Widespread Platumin Coating Flaking On Seating serfaces in BWR Safety Releif Valves.

Update Aug 19

Experiments: stellite, stellite B and now platium. 

Update Aug17

I spent about 2 hours talking to the Brunswick inspector and the public affairs people yesterday. I called the chairman's office and they got a Washington public affairs officer to talk to me. My aim with the chairman's office was gain me more influence than I deserve. If NRC officials know this emendated from chairman's office they will pay more attention to me. Basically went over the territory with BWR SRV testing failures. It is surprising the inspectors only look at this from their plant perspective...don't understand the big picture. My total objective with calling the chairman's was I want to take to their expert on Safety Relief Valves. He would have broad experience with all the troubles with the SRV. The PA guy said he would do the best that he could do. 

I got wind their is a secret white paper on SRV problems. This isn't disclosed to outsiders. I guess it is a NRC internal position paper, more about the internal agency deliberations. I asked the NRC to release the paper to me. I wonder how many secret white papers, over a host of other issues, there are within the NRC.

***The failure of the platinum to stick to the seating surfaces is a unreviewed safety problem. The problem is only looked at on a single plant bases. It is going on all over the place. Platinum is flaking off at untold BWR facilities including Browns's Ferry and Brunswick. Platinum is just the latest coating scam. This is leading to some eye popping magnitude pressure testing failures never seen before. Worst, the flaking is leading to steam cutting of the seats. This direly threatens the operability of the valves. Steam cutting the pilot valves does lead to a high probability of the valves to fling open at power and inability to shut the valve. It is also implicated in the valve not being about to open on demand and in remote. Pilgrim Plant had similar problems with SRVs. The plant grade by the NRC was severe downgraded and is directly implicated with their permanent closure. 

1. FACILITY NAME 2. DOCKET NUMBER 3. LER NUMBER
YEAR
Browns Ferry Nuclear Plant, Unit 2 05000260
SEQUENTIAL NUMBER
REV NO.
2019 - 002
NARRATIVE  

I. Plant Operating Conditions Before the Event
At the time of discovery, Browns Ferry Nuclear Plant (BFN) Unit 2 was in Mode 1 at approximately 91 percent power.
II. Description of Event  

A. Event Summary 

On May 29, 2019, NWS Technologies notified the Tennessee Valley Authority (TVA) with the as-found testing results of the thirteen Main Steam Relief Valves (MSRVs) [EIIS:RV], which were removed during the Spring 2019 Unit 2 Refueling Outage 20 (U2R20). Three MSRVs (BFN-2-PCV-001-0004, BFN-2-PCV-001-0018, and BFN-2-PCV-001-0019) had as-found lift settings which were outside of the +/- 3 percent band of their setpoints required for operability. 

Technical Specification (TS) 3.4.3, Safety/Relief Valves (S/RVs), requires twelve of the thirteen S/RVs to be operable for S/RV system operability. The three MSRVs were found to have been inoperable for an indeterminate period of time during the entire operating cycle between March 29, 2017, and March 2, 2019, and longer than permitted by TS 3.4.3. 

Previously, on April 27, 2018, MSRV BFN-2-PCV-001-0041 had been declared inoperable due to excessive leakage. It was removed and replaced during a midcycle outage in May 2018. Subsequent testing by NWS Technologies identified the as-found lift setting was outside the required +/- 3 percent band. At the time, BFN-2-PCV-001-0041 was the only known inoperable MSRV and the requirement of TS 3.4.3 (twelve of the thirteen S/RVs operable) was considered to be met.
Throughout this event, the two-stage MSRV pilot valves remained capable of maintaining the reactor pressure below 1375 pounds per square inch gauge (psig), which is the American Society of Mechanical Engineers (ASME) code limit of 110 percent of the vessel design pressure. The valves remained capable of performing their required safety function. 

The TVA is submitting this report in accordance with Title 10 of the Code of Federal Regulations 50.73(a)(2)(i)(B), as any operation or condition which was prohibited by the plant's TS.

B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event
There were no structures, systems, or components (SSCs) whose inoperability contributed to this event.

NARRATIVE 

C. Dates and approximate times of occurrences
Dates March 29, 2017
April 27, 2018
May 7, 2018
May 10, 2019
March 2, 2019
May 29, 2019  

Occurrence Unit 2 entered Mode 2 at beginning of cycle 20 (U2C20) 

BFN-2-PCV-001 -0041 declared inoperable due to excessive leakage  

Unit 2 entered Mode 4 for the replacement of BFN-2-PCV-001 -0041 

Unit 2 entered Mode 2 following the replacement of BFN-2-PCV-001 -0041 

Unit 2 entered Mode 4 for U2R20 

NWS Technologies notified the TVA with as-found testing results of the thirteen Unit 2 MSRV pilot valves removed during U2R20 

D. Manufacturer and model number of each component that failed during the event
The failed components were all Target Rock Corporation two-stage pressure control valves, model number 7567F. 

E. Other systems or secondary functions affected
No other systems or secondary functions were affected by this event. 

F. Method of discovery of each component or system failure or procedural error 

Failure of MSRVs BFN-2-PCV-001-0004, BFN-2-PCV-001 -0018, and BFN-2-PCV-001-0019 was discovered at NWS Technologies during their as-found testing of the thirteen MSRV two-stage pilot valves which were removed during U2R20. Failure of MSRV BFN-2-PCV-001-0041 was discovered based upon leakage estimates during plant operation and subsequent testing by NWS Technologies. 

