February 19, 2015, with Millstone Power Station Unit 3 (MPS3) at 100% power and in operating mode 1, an individual on a fire watch rove processed through a dual train high energy line break (HELB) door normally and upon checking the door after passage the individual noted the door did not latch. The Control Room was promptly notified. An operator was dispatched to investigate. The operator exercised the door lock-set mechanism freeing the latch allowing the door to properly latch. The door was inoperable for approximately 7 minutes. Technical Specification 3.0.3 was entered and exited appropriately.

Although no definite failure mechanism was identified, the door was experiencing high usage due to compensatory fire watch roves entering/exiting the door. The door lockset mechanism was manually manipulated and then tested several times satisfactorily by maintenance personnel. Further, the door design has the door swing such that the HELB event would act to open the door when the lockset mechanism fails. Engineering is evaluating the adequacy of the preventive maintenance frequency. Additionally, a design change to reverse the door swing such that the HELB event would cause the door to close and thus not rely on the lock-set mechanism is being considered. Additional corrective actions are being taken in accordance with the station's corrective action program.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(v)(D), as a condition that could have prevented the fulfillment of a safety function for systems needed to mitigate the consequences of an accident.

1. EVENT DESCRIPTION:

On February 19, 2015, with Millstone Power Station Unit 3 (MPS3) at 100% power and in operating mode 1, an individual on a fire watch rove processed through a dual train high energy line break (HELB) door normally and upon checking the door after passage the individual noted the door did not latch. The Control Room was promptly notified. An operator was dispatched to investigate. The operator exercised the door lock-set mechanism freeing the latch allowing the door to properly latch. The door was inoperable for approximately 7 minutes. Technical Specification 3.0.3 was entered~and exited appropriately. This event was reported to the NRC pursuant to 10 CFR 50.72(b)(3)(v)(D), (NRC event # 50836) as a condition that could have prevented the fulfillment of a safety function for systems needed to mitigate the consequences of an accident. This event is also being reported pursuant to 10 CFR 50.73(a)(2)(v)(D), as a condition that could have prevented the fulfillment of a safety function for systems needed to mitigate the consequences of an accident.

BACKGROUND:

This door fulfills the requirements of a Security Door, Technical Requirement Manual Fire Door, C02 Door, Dual Train Protection Door, and a HELB Door. It is a key card actuated door with a crash bar on one side and a thumb latch on the other side. The door is part of the HELB barrier for the A and B 480 volt switchgear.

2. CAUSE:

Although no definite failure mechanism was identified, the door was experiencing high usage due to compensatory fire watch roves entering/exiting the door. Further the door design has the door swing such that the HELB event would act to open the door when the latch fails.

3. ASSESSMENT OF SAFETY CONSEQUENCES:

Given the low likelihood of an Auxiliary Building HELB occurring during the time the door was not properly latched (7 minutes), the consequences of this event was of very low safety significance.

4. CORRECTIVE ACTION:

Since this event occurred on the back shift, a maintenance technician was called in to inspect the door lock-set mechanism and affect any necessary repairs. The technician reported his inspection was satisfactory. He exercised the door lock-set mechanism from both the crash bar and the thumb release mechanisms approximately 30 times without any repeat indications of the latch sticking or not functioning. He also noted he tightened one screw on the mechanism that he found loose during this inspection. Continued exercises of the door mechanism after tightening the screw showed no difference in the smooth and proper operation of the door lockset mechanism. It was identified that the door was experiencing high usage due to compensatory fire watch roves entering/exiting the door. Equipment repairs have been completed eliminating the need for this high frequency fire rove activity. Additionally, the preventive maintenance for the door lock-set mechanism has been changed.

A design change to reverse the door swing such that the HELB event would cause the door to close and thus not rely on the lock-set mechanism is being considered.

Additional corrective actions are being taken in accordance with the station's corrective action program.

5. PREVIOUS OCCURRENCES:

* MPS3 LER 2014-004-00, Unlatched Dual Train HELB Door Results in Potential Loss of Safety

Function.

6. Energy Industry Identification System (EIIS) codes:

* Door- DR

* Switchgear - SWGR

Pilgrim's Fuel Vendor Depended On Non Safety Equipment For Core Safety

Basically generic letter 89-19 worries about a non safety system protecting a reactor core. Their concerns relate to: 

1) Redundant and diverse power supplies to these non safety system. 

2) The non safety systems components come up nuclear safety quality.

This document don't give any assurance Pilgrim meets these nuclear safety needs as stated in GL 89-19.

I had problems with River Bend scrams...they always were over feeding the vessel on scrams testing the MFP trips. You are not as habit allow to kept testing protection system...you are supposed control the situation. 

So how often is this going on at other plants and do these guys get to be intentionally evasive to skirt government safety rules. 

You know what bothers me the most, Pilgrim doesn't have total control of safety at the site. The fuel vendor and Pilgrim were like ships passing each in the night without seeing each other.  