G. The failure mode, mechanism, and effect of each failed component  

MSRVs BFN-2-PCV-001 -0004, BFN-2-PCV-001-0018, and BFN-2-PCV-001-0019 failed due to corrosion bonding to the valve seats as a result of the platinum anti-corrosion coatings flaking off. Two additional test lifts on each valve were within the acceptance criteria of +/- 3 percent of the required setpoint, indicating corrosion bonding caused each pilot valve to initially lift high.

NARRATIVE 

2019 - 002
Failure of MSRV BFN-2-PCV-001-0041 was caused by a delamination of a portion of the platinum anti-corrosion coating which lead to the leak. 

These guys crossed their quality and QA regulations. This platinum coating should have been proofed in a test stand with the same environment and duration as the SRVs in the reactor with a generous margin of safety.  The guys are doing unapproved experiments in a reactor.  

H. Operator actioans  

There were no operator actions associated with this event.
I. Automatically and manually initiated safety system responses 

There were no automatic or manual safety system responses associated with this event. 

Ill. Cause of the event 

A. Cause of each component or system failure or personnel error  

MSRVs BFN-2-PCV-001-0004, BFN-2-PCV-001 -0018, and BFN-2-PCV-001 -0019 failed above their setpoint band due to valve disc corrosion bonding to the valve seat as a result of the platinum anti-corrosion coating flaking off. In the case of MSRV BFN-2-PCV-001-0041, the loss of the platinum anti-corrosion coating led to leakage which affected its lift setpoint. 

B. Cause(s) and circumstances for each human performance related root cause 

No human performance related root causes were identified. 

IV. Analysis of the event 

BFN Unit 2 TS Limiting Condition for Operation (LCO) 3.4.3 requires twelve Operable S/RVs during Modes 1, 2, and 3. If one or more required S/RVs become inoperable, Required Action A. 1 requires entering Mode 3 within 12 hours and Required Action A.2 requires entering Mode 4 within 36 hours. S/RV Operability requires that S/RVs be within a +/- 3 percent band of their setpoint values in accordance with Surveillance Requirement (SR) 3.4.3.1. BFN Unit 2 has thirteen MSRVs to satisfy this requirement with margin.
When tested, the following four S/RVs were outside the allowable+/- 3 percent band. 

S/RV Number Set~oint Test Result Difference BFN-2-PCV-001-0004 1155 1215 +5.2% BFN-2-PCV-001-0018 1145 1197 +4.5% BFN-2-PCV-001-0019 1135 1214 +7.0% BFN-2-PCV-001-0041 1155 1230 +6.5%
00

NARRATIVE 

Prior to startup from U2R20, all thirteen MSRV pilot valves were replaced with refurbished valves which were certified to lift within +/- 1 percent of their setpoint. Operating Experience has shown that Target Rock two-stage MSRV setpoint drift is not a uniform, linear process. The corrosion bonding increases at a random rate. Without an accurate and reliable model for predicting or estimating the setpoint drift development, the point in time where the setpoint exceeded the +/- 3 percent limit cannot be reliably determined. Since this drift occurred during the operating cycle when the MSRVs were installed, MSRVs BFN-2-PCV-001-0004, BFN-2-PCV-001 -0018, and BFN-2-PCV-001-0019 are conservatively considered to be inoperable for an indeterminate period of time between March 29, 2017, and March 2, 2019. MSRV BFN-2-PCV-001 -0041 was considered to be inoperable from April 27, 2018 until entering Mode 4 on May 7, 2018, for replacement. More than one MSRV was considered to be inoperable during the entire operating cycle and longer than permitted by TS 3.4.3. 

TS LCO 3.0.4 states that when an LCO is not met, entry into a Mode or other specified condition in the Applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Because Unit 2 made Mode changes from a mid-cycle on May 10, 2018, while TS 3.4.3 was not met, Unit 2 was in violation of TS 3.0.4. 

V. Assessment of Safety Consequences 

System availability was not impacted by this event. The failure of the MSRV pilot valves to meet their TS 3.4.3 specified mechanical setpoints does not impact their remote-manual operation or activation through the MSRV Automatic Actuation Logic, since these operating modes and functions rely upon electrically signaled control air solenoids to open the MSRV pilot valves. 

The bounding maximum over-pressurization analyses are performed each fuel cycle to show that the requirements of the ASME code regarding overpressure protection are met. The analyses are performed specifically to show that the dome pressure TS limit of 1325 psig is not exceeded and that the vessel pressure does not exceed the limit of 1375 psig. In addition, the Anticipated Transient Without Scram (A TWS) pressurization analyses are also performed to demonstrate that the 1500 psig peak vessel pressure limit is not exceeded. 

For the ASME analysis, the existing analysis setpoint groupings conservatively bound the eleven lowest as-found MSRV opening setpoints; however, the highest as-found valve opening setpoint falls outside the bounds of the existing analysis valve groupings. Therefore, the limiting ASME overpressurization event, identified as the ASME with main steam isolation valve closure at 102 percent rated power/ 105 percent rated flow at coastdown, was re-analyzed based on the as-found lift settings. The re-analysis determined a maximum dome pressure of 1272 psig and maximum vessel pressure of 1305 psig, which are within the ASME limits.
For the A TWS analysis, the existing analysis valve setpoint groupings conservatively bound the eight lowest as-found MSRV opening setpoints; however, the four highest valve setpoints fall outside the existing analysis valve groupings. Therefore, the limiting A TWS overpressurization event, identified as the A TWS pressure regulator failed open at 100 percent rated power I 81 percent rated flow at beginning of cycle, was re-analyzed based on the as-found lift settings. The re-analysis determined a maximum vessel pressure of 1414 psig, which is within the ATWS limit.
TS Bases 3.4.3 states that the overpressure protection system must accommodate the most severe pressurization transient. The MSRVs remained capable of maintaining the reactor pressure below 1375 psig, which is the ASME code limit (110 percent of the vessel design pressure). The valves remained capable of performing their required safety function. Therefore, as defined in NEI 99-02, failure of the MSRV pilot valves was not a safety system functional failure. 

Based on the above, the TVA has concluded that sufficient systems were available to provide the required safety functions needed to protect the health and safety of the public... 

River Bend Isn't A World Class Nuclear Plant

Update

This is what a good plant looks like.
 
Comanche Peak 

PLANT STATUS
Unit 1 began this inspection period in coast down at 91 percent rated thermal power. The unit coasted down until April 20, 2019, when the unit was shut down to commence a refueling outage. On May 26, 2019, the unit began a reactor startup and reached rated thermal power on May 31, 2019. The unit remained at or near rated thermal power for the remainder of the inspection period.
 Unit 2 operated at or near rated thermal power for the entire inspection period. 

***There are on the cusp of a plant trip with the conditions of the employees and maintenance.  

River Bend Station began the inspection period shut down in Mode 5 for refueling outage RF-20.  The station performed a reactor startup on May 11, 2019, ending the outage.  The unit reached 87 percent power and remained there while waiting to perform a rod pattern adjustment.

On May 22, 2019, the operators reduced power to 36 percent due to an oil leak from a reactor recirculation pump transformer.  On May 24, 2019, the operators restored the reactor recirculation pump and resumed the reactor startup.  The unit reached 90 percent power and remained there while waiting to perform a rod pattern adjustment.

On May 28, 2019, the operators reduced power to 65 percent due to a feedwater heater leak.  The station commenced a normal reactor shut down on May 31, 2019, to repair the leak. 

On May 31, 2019, during the planned shutdown, the operators inserted a manual scram due to an unplanned loss of reactor feedwater.  The plant remained shut down following the scram to repair the feedwater heater leak. 

On June 6, 2019, the station performed a reactor startup.  On June 8, 2019, the unit reduced power from 76 percent to 36 percent due to another oil leak on the reactor recirculation pump transformer.  On June 11, 2019, the operators restored the reactor recirculation pump and resumed the reactor startup. 

On June 17, 2019, the unit reached 100 percent power.  Power remained at or near 100 percent for the remainder of the inspection period.

Semiannual Trend Review (IP Section 02.02) (1 Sample)
The inspectors reviewed the licensee’s corrective action program for potential adverse trends in the area of corrective maintenance that might be indicative of a more significant safety issue.  During the first half of the year, the inspectors observed an adverse trend in corrective maintenance.  The licensee performed maintenance on several safety--related or risk--significant components that subsequently failed, requiring rework.  The inspectors noted examples including a repeat failure of an intermediate range monitor, a repeat failure of a recirculation pump transformer, and inadequate maintenance of a recirculation pump motor bearing lubrication system.  These failures have been documented in the corrective action program.

Grand Gulf: Example Of a Bad Plant and Fuel Failures

This guy is the most dangerous plant is the USA. Check out the sickening issues with the employees and manager impeding the NRC without a harsh penalty. 

Generally when a plants declares they got fuel failure it implies to the public the plant is dirty. Being dirty inflames the minds of most people.This has severe public relations issues So that is why nobody openly discloses fuel failures.  

August 14, 2019  

SUBJECT: GRAND GULF NUCLEAR STATION – INTEGRATED INSPECTION REPORT 05000416/2019002 
 

PLANT STATUS Unit 1 began the inspection period at full power.  On April 6, 2019, operators reduced power to approximately 56 percent for control rod pattern adjustments.  Full power operation resumed on April 8, 2019.  On April 9, 2019, to allow for repair of a high pressure feedwater heater leak, power was reduced to approximately 66 percent.  Power was increased to approximately 94 percent on April 12, 2019, with the high pressure feedwater heaters isolated.  On April 16, 2019,

Sounds like fuel failures. I have reason to believe there is generic issues with fuel failure.  

power was reduced to approximately 65 percent to perform power suppression testing.  Following repairs to the feedwater heater and power suppression testing, the unit reached full power on April 21, 2019.  On May 12, 2019, due to a significant loss of plant service water, operators manually shut down the unit.  Following repairs to the plant service water electrical distribution system, the unit was started up on May 19, 2019, and full power was achieved on May 30, 2019.  On June 26, 2019, in order to support replacement of a recirculation flow control valve controller, operators reduced power to approximately 96 percent.  On June 27, 2019, the unit reached full power and remained at or near full power for the remainder of the inspection period. 