April 13, 2015U.S. Nuclear Regulatory Commission 

ATTN: Document Control DeskWashington, DC 20555-0001SUBJECT: Correction of Information Provided in a Response to a Request forInformation Related to Feedwater Pump Trip Technical Specifications(TAC No. M474981)Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power StationDocket No. 50-293License No. DPR-35 

REFERENCE: Boston Edison Company letter to NRC, "Pilgrim Nuclear Power Station Response to Feedwater Trip Technical Specifications (TAC No.M474981), dated October 3, 1994 (BECo Ltr. #94-120) 

Dear Sir or Madam: 

This letter corrects information previously provided to the U.S. Nuclear Regulatory Commission (NRC) in response to a request to submit reactor vessel overfill protection Technical Specifications. 

Specifically, Pilgrim Nuclear Power Station (PNPS) staff identified that a Boston Edison Company letter to the NRC dated October 3, 1994 contained two incorrect statements concerning the function of the feedwater pump trip and how it was credited in the station's safety analysis. The correspondence was related to PNPS' response to Generic Letter 89-19, Request for Actions Related to Resolution of Unresolved Safety Issue A-47 "Safety Implication of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54(f) and directly responded to NRC's request for information on PNPS' plans to add a Technical Specification requirement for the feedwater pump trip.The October 3, 1994 letter stated, "It can therefore be seen

If the Feed water trip was safety related, does that mean they would have to up grade the safety quality of the Feedwater trip instrumentation. Was Pilgrim basically lying to the NRC to save money, and the NRC approving it being less than truthfully.

that this trip is a plant design feature incorporated to protect equipment, but has no nuclear safety-related function." It further stated "The MFP (Main Feedwater Pump] trip is not credited for fuel protection in any design basis accident or abnormal operational transient described in Pilgrim's Updated Final Safety Analysis Report." However, even though it was PNPS' intent not to credit the trip consistent with the assertions in the letter, while evaluating a condition reported in the PNPS corrective action program in late 2014, station staff discovered that the reload analysis conducted by the nuclear fuel vendor, in fact, had credited the trip in the plant-specific transient analysis for many cycles of operation, including the current cycle. Since the fuel vendor performs this analysis, PNPS staff were unaware that this assumed trip actually terminated the transient and reduced the maximum Critical Power Ratio impact calculated for the event. In Cycles 15 through 20, PNPS input to the transient analysis clearly stated that this trip was not an available feature for transient analysis use by the fuel vendor. 

Following discovery, PNPS took actions to ensure for the remainder of Cycle 20 that Turbine Control Valve Fast Closure and Turbine Stop Valve position scrams were active at power levels exceeding 25%, consistent with the use of thermal limits used by 3D Monicore to ensure thermal limit protection for analyzed transients. These actions were formalized in an Operations department Standing Order. PNPS also completed its evaluation of the cause for the discrepancy between PNPS' intent to not credit the high-level feedwater pump trip and the fuel vendor's calculation input assumptions. The cause was that there was no means specified for direct verification of inputs used by the fuel vendor. Additionally, PNPS evaluated the past three years of operation using the more restrictive Minimum Critical Power Ratio and Linear Heat Generation Rate limits calculated for Cycle 20 if the feedwater pump high-level trip is not credited and concluded that there were no thermal limit violations in the preceding three year time period using the more restrictive calculated limits. 

Corrective actions have been put into place to ensure discrepancies of this type do not occur in the future. PNPS has confirmed that the reload analysis for Cycle 21 does not credit the high level feedwater pump trip and that the fuel vendor has taken no deviations from PNPS provided inputs in performance of the transient analyses. Accordingly, going forward, beginning with Cycle 21, the statements in the October 3, 1994, letter will be complete and accurate as originally intended. 

Please contact me at (508) 830-8323 if you have any questions.

Limerick: Information On Testing US Reactor Vessels and Belgium Cracks

It is basically the blind leading the blind with my understanding this???

Big questions:

1) What does this mean and is it going on in the USA?  

"In carrying out tests related to theme 2 during the spring of 2014, a fracture toughness test revealed unexpected results, which suggested that the mechanical properties of the material were more strongly influenced by radiation than experts had expected."

Alliance for a Clean Environment response from NRC

May 5, 2015

Dr. Lewis Cuthbert

President

Alliance for a Clean Environment

1189 Foxview Road

Pottstown, PA 19465

Dear Dr. Cuthbert:

May 5, 2015

On behalf of the U.S. Nuclear Regulatory Commission (NRC), I am responding to your e-mail dated March 16, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15076A480), expressing concerns primarily about cracking, due to embrittlement, of the reactor pressure vessels at Limerick Generating Station, Units 1 and 2. I have included answers to your specific concerns in the enclosure to this letter. Thank you for contacting the NRC with your concerns.