 OTHER ACTIVITIES – TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL 92723 - Follow Up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period (1 Sample) From December 9, 2016, through February 7, 2019, the NRC issued 10 Severity Level IV traditional enforcement violations associated with impeding the regulatory process.  These violations, which are listed below, were issued in the following NRC Inspection Reports: 

• 05000416/2016007 (ADAMS Accession No. ML16348A222), dated December 9, 2016  • 05000416/2017002 (ADAMS Accession No. ML17220A152), dated August 3, 2017 • 05000416/2017007 (ADAMS Accession No. ML17339A154), dated December 1, 2017 • 05000416/2018004 (ADAMS Accession No. ML19038A437), dated February 7,2019
 Specifically, the violations included:
 (1) Failure to Obtain NRC Approval for Changes to the Reactor Protection System, NRC-identified NCV 05000416/2016007-02;
 (2) Failure to Obtain NRC Approval for Changes to Diesel Generator Trips and Flood Mitigation Strategy, NRC-identified NCV 05000416/2016007-03;
 (3) Failure to Evaluate Delaying Inspection of Diesel Fuel Oil Storage Tank, NRC-identified NCV 05000416/2016007-04; 

(4) Failure to submit an annual effluent report in accordance with 10 CFR 72.44(d)(3); licensee-identified NCV documented in Inspection Report 05000416/2017002;
 (5) Failure to report the results of the visual inspections of all accessible, susceptible locations of the steam dryer to the NRC staff within 60 days following startup in accordance with License Condition 2.C(46)(f), licensee-identified NCV documented in Inspection Report 05000416/2017002;
 (6) Failure to submit a long-term steam dryer inspection plan based on industry operating experience along with the baseline inspection results for NRC review and approval in accordance with License Condition 2.C(46)(g), licensee-identified NCV documented in Inspection Report 05000416/2017002;
 (7) Failure to notify the NRC within 4 hours of the occurrence of any event or condition that resulted in actuation of the reactor protection system when the reactor was critical in accordance with 10 CFR 50.72(b)(2)(iv)(B), licensee-identified NCV documented in Inspection Report 05000416/2017002;
 (8) Failure to Update the Final Safety Analysis Report, NRC-identified NCV 05000416/2017007-04;
 (9) Failure to make a timely event report for an event or condition that could have prevented fulfillment of a safety function (accident mitigation), licensee-identified NCV documented in Inspection Report 05000416/2018001;
 (10) Failure to Update the Updated Final Safety Analysis Report, NRC-identified NCV 05000416/2018004-01.
 In April 2019 the inspectors performed Inspection Procedure 92723, “Follow-up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period.”  The inspectors reviewed the licensee’s cause evaluations and corrective actions associated with these issues in order to determine whether the licensee’s actions met the Inspection Procedure 92723 inspection objectives, which include:  (1) providing assurance that the cause(s) of multiple Severity Level IV traditional enforcement violations are understood by the licensee; (2) providing assurance that the extent of condition and extent of cause of multiple Severity Level IV traditional enforcement violations are identified; and (3) providing assurance that licensee corrective actions to traditional enforcement violations are sufficient to address the cause(s).
 INSPECTION RESULTS Observation:  Follow-up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period 92723 Background
 The inspectors noted that from January 1 to December 31, 2013, the NRC issued three Severity Level IV traditional enforcement violations associated with impeding the regulatory process, as documented in Annual Assessment Letter 05000416/2013001 (ADAMS Accession No. ML14063A338), dated March 4, 2014.  Inspection Procedure 92723 was performed in response to four Severity Level IV traditional enforcement violations, as documented in Inspection Report 05000416/2014005 (ADAMS Accession No. ML15033A479), dated February 2, 2015.  These violations involved accuracy and completeness of the Updated Final Safety Analysis Report, accuracy and completeness of information in the license renewal process, failure to report changes to the emergency plan, and failure to obtain a license amendment as required.

 The inspectors also noted that from January 1 to December 31, 2015, the NRC issued seven Severity Level IV traditional enforcement violations associated with impeding the regulatory process, as documented in Annual Assessment Letter 05000416/2015006 (ADAMS Accession No. ML16061A361), dated March 2, 2016.  Inspection Procedure 92723 was again performed in response to these Severity level IV traditional enforcement violations, as documented in NRC Inspection Report 05000416/2016003 (ADAMS Accession No. ML16315A372), dated November 10, 2016.  These violations involved two failures to update the Final Safety Analysis Report; a failure to maintain a safety-related cable tray overfill analysis record; an incomplete and inaccurate response to NRC Bulletin 88-04; a failure to obtain a license amendment; and two failures to make required event notifications or reports.
 Considering these facts, the inspectors sought to understand why failures to update the Final Safety Analysis Report, failures to make required reports, and failures to obtain license amendments were issues that have continued to be identified and documented as recently as 2019.
 Licensee Evaluation
 The inspectors reviewed the licensee’s collective evaluations (completed at the “Condition Analysis” level) associated with the 10 CFR 50.59 issues and reporting issues, which included 10 CFR 50.71(e) and reporting issues.  The inspectors also reviewed individual/combined evaluations associated with each of the 10 non-cited violations.  The inspectors were not able to review in detail the latest revisions of the collective evaluation associated with Condition Report CR-GGN-2019-01002 (associated with 50.59 issues) and the adverse condition analysis associated with Condition Report CR-GGN-2018-01595 (associated with a late 8-hour notification) because preliminary versions of the evaluations were provided during the inspection.  Evaluations reviewed included:

• CR-GGN-2019-01002, “50.59 Evaluations not Conducted as Required,” which evaluated the causes of 10 CFR 50.59 related violations (three issues with four total examples).  CR-GGN-2017-01483 had previously been completed to address this group of issues.  The individual issues included within this collective evaluation were evaluated by the following cause evaluations:
 1. CR-GGN-2019-01007, dated March 14, 2019, which revised the licensee’s evaluations associated with CR-GGN-2016-09757 (two revisions); these cause evaluations were associated with violation (3), as listed above.
 2. CR-GGN-2019-01185, dated March 20, 2019, which revised the licensee’s evaluations associated with CR-GGN-2016-08298 (two revisions); these cause evaluations were associated with violations (1) and (2), as listed above.

• CR-GGN-2019-01003, “Failure to Submit Reports to the NRC,” which evaluated the causes of failures to make reports to the NRC (five issues/examples) and failures to 

update the Final Safety Analysis Report (two issues with four total examples).  CR-GGN-2017-07970 had previously been completed to address this group of issues.  The individual issues included within this collective evaluation were evaluated by the following cause evaluations:
 1. CR-GGN-2017-03404, dated May 22, 2017, and later revised on March 1, 2019; this cause evaluation was associated with violations (5) and (6), as listed above
 2. CR-GGN-2017-12284, dated February 20, 2018, and later revised on March 1, 2019; this cause evaluation was associated with violation (8), as listed above
 3. CR-GGN-2018-01595, dated March 27, 2018, and later revised March 1, 2019; this cause evaluation was associated with violation (9), as listed above
 4. CR-GGN-2019-01047, dated March 22, 2019; this cause evaluation was associated with violation (10), as listed above
 5. CR-GGN-2019-01390, dated March 1, 2019, which elevated the level of evaluation of CR-GGN-2017-03092, which was documented on March 27, 2017, at the “C” significance level; these were associated with violation (4), as listed above
 6. CR-GGN-2019-01391, dated March 14, 2019, which elevated the level of evaluation of CR-GGN-2017-03331, which was documented on April 4, 2017, at the “C” significance level; these were associated with violation (7), as listed above 

 Assessment
 Considering all the cause evaluations and relevant condition reports the inspectors identified 16 examples where the responsible manager failed to ensure all available and relevant information was acquired by reviews of pertinent industry events in accordance with step 4[2]e of Procedure EN-LI-118, “Cause Evaluation Process,” Revision 23.  The licensee documented Condition Reports CR-GGN-2019-02684, CR-GGN-2019-02716, CR-GGN-2019-02733 to address these concerns.  These examples included:

• Failure to review all relevant condition reports (such as CR-GGN-2015-05057) in the relevant internal operating experience section of the adverse condition analyses associated with CR-GGN-2019-01002 and CR-GGN-2019-01003

• Failure to review 17 relevant condition reports in the relevant internal operating experience section of the adverse condition analysis associated with CR-GGN-2017-12284

• Failure to include relevant internal operating experience detailed review attachments with the adverse condition analyses associated with CR-GGN-2017-12284, CR-GGN-2017-03404, CR-GGN-2019-01391, and CR-GGN-2018-01595
 

• Failure to perform internal operating experience repeat event reviews with the adverse condition analyses associated with CR-GGN-2018-01390, CR-GGN-2019-01047, and CR-GGN-2019-01391

• Failure to perform an adequate review of external operating experience with the adverse condition analyses associated with CR-GGN-2017-03404 and CR-GGN-2018-01390

• Failure to include relevant external operating experience detailed review attachments with the adverse condition analyses associated with CR-GGN-2017-12284, CR-GGN-2017-03404, CR-GGN-2019-01391, and CR-GGN-2018-01595
 The inspectors also identified other issues during the review:

• The inspectors identified that Attachment 9.2, “UFSAR Change Process Flow Chart,” of Procedure EN-LI-113-01, “Updated Final Safety Analysis Report Change Process,” Revision 3, included numerous incorrect procedure step references.  The licensee documented CR-GGN-2019-02674 to address the concern.

• The inspectors identified that a corrective action was not implemented to address a missed barrier identified as a factor in adverse condition analysis CR-GGN-2018-01595.  This was associated with providing a procedure barrier to operators and licensing personnel when completing reportability evaluations.  The licensee initiated CR-GGN-2019-02690 to address this observation.