'NRC Response to Concerns in March 16, 2015, E-Mail From the Alliance for a Clean Environment Regarding Limerick Generating Station, Units 1 and 2'

Background

This enclosure provides the U.S. Nuclear Regulatory Commission's (NRC's) response to concerns regarding Limerick Generating Station, Units 1 and 2 (Limerick), as discussed in the March 16, 2015, e-mail from the Alliance for a Clean Environment (ACE) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15076A480). ACE previously raised some of the same concerns in an e-mail dated November 9, 2014(ADAMS Accession No. ML 14321A054). The NRC staff provided a response in an e-mail dated December 8, 2014 (ADAMS Accession No. ML 14345A078).

Destructive Testing

ACE raised concerns regarding embrittlement of the Limerick reactors and asked if destructive testing of the reactors had been performed.

When a nuclear plant is operated, neutron radiation from the reactor core causes embrittlement of the reactor pressure vessel (RPV). Embrittlement refers to a decrease in the fracture toughness of RPV materials and affects the vessel materials in the section closest to the reactor fuel, referred to as the vessel's "beltline."

Section 50.60 of Title 10 of the Code of Federal Regulations (10 CFR), "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," requires compliance with the

Basically this is saying a plant is licensed with a vessel with these kinds of metallurgical properties based on 1960s technology and knowledge. No other inspections is necessary throughout the life of the plant.  

fracture toughness and material surveillance program requirements set forth in Appendices G and H to 1 O CFR Part 50. Compliance with the requirements of this rule, and the associated appendices, provides assurance regarding the structural integrity of the reactor coolant pressure boundary (RCPB) and, specifically, the RPV.

Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements," requires nuclear power plant licensees to implement RPV surveillance programs to monitor changes in the fracture toughness properties of ferritic materials in the RPV beltline region which result from exposure of these materials to

For a host of reasons, I don't think the coupon specimen surveillance is representative of the properties of the vessel. Basically a crack in the reactor vessel bypasses all the safety designs of the plant and a accident could be so severe, we need actual ultrasonic testing of the vessel. A stand in coupon testing is no longer a guarantor of safety. 

neutron irradiation. The RPV surveillance programs require destructive testing of material test specimens that are representative for the materials in the reactor. Two specific alternatives are provided for the design of a facility's RPV surveillance program to address the requirements of Appendix H to 10 CFR Part 50.

The first alternative, provided in Appendix H to 1 O CFR Part 50, is the implementation of a plant specific RPV surveillance program consistent with the requirements of American Society for Testing of Materials (ASTM) E 185, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."

The second alternative, provided in Appendix H to 10 CFR Part 50, is the implementation of an Integrated

You get it, always a stand in type testing, never inspection the real deal. Basically it is too expensive and disrupting the capacity factor.  

Surveillance Program (ISP). When a licensee uses an ISP, representative materials chosen for surveillance of a reactor are irradiated in one or more other reactors that have similar design and operating features.

As discussed in the NRC staff's safety evaluation for a Limerick amendment dated November 4, 2003 (ADAMS Accession No. ML032310540), Limerick, Units 1 and 2, have implemented the Boiling Water Reactor Vessel and Internals Project (BWRVIP) ISP as the basis for demonstrating compliance with the requirements of Appendix H to 1 O CFR Part 50.

To comply with Appendix H to 10 CFR Part 50, the entire fleet of operating U.S. reactors, including the Limerick RPVs, contain material specimens, representative of the materials in RPV beltline region, in surveillance capsules. These surveillance capsules are removed for

So they got specimens designed to be remove for testing from the vessel, but are never required to be tested. 

destructive testing of the material specimens as necessary. None of the surveillance capsules in the Limerick RPVs have been removed to date. In addition, as discussed in a Limerick license amendment dated April 8, 2011 (ADAMS Accession No. ML 110691095), based on the BWRVIP ISP, Limerick is not scheduled to remove any surveillance capsules in the future. 

Instead, the limiting weld and plate materials for the Limerick RPVs are monitored through representative material specimens that are exposed to irradiation in other boiling water reactors, as part of the BWRVIP ISP. The BWRVIP ISP was found acceptable by the NRC staff to satisfy the requirements of Appendix H to 10 CFR Part 50, during the period of extended operation for Limerick, as discussed in the Section 3.0.3.1.11 of NUREG-2171, "Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2" (ADAMS Accession No. ML 14276A156).

In summary, destructive testing has not been performed on the material specimens in the Limerick RPV surveillance capsules. Destructive testing has been and will continue

Ultrasonic like I want is not destructive and the destructive testing of drilled out specimens  I want is from permanently  shutdown plants.  Further I want ultrasonic testing of the permanently shutdown vessels...if no defects are discovered then no further inspection are needed.  

to be performed, for material specimens representative of the materials in the Limerick RPV, to meet the requirements in Appendix H to 10 CFR Part 50, as part of the BWRVIP ISP.

Material Fatigue Testing

ACE requested that the NRC "require independent 'material fatigue' testing of both Limerick Nuclear Plant reactors, with the results of testing immediately reported to the public."