• The licensee initiated CR-GGN-2019-02732 in response to the inspectors' questions on licensing basis document change request (LBDCR)/Updated Final Safety Analysis Report tracking not being defined.
 Conclusions
 The inspectors could not conclude that the inspection objectives were met because substantially relevant operating experience was not adequately considered and evaluated by the licensee when determining causes, extent of conditions, extent of causes, or corrective actions.  Specifically, the inspectors identified that the conditions, scope, and causes associated with the two main collective evaluations and the previously performed root cause CR-GGN-2015-05057, which was associated with previous examples of incorrect maintenance of design and license basis documents, revealed substantive commonalities involving implementation and control of design change and license basis document changes; however, the CR-GGN-2015-05057 root cause analysis was not appropriately evaluated as relevant operating experience in the current evaluations.  Corrective actions to prevent recurrence of root causes would be expected to prevent recurrence of the same performance deficiency.
 Additionally, the licensee continues to identify LBDCR performance issues.  Specifically, engineering changes with LBDCRs are still being evaluated to ensure all LBDCRs from 2000-2013 are appropriately tracked and coordinated.  Also, a trend of incorrectly implemented LBDCRs during a software conversion is still being evaluated to determine causes and to ensure that conversion issues have been identified.  

Finally, the inspectors could not conclude that the inspection objectives were met because the collective evaluations and additional causal analyses were being revised as of the completion of the inspection.  As a result of this inspection, the licensee voluntarily determined that the conclusions of its collective evaluations needed to be reconsidered.  The licensee planned to re-evaluate and revise its collective evaluations (CR-GGN-2019-01002 and CR-GGN-2019-01003) associated with the 10 violations and review the previous CR-GGN-2015-05057 root cause analysis to ensure that the causes, extent of conditions, extent of causes, and corrective actions were adequate.  The licensee initiated CR-GGN-2019-02717 to document that adverse condition analyses associated with CR-GGN-2019-01002 and CR-GGN-2019-01003 did not adequately evaluate issues with configuration control of design and licensing basis documentation.  The licensee informed the inspectors that a cause evaluation will be completed to address this concern.

What The Hong Kong Experiment Portends For All Future Uprising

Will the new weapons of choice for all protestor be hand held lasers. Can you tell of it is a harmless laser or one that can do much damage to the eyes. These god damn ultra modern societies? 

Did you see the pictures of the multitudes of protestors laser dots on the Hong Kong. How will this play out in all future protest? 

Will it be coming to a theater near you( US protest)? 

Can you tell the deference between a pointer laser or a high powered rifle laser gunsight?        

Brunswick SRVs: Another Half Ass Fix Didn't Work

They should have worked this out in comprehensive testing in a similar environment.

Inadequate Procedure Resulted in Inoperable Safety Relief Valves Cornerstone Significance Cross-Cutting Aspect
Report Section
Mitigating Systems

Green NCV 05000324,05000325/2019002-03  Closed
[H.1] - Resources 71153
A self-revealed Green non-cited violation (NCV) of TS 3.4.3, “Safety/Relief Valves (SRVs)”, was identified when the licensee discovered two of the 11 safety relief valves (SRVs) asfound lift set points were outside of the +/- 3 percent pressure band required for their operability. Description:  Licensee event report (LER) 05000325/2018-003-00 was associated with two of the 11 SRVs as-found setpoints being outside of the +/- 3 percent pressure setpoint band required for their operability. This was discovered on June 11, 2018, following as-found

17

testing results conducted on all 11 SRVs that were removed during the refueling outage. The licensee determined that the out of tolerance lift pressure of the two SRV pilot discs was due to corrosion bonding of the pilot disc to the valve seat. The licensee determined that these two SRVs were inoperable for an indeterminate period of time from March 23, 2016, when the unit entered Mode 2 (beginning of operating cycle) to March 3, 2018, when the unit entered Mode 4 (beginning of refueling outage). The inspectors reviewed the licensee event report and determined that the report adequately documented the summary of the event including the cause and potential safety consequences.   Corrective Actions:  The licensee replaced all eleven of the Unit 1 SRV pilot valves with refurbished valves during the Spring 2018 Unit 1 refueling outage. Corrective actions have been completed which included revised procedures and work instructions to ensure a more consistent surface preparation and proper quality checks of SRV pilot disc surface conditions prior to applying the platinum coating. Additionally, the licensee is part of the industry-led boiling water reactors owners group which is researching several new corrective actions aimed to eliminate the SRV setpoint drift issue due to corrosion bonding of the pilot valves.   Corrective Action References:  NCR 2212540 Performance Assessment:   Performance Deficiency:  Failure to provide an adequate procedure and work instructions with sufficient detail to ensure consistent pilot valve surface preparation prior to platinum coating was the Performance Deficiency. Specifically, the inadequate procedure led to a degraded platinum coating on the pilot valve seating surfaces that allowed corrosion bonding of the SRV pilot discs to pilot seats which resulted in the out of tolerance lift setpoints of the two SRVs.
 Screening:  The performance deficiency (PD) was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).   Significance:  The significance of this finding was evaluated using IMC 0609, Appendix A, “The Significance Determination Process (SDP) for Findings At-Power,” dated June 19, 2012.  This finding was determined to be Green, very low safety significance, because all of the associated mitigating systems screening questions were answered No.  Additionally, the licensee’s Cycle 21 reload safety analysis report determined that the SRVs remained capable of performing their safety function to prevent over-pressurization of the reactor coolant system (RCS).
 Cross-Cutting Aspect:  H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. The inspectors determined the finding had a cross-cutting aspect of Resources in the Human Performance area because adequate procedures and work instructions were not provided to ensure an adequate application of the platinum coating of the SRV pilot valve seats.
 Enforcement:   Violation:  Brunswick Steam Electric Plant, Unit 1 Limiting Condition of Operation (LCO) 3.4.3, ”Safety/Relief Valves (SRVs)” required the safety function of ten (10) SRVs shall

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be operable in Modes 1, 2 and 3. When the LCO was not met, Condition A was applicable which required that with one or more required SRVs inoperable, that the unit be in Mode 3 in 12 hours and Mode 4 in 36 hours. Contrary to the above, two required SRVs were inoperable from March 23, 2016, to March 3, 2018, and Unit 1 was not placed in Mode 3 and Mode 4 in 12 hours and 36 hours, respectively.
 Enforcement Action:  This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy...

At A Boy Entergy, For First Time This Summer All Region IV Plants are at Power

Hindale-Brattleboro Route 119 Bridge Construction Project

I talked to the project manager yesterday. I had been talking to many people yesterday about we need a  two way blog or Facebook page concerning the new bridge project. I am concerned with the transparency of the commissions.

As far as the project manager, he said the property issues with be clear up spring next year. The bridge construction project will kick off beginning next fall with pier work.  

construction

Bam, ANO 2 IS Up At Power And Waterford Took A Forced Shutdown

Entergy has had one plant down out of their southern fleet all summer long.  Collectively they have had a poor record with protecting their grid this summer.

So lazy and incompetent Waterford found a crack in the charging pump discharge header piping and put off emediately fixing it. Then next check of the area it worsened leading to the shutdown. Is their predicting piping flaw growth any good.

Or worst, the first flaw test was inaccurate?    

Power Reactor Event Number: 54191 Facility: WATERFORD Region: 4     State: LA Unit: [3] [] [] RX Type: [3] CE NRC Notified By: BRIAN BUSCHBAUM HQ OPS Officer: DONALD NORWOOD Notification Date: 07/31/2019 Notification Time: 16:20 [ET] Event Date: 07/31/2019 Event Time: 12:06 [CDT] Last Update Date: 07/31/2019 Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(i) - PLANT S/D REQD BY TS 50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION Person (Organization): JEREMY GROOM (R4DO) Unit SCRAM Code RX Crit Initial PWR Initial RX Mode Current PWR Current RX Mode 3 N Y 100 Power Operation 65 Power Operation Event Text TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO INOPERABLE BORON INJECTION FLOW PATHS AND CHARGING PUMPS "On July 31, 2019, at 1206 CDT, Waterford 3 commenced initiation of a plant shutdown as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3. Prior to this, on July 31, 2019, at 1108 CDT, the boron injection flow paths were declared inoperable in accordance with LCO 3.1.2.2, 'Flow Paths - Operating,' and the charging pumps were declared inoperable in accordance with LCO 3.1.2.4, 'Charging Pumps-Operating.' This was due to visual examination identifying that propagation had progressed on a previously identified flaw on piping upstream of the header supplying the charging pumps. TS LCO 3.0.3 was entered due to the action statements of LCOs 3.1.2.2 and 3.1.2.4 not being met. LCO 3.0.3 requires that action shall be initiated within one hour to place the unit in a mode in which the specification does not apply by placing it in hot standby within the next 6 hours and cold shutdown within the next 30 hours. At 1206 CDT, Waterford 3 commenced direct boration to the reactor coolant system. "This condition meets the reporting criteria of 10 CFR 50.72(b)(2)(i) due to the initiation of plant shutdown required by Technical Specifications and 10 CFR 50.72(b)(3)(v)(A) and (D) due to an event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (A) shutdown the reactor and maintain it in a safe shutdown condition and (D) mitigate the consequences of an accident."

Hinsdale-Brattleboro Route 119 Bridge

With new Hinsdale-Brattleboro bridge on horizon, committee plans old spans' future

By Meg McIntyre Sentinel Staff
Jul 28, 2019 Updated 34 min ago 

With the completion of a new bridge connecting Hinsdale to Brattleboro still several years away, a local subcommittee is already hard at work to determine the future of the existing bridges between the two towns.

Remember this wasn't the original names of the bridges. I think it had no name. I wonder when the new bridge inspection is comming. Remember it was red listed last year. It should have been on the highly politicalized list for many decades.    

Named after Charles Dana and Anna Hunt Marsh, the Route 119 bridges are Pennsylvania truss-style spans built in 1920 and rehabilitated in 1988. A new bridge is set to be built several hundred feet downstream, with construction slated to begin next year and finish in 2023.

A project to replace the bridges has been included in New Hampshire’s 10-year transportation improvement plan since fiscal year 1994, with its start date being delayed several times.

A subcommittee comprising representatives from both communities has been meeting to discuss the existing bridges for a little more than a year, according to J.B. Mack, principal planner for the Southwest Region Planning Commission. Mack said he’s facilitating the committee process to gather feedback and input from community stakeholders.

An environmental assessment document created as part of the state’s overall bridge replacement project identified the existing bridges as a historic “extension of downtown Brattleboro,” according to Mack. That document proposes using the bridges for pedestrian and bicycle use.

I never seen the term "Hinsdale Island" before. It was once termed as Island Park. It seems the floating name of the island is a open question. As with everything in this area very little public information gets to the public including metting times and dates.  

The subcommittee is also looking at the future of Hinsdale Island, which connects the two bridges, Mack said. After speaking with the N.H. Department of Environmental Services, it appears that extensive construction on the island would not be possible because it is in the floodplain zone and is primarily composed of alluvial deposits.