All U.S. nuclear RPVs are designed and fabricated for operational cyclic stresses caused by all postulated loadings, including startup, shutdown, and scram events. Fatigue is explicitly evaluated as a part of the design

Based on 1960s technology and knowledge!!! 

process. Once the RPV is designed and fabricated and placed into service, licensees are required to track operational events, such as startups and shutdowns, to ensure they remain within their design bases with respect to fatigue. The NRC staff found that Limerick's fatigue program satisfied these requirements for the extended period of operation, as discussed in Section 4.3 of NUREG-2171, "Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2" (ADAMS Accession No. ML 14276A 156). As a result of satisfying these requirements, there is no demonstrated need for material fatigue testing at Limerick.

The NRC's regulations in 10 CFR 2.206 describe the

At least I got this right... 

petition process, which is the primary mechanism for the public to request enforcement-related action by the NRC in a public process. This process permits anyone to petition the NRC to take enforcement-related action associated with NRC licensees or licensed activities. Depending on the results of its evaluation, NRC could-modify, suspend or revoke an NRC-issued license or take any other appropriate enforcement related action to resolve a problem.

Although ACE's e-mail dated March 16, 2015, did not specifically cite the 10 CFR 2.206 process, it did request enforcement-related action (i.e., ACE's request to require material fatigue testing at Limerick). The NRC staff has previously offered the use of the petition process to address concerns where enforcement-related action was requested by ACE (e.g., NRC e-mail dated April 23, 2014 (ADAMS Accession No. ML 14129A184)). However, ACE has previously rejected use of the NRC's petition process to address its concerns (e.g., letter dated July 25, 2014 (ADAMS Accession No. ML 14216A339)). Nevertheless, the NRC staff considers the 10 CFR 2.206 petition process to be the appropriate process to address requested enforcement-related action. The NRC's petition process is available if ACE disagrees with the NRC's findings and has information the NRC did not consider in making its findings.

Belgium Reactor Operating Experience

ACE cited issues with cracking that had been reported in two reactor vessels in Belgium.

The NRC staff is well aware of this issue. Evaluations in Belgium and the U.S. demonstrated that because these many flaws are oriented nearly parallel to the direction of stress in the reactor vessel shell, they do not pose a significant safety concern. Additionally, it should be

Suggest isn't proof it is safe. Again this is a inference a vessel is safe, not acceptable proof it is safe. 

It is generially a new kind of corrosion/ hydrogen process...hydrogen collecting deep into the vessel leading to crack and flaws.

As far as I can see, nobody has taken a sample and cut into metal, thus nobody really what is going on.  

noted that information available from Belgium suggests that the flaws occurred as part of the initial fabrication process (i.e., flaws are not service-induced). As such, there is no indication that the flaws in the Belgium reactors are in any way related to fatigue damage.

Notes:

*** At least one reactor in Switzerland, another in Belgium and two in Spain have components produced by the same Dutch firm, Rotterdam Drydock Company, which has gone bankrupt since producing the equipment. The U.S. Nuclear Regulatory Commission said Friday it has been informed that 10 American reactors may have used the component in question, but it hasn't yetverified that information with U.S. nuclear operators.

***Doel 3/Tihange 2: new update

After a large number of flaw indications was discovered in the walls of the reactor pressure vessels (RPVs) of Doel 3 and Tihange 2 during a scheduled maintenance in the summer of 2012, the Belgian nuclear safety authorities (FANC and Bel V) decided that Electrabel had to submit a Safety Case to justify the restart of both reactors. Electrabel had to demonstrate specifically and convincingly in its Safety Case that the flaw indications in the walls of the RPVs do not compromise its structural integrity. 

After an analysis of the safety cases of both reactors, the FANC and Bel V decided on May 17, 2013 that Doel 3 and Tihange 2 could be restarted. Linked to this agreement, however, was the condition that Electrabel had to perform a series of medium-term actions to consolidate the hypotheses of its Safety Case. These actions were divided into three major themes:

1. The ultrasonic inspection technique of the RPVs: detection and measurement of hydrogen-induced flaw indications
2. Material properties of steel containing hydrogen flakes
3. Structural integrity of a rpv containing hydrogen flakes

The results of the actions on issues 1 and 2 provide the input for theme 3. 

"In carrying out tests related to theme 2 during the spring of 2014, a fracture toughness test revealed unexpected results, which suggested that the mechanical properties of the material were more strongly influenced by radiation than experts had expected." As a precaution both reactors were immediately shut down again.

Electrabel launched a test campaign to find an explanation for the unexpected test results.
At the same time, the licensee continued the execution of the medium termed-action plan. In the mean, this has led to the following results:

More accurate information about the flaw indications 

In February 2015, Electrabel completed the actions related to the theme of the ultrasonic inspection technique. 

This technique was originally designed for the control of the welding and the cladding of the RPV, but it also proved to be able to detect flaw indications in the wall of the RPV. Electrabel had to qualify the technique, i.e. prove that all hydrogen-induced flaw indications can be found and can be measured correctly using the ultrasonic inspection. By doing so, Electrabel found that the inspection procedure had to be slightly modified and that the detection threshold of the probes had to be lowered to ensure the proper detection of all flaw indications.
In 2014, a further inspection was carried out based on the improved procedure and the modified settings of the machine, resulting in the detection of a greater number of flaw indications than was measured in 2012 and 2013. This means that Electrabel now has to take into account 13 047 flaw indications for Doel 3 and 3149 flaw indications for Tihange 2 in its calculations. These additional flaw indications are similar to those which were previously considered and are located in the same area of the RPV. 

New sequence of material testing 

At the same time, Electrabel also continues its research on the material properties of the RPV and the unexpected results of the previous fracture toughness test. 

Currently a 4th irradiation campaign is being executed in the research reactor BR2 of the SCK, where, next to hydrogen-flaked samples of the French VB395 test material, other hydrogen-flaked samples of another test material of German origin are also being irradiated. The results of this irradiation campaign and of the subsequent material tests are expected by April 2015. 

New meeting of the international review board 

Electrabel provides the FANC and Bel V with results of ongoing tests and analyses on a regular basis. The Belgian security authorities need time to look into this new information and will continue their analysis during the first months of 2015. Therefore, they call in the help of international experts who are specialized in damage mechanisms caused by radiation and in mechanical resistance tests. This international expert panel (International Review Board) met for the first time in Brussels at the start of November 2014. The main conclusion of this meeting was that the methodology used by Electrabel was not yet sufficiently developed to make a well-grounded judgment. The international experts have formulated some suggestions for further actions and studies. Based on these suggestions and on the documents already analyzed, the Belgian security authorities have passed a series of additional demands and suggestions to Electrabel, so that the licensee can adjust its methodology and validate the underlying hypotheses of its arguments. 

In April 2015, the FANC will organize a new meeting of the international panel of experts to obtain their advice on the results of the new material tests and on the new data provided by Electrabel.

***Good Back Ground info: Report Activities in WENRA countries following the Recommendation regarding flaw indications found in Belgian reactors

Strange Sentence in New Brunswick Nuclear Plant Inspection Report???

I was looking for Diesel Generator problem, but found this? It is in a operability determination. 

Some kind of Ariva fuel problem going to impact Brunswick in 2015? How widespread is it in the USA?  

April 30, 2015: BRUNSWICK STEEM - NRC INTEGRATED INSPECTION REPORT NOS.: 05000325/2015001 AND05000324/2015001 

"Units 1 and 2, Atrium-10 fuel assembly load chain failure event at Chinshan – impact for operating cycle, March 30, 2015"

Chinshan nuclear plant is in Taiwan???

 

Jan 2014 Messages to Senator Shaheen about Seabrook

Jan 2014 Messages to Senator Shaheen about Seabrook

Bottom line, they only replaced a small percentage of the defective service water piping.  

Apr 28 2015

Service Water Piping Replacement for the Diesel Generator and Primary Component 

Cooling Water Heat Exchangers 

a. Inspection Scope The team reviewed EC 274172 which replaced degraded Plastisol-lined service water (SW) piping on the supply and return of the ‘A’ and ‘B’ emergency diesel generator (EDG) heat exchangers and degraded cement-lined SW piping on the supply side of the ‘A’ primary component cooling water (PCCW) heat exchanger. NextEra performed the modification to replace degraded SW piping with a corrosion resistant material to ensure long-term system pressure boundary integrity. NextEra replaced the carbon steel lined piping with AL-6XN, an austenitic stainless steel material, suitable for seawater service without the need for internal lining or protective coating.

I give Senator Shaheen a B+ on this. She got it all on the record. But she never had any active involvement or follow-up. The office never stayed on top of it. The NRC read her like a book..felt she would be happy just getting this on record. 

Writing to Senator Shaheen office and on the NRC docket
Remsburg, KristyFrom: Holmes, Sarah (Shaheen) [mailto:Sarah Holmes@shaheen.senate.gov]
Sent: Tuesday, January 07, 2014 10:09 AM
To: Dacus, Eugene
 Subject: FW: Seabrook issues
 Gene, Happy New Year- I hope you and your family had a nice holiday and are staying warm. Please see the note below from the Senator's constituent Mike Mulligan . He is very concerned about issues with deteriorating plumbing around the plant at Seabrook. could you please provide us with a formal reply to his concerns/allegations Thank you, please let me know if you need any additional information Kind Regards,

Sarah
From: Michael Mulligan [mailto:steamshovel2002@yahoo.com]
Sent: Wednesday, December 18, 2013 1:22PM
To: Holmes, Sarah (Shaheen)
Subject: Re: Seabrook issues
 Sarah, If you don't talk tough to these guys, they will allow Seabrook to deteriorate into a Vermont Yankee. This is very similar to their concrete problem where it took much more NRC action. I never have any confidentiality or anonymity needs what so ever! Mike  

Dear Senator Shaheen, These comments below are from the NH Union Leader newspaper by Seabrook Station's Local 555 Union President Ted Janis on Nov 26, 2013. They were negotiating a union contract.

"Their battle cry is 'natural gas' is killing us. We are not making the money we were making five years ago,'" said Jenis. "But it's hard for us to sit here and see these raises go out to management."

"This is a workplace that has been beaten down over the last few years," he said."

"There seems to be a total attitude change toward the workers from the corporate level."

"This is a workplace that has been beaten down over the last few years," he said."

Seabrook nuclear plant was brought on line in 1990 with cheap and non-corrosion resistant carbon steel service water piping. Within two years, piping integrity problems began showing up with pitting and local corrosion. And this problem has only gotten worst and it's running out of control as I write. It is corrupting the staff of this organization.

In 2011 they replaced a 30 year old 8 foot section of 24 inch {huge) width pipe qn the service water strainer by pass line. I think because of corrosion issues. It seemingly had a secret failure of some sort in 2011, as the NRC didn't disclose it in their most recent inspection report (2013001).

They replaced it with new carbon steel piping that was lined with so called super epoxy material Belzona. It failed within three years during August of this year. This is called progress. How do we know if the Belzona isn't going to clog again the emergency diesel generator cooling water orifices?

As it stands right now, the pipe only has a Band-Aid over the wound till the next outage {late spring 2014). Seabrook and the NRC will tell you they ultrasonically tested the hell out of this section of pipe once they detected it leaking. This device shows you the thickness of the metal piping. This is a nuclear plant and a crucial nuclear safety component...one which just failed mysteriously after 2 years ... why weren't they UTing the hell out the pipe before it leaked, as they knew the carbon steel service water piping was seriously corrosion prone? Why wasn't there an intense program to uncover any corrosion throughout the system and especially on the strainer bypass line that already failed?

Why didn't they catch the defect before it first leaked ... then catch it before the tinfoil thickness pipe wall burst and the leak got even bigger threatening the design of the plant? This is a matter of trusting them and their integrity. This is a matter of the NRC prodding them over and over again about following their procedures and using conservative engineering ethics. 

Seabrook through August this year didn't want to shut down over a pipe leak fearing a summer grid emergency with limited electricity and in a heat wave with expensive replacement electricity. Was this all about money and very little about public safety?

Seabrook obtained regulatory good will and forbearance to not shutdown to fix this dangerous leak even after botching the UT reading. The American Society of Mechanical Engineers sets the engineering standards that the NRC requires Seabrook to abide by. The ASME nuclear piping codes requires Seabrook to repair the pipe ... not a temporary repair like the NRC gave them permission to do. I bet you they want Seabrook to actually see the damage inside by eye to make certain they know what is going on ... not guessing. They could have kept this plant up at power if they first designed this plant prior to construction with sufficient extra service water capacity and flexibility in this area.

ASME Standards used in over 100 countries

ASME members provide engineering and technical expertise to policy makers in Congress, the White House Office of Science and Technology policy, and key federal agencies"

You get it, the poor initial plant design of the service water system sets Seabrook up to cry tike a baby to the NRC with the burdens of code and agency compliance. Your brother Pilgrim (Entergy) plant up north and the NRC doesn't have a care in the world with any "shutting the plant down in mid-cycle creates undue and unnecessary stress on plant systems, structures, and components" during the last year over all the multitudes of shutdowns and scrams they had caused by their poor plant upkeep and maintenance. These are nothing but excuses of convenience and it borders on another falsification in federal documents. 

It is impractical to complete a Code-acceptable repair to the identified SW leak at Seabrook Station without shutting the plant down. Shutting the plant down in mid-cycle creates undue and unnecessary stress on plant systems, structures, and components." (Sept 4, 2013)

I don't think these guys deserved any regulatory forbearance. They should have prepared their service water system years before for the rigors of summertime operations. This is how you protect the consumers from the potential of electricity shortages and maintain nuclear safety. Nuclear safety never comes from undeserving regulatory good will. Honestly, they need to spend big bucks to fix their service water. Course, the grid might be more vulnerable in winter time operations and with our limited natural gas piping capacity. They should have spent our good money towards the aims of making this plant reliable without regulatory nuclear safety forbearance during these critical summer months. Do you think the NRC's regulatory good will and forbearance will get us a reliable service water system for the rest of the life of this plant?

The service water cooling system supports all the reactor core cooling and the emergency diesel generator. This is certainly their Fukushima nuclear safety system. I spent considerable time talking to the NRC senior resident inspector and his boss the branch chief. The senior NRC resident frames the quality of the carbon steel service water piping system as "crap". Every professional in the field knows this is grossly inappropriate material for a salt water system.

Because of the poor quality of the carbon steel and its reckless susceptibility to early failure and all sorts of corrosion, they have lined (inside) portions of the piping with concrete, plastisol and Belzona. Seabrook began using plastisol in 1992 two years after first operation of the plant. The station was oblivious to the fact that the plastisol only has a service life of fifteen years. The NRC had to remind them of this. The brittle and pitted plastisol then sheeted off the piping and clogged a cooling orifice into a Fukushima emergency diesel generator. The machine didn't have enough cooling water and the station botched the "operability determination" over this twice. The big event you should be worrying about is any of the cement, plastisol or Belzona detaching from the inside of the piping and clogging up the water flow or damaging any of the valves.


I believe Seabrook knowingly falsified internal paperwork (prompt operability determination (POD)) the NRC depends on it to make a regulatory judgment. Is it safe to stay up at power orshould they shut down? Seabrook has made a string of bad "operability determinations" over the recent years and nothing the NRC does seem to turn these guys around into making accurate operability determinations. Seabrook had a leak in their service water system and they used a ultrasonic detector to measure the nature of hole in the pipe. They had information the hole was a very dangerous type which could leak big amounts of water ... but they put on the POD document it was a safe and stable hole. Within weeks the hole widened and leaked significant amount of water inside the plant threatening other safety equipment.

I questioned the Branch Chief and senior NRC inspector. They tell me Seabrook didn't adequately support their prompt operability determination (POD). This is a basic operation's safety function at a nuclear plant and they are all trained much on it. What are they even up at power for if they can't perform this simple determination? These guys are all extremely educated and there are many employees with advanced engineering degrees who ultimately make these determinations. It doesn't wash with these really smart and educated people making these kinds of simple mistakes. What they are really good at is covert-ups and playing stupid. Like I said, this plant has had lots of bum service water safety operability determinations lately ... why isn't the punishment cumulative? They had at least two stupid and inaccurate operability determinations with erratic cooling water flow indication to an emergency diesel generator. The dangerous brittle and over aged so called protective plastisol that sheeted off the sw pipes. What does it take to turn their hearts? What does it take to make accurate and safe operability determination? How will this faux stupidity end? This revolved around
an accurate UT scan of the pipe hole on day one and the staff blowing it with getting the information into the "Prompted Operability Determination".(wink wink)

I questioned the NRC inspector Mr. Cataldo about if it was a falsification of documents or if the NRC interpretation was Seabrook didn't adequately support the POD. How could these really smart and highly trained employees ever make that kind of simple mistake? He said, "Mike, it was just gross staff incompetence" surrounding the reading of the UT and getting the correct information into the POD." I still believe it was an intentional willful! falsification of documentation and the NRC is sweet talking this event into a poor support of the POD. But the great problem now is; why didn't the NRC accurately characterize this event as "gross Seabrook staff incompetence" surrounding the UT and the characterization of the hole in the NRC's inspection report? Does the NRC have two tiers of reporting, the prettified talk in the inspection reports for the community and the actual events at the plant?

I have real issues with the early failure of the new carbon steel piping and its super epoxy material Belzona. The nuclear industry is riddled with issues of improper heat treatment of metals and using the wrong type of metal. Remember, the old section of pipe failed mysteriously after 30 years. You get it, they never depressurized this section of piping. They never eyeballed the flaw inside the pipe and taken samples for sophisticated metallurgical analysis at an approved engineering laboratory. And these guys are terrible at guesswork. It could be related to microbial corrosion, electrochemical reactions with dissimilar metals and cement is a great worry.

"As previously stated in Section 7.2, the cause of the degradation is from localized corrosion.The typical corrosion rate used in Seabrook Station Service Water piping evaluations is 30 mils per year (mpy). However, the current identified wall defect resides in piping which was recently replaced during Refueling Outage 14 during April 2011, concluding that an accelerated (presently unknown) mechanism exists within the bounding area." (Sept 4, 2013)

The new carbon steel and the super coating failed after two years. The NRC's branch chief says the seawater in the bypass line is stagnant, but is open at the downstream connection. Nowhere in their documentation does it explain why the new section of piping failed so quickly other than to imply it is the same corrosion mechanism that destroyed the 30 year old first pipe. I think it is a new failure mechanism and other areas of the pipe could also fail quickly.

And believe me, there is no way to get an objective and independent interpretation of what went on here. The NRC and Seabrook have dog in this race with protecting their credibility ... you would need a recording (voice, visual) of the initial control room discussion about this hole with the NRC and then a recording of any subsequent discussion on this. Can you even imagine in a nuclear plant's safety cooling water piping, the NRC would allow the metal destruction mechanism to remain unknown?

Senior inspector Paul Cataldo told me he fought like hell with his bosses trying to get a bigger violation over this. He talked to me about the burdens the agency only gives him with a very limited weekly or month time budget with events at the plant. He put in a lot of time with this non violation. No overtime and certainly no paid overtime. I got the impression he thinks his bosses don't fully support him as they should and he is worried Seabrook's management doesn't respect him for his Federal oversight role at the plant.

I have called Seabrook's security gate and left a message asking to speak to an engineer about the sw strainer piping leaks. 

Better, somebody in the know within the operations department. I am still waiting for that call back?

Sincerely,
Mike Mulligan
Hinsdale, NH
(Ce11)16032094206

'The Popperville Town Hall'
http://steamshovel2002.blogspot.com/
On Friday, December 13, 2013 2:35 PM, Michael Mulligan wrote:
Sarah,
There we go!
Mike


On Friday, December 13, 2013 2:27PM, "Holmes, Sarah {Shaheen)" wrote:

Hi Mike,
Let's try this one more time.
Sarah

From: Holmes, Sarah (Shaheen)
Sent: Thursday, December 12, 2013 10:17 AM
To: 'steamshovel2002@yahoo.com'
Subject: Seabrook issues

Mike, 

Nice talking with you today. As we discussed, can you please send me a summary of the issues with leaks at Seabrook, that would be very helpful. I will forward along to the NRC with a cover note from the Senator asking for a response and additional information.

Kind Regards,
Sarah
Sarah Holmes
Special Assistant for Policy and Projects
Office of U.S. Senator Jeanne Shaheen
340 Central Avenue, Suite 205
Dover, NH 03820
P: (603) 750-3004
F: (603) 750-4046
sarah holmes@shaheen.senate.gov
s1gn up '"or
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"Their battle cry is 'natural gas' is killing us. We are not making the money we were making five years ago,'" said Jenis. 

"But it's hard for us to sit here and see these raises go out to management."

"There seems to be a total attitude change toward the workers from the corporate level."

ASME Standards used in over 100 countries

ASME members provide engineering and technical expertise to policy makers in Congress, the White House Office of Science and Technology policy, and key federal agencies"

"6. Burden Caused by Compliance

"As previously stated in Section 7.2, the cause of the degradation is from localized corrosion. The typical corrosion rate used in Seabrook Station Service Water piping evaluations is 30 mils per year (mpy). However, the current identified wall defect resides in piping which was recently replaced during Refueling Outage 14 during April 2011, concluding that an accelerated (presently unknown) mechanism exists within the bounding area." (Sept 4, 2013)

5

More Crap Service Water Piping At Seabrook!

“replace degraded SW piping with a corrosion resistant material: AL-6XN austenitic stainless steel material

SeabrookNuclear Station's "Crap" Service Water Piping System

Senator Shaheen's Help With Seabrook's Deteriorated Service Water Piping

Seabrook's Crap Service Water System

So now the truth comes out. I have gone over a lot of Seabrook’s documents with degraded service water pipes. I knew Seabrook used inappropriate piping materials leading to premature corrosion and pipe leaks. The Seabrook solution was to put a ban aid on the gaping pipe wound...the cement liner and plastisol lined pipe. The foreign material lined piping caused a flow reduction and the material has sloughed off causing blockage in other areas of the plant including the diesel generator.

"...replaced degraded Plastisol-lined service water (SW) piping on the supply and return of the ‘A’ and ‘B’ emergency diesel generator (EDG) heat exchangers and degraded cement-lined SW piping on the supply side of the ‘A’ primary component cooling water (PCCW) heat exchanger..."

Most troubling in all the Seabrook and NRC degraded service water documents...the NRC and Seabrook never admitted the root cause of their trouble was inappropriate pipe metal. Because if they admitted it, all the piping would have needed to be replaced.

Just to be clear, if all of the service water piping was made of AL-6-XN austenitic stainless steel, there never would have been a need of any plastisol or cement service water piping liners. The NRC says this piping was degraded, why wasn't this in a LER and why wasn't Seabrook forced into a tech spec LCO?  

It just  gets you wonder how much more safety information they withhold from the public because it reflects bad on the agency. 

April 28, 2015: SUBJECT: SEABROOK STATION, UNIT NO. 1 – NRC EVALUATION OF CHANGES,  TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONSTEAM INSPECTION REPORT 05000443/2015007 Service Water Piping Replacement for the Diesel Generator and Primary ComponentCooling Water Heat Exchangersa. Inspection Scope The team reviewed EC 274172 which replaced degraded Plastisol-lined service water (SW) piping on the supply and return of the ‘A’ and ‘B’ emergency diesel generator (EDG) heat exchangers and degraded cement-lined SW piping on the supply side of the ‘A’ primary component cooling water (PCCW) heat exchanger. NextEra performed the modification to replace degraded SW piping with a corrosion resistant material to ensure long-term system pressure boundary integrity. NextEra replaced the carbon steel lined piping with AL-6XN, an austenitic stainless steel material, suitable for seawater service without the need for internal lining or protective coating. 

The team reviewed the modification to determine if the design basis, licensing basis, or performance capability of the EDG, PCCW, and SW systems had been degraded by the modification. The team interviewed design engineers and reviewed design drawings and calculations to determine if the new SW piping met design and licensing requirements.
Additionally, the team reviewed non-destructive examination (NDE) results and associated maintenance work orders to determine if NextEra appropriately implemented the modification. The team performed several walkdowns of the accessible portions of the replaced SW piping, including a walkdown during a prolonged ‘A’ EDG run on March 5, 2015, to verify that NextEra had adequately implemented the modification, maintained pressure boundary integrity and configuration control, and had not impacted the function of other safety-related SSCs located in the vicinity. The team also reviewed corrective action CRs and the EDG, PCCW, and SW system health reports to determine if there were reliability or performance issues that may have resulted from the modification. Additionally, the team reviewed the 10 CFR 50.59 screen and engineering evaluation associated with this modification. The documents reviewed are listed in the Attachment.