“They’ve talked about maybe putting a gazebo or having a park out there, but in terms of intensive development, it doesn’t look like it’s a good idea,” Mack said.

The group is also looking at the possibility of creating a trail loop connecting the old and new bridges that would allow pedestrians and bicyclists to cross on one bridge and return on the other, he said.

“We’ve talked about working with local groups to do some interpretive signage where basically people can learn more about the history of the river, or people that used to live on the river,” he said. “The native Abenaki has been one idea, and sort of the industrial past of Brattleboro.”

Funding is still being secured for the refurbishment of the existing bridges, according to Mack, but the N.H. Department of Transportation has applied for a federal BUILD Grant for the project.

The subcommittee has met three times so far, according to Mack. At this point, there is no concrete timeline for the group to complete its work, he said, but members plan to continue brainstorming before ultimately bringing their proposal to the community for feedback.

“We’ve got quite a bit of time to do our planning, but at the same time, it’s so close especially to Brattleboro that it’s an important resource, and we’ve got to start planning for it as soon as possible,” he said.

The Existing Bridges Subcommittee plans to meet again in August, Mack said, but a specific date and time have not yet been set. Once scheduled, information about the next meeting, along with general information about the wider bridge replacement project, will be available at www.nh.gov/dot/projects/hinsdalebrattleboro12210.

Nuclear Agency Considers Reducing Inspections

U.S. NEWS

07/17/2019 06:58 am ET

Nuclear Agency Considers Reducing Inspections Of Reactors
Opponents denounce the cost-cutting move as a threat to public safety.
 

Associated Press 

The construction site of Vogtle Units 4 at the Alvin W. Vogtle Electric Generating Plant is see on March 22, 2019 in Waynesboro, Ga. The Nuclear Regulatory Commission will look at cutting back on inspections of the country’s nuclear reactors. 

WASHINGTON (AP) — Nuclear Regulatory Commission staff is recommending that the agency cut back on inspections at the country’s nuclear reactors, a cost-cutting move promoted by the nuclear power industry but denounced by opponents as a threat to public safety.

The recommendations, made public Tuesday, include reducing the time and scope of some annual inspections at the nation’s 90-plus nuclear power plants. Some other inspections would be cut from every two years to every three years.

Some of the staff’s recommendations would require a vote by the commission, which has a majority of members appointed or reappointed by President Donald Trump, who has urged agencies to reduce regulatory requirements for industries.

The nuclear power industry has prodded regulators to cut inspections, saying the nuclear facilities are operating well and that the inspections are a financial burden for power providers. Nuclear power, like coal-fired power, has been struggling in market completion against cheaper natural gas and rising renewable energy. 

While Tuesday’s report made clear that there was considerable disagreement among the nuclear agency’s staff on the cuts, it contended the inspection reduction “improves efficiency while still helping to ensure reasonable assurance of adequate protection to the public.”

Commission member Jeff Baran criticized the proposed changes Tuesday, saying reducing oversight of the nuclear power industry “would take us in the wrong direction.”

“NRC shouldn’t perform fewer inspections or weaken its safety oversight to save money,” Baran said.

The release comes a day after Democratic lawmakers faulted the NRC’s deliberations, saying they had failed to adequately inform the public of the changes under consideration.

“Cutting corners on such critical safety measures may eventually lead to a disaster that could be detrimental to the future of the domestic nuclear industry,” Rep. Frank Pallone, D-N.J., chair of the House Energy and Commerce Committee, and other House Democrats said in a letter Monday to NRC Chairwoman Kristine Svinicki.

Asked for comment Tuesday, NRC spokespeople pointed to the staff arguments for the changes in the report. Trimming overall inspections “will improve effectiveness because

There is a common definition of a word and the industry's and nrc's definitions of words. "Safety" and "significance" are such slimy words. Those words means anything the agency wants?  

inspectors again will be focused on issues of greater safety significance,” staffers told commission members in the recommendations.

Edwin Lyman, a nuclear-power expert at the nonprofit Union of Concerned Scientists, faulted the reasoning of commission staff that the good performance of much of the nuclear power industry warranted cutting back on agency inspections for problems and potential problems.

“That completely ignores the cause-and-effect relationship between inspections and good performances,” Lyman said.

Brattleboro-Hinsdale Bridge: My Vision of a New Island Park.

Update

I will probably get a email from Brattleboro Reformer I am infringing again on their newspaper copy rights. But are they a newspaper?

***Right, I been thinking what to do with this Park since I parked my ass next to the bridges during my protesting days. I was the first one who conceived of this project.

'Pie in the sky' vision for island, bridges

Posted Tuesday, July 16, 2019 7:51 pm

By Bob Audette, Brattleboro Reformer

BRATTLEBORO — While the new bridge connecting Hinsdale, N.H., to downtown Brattleboro won't be complete until the fall of 2023, discussions are underway as to what to do with the two existing bridges and the island between them.

On Tuesday, the Existing Bridges Committee, which is a subcommittee of the Hinsdale-to-Brattleboro Project Advisory Committee, met at the Brattleboro Municipal Building to begin the process of deciding how the bridges and the island might be used once traffic is diverted to the new bridge.

The replacement bridge will cross the Connecticut River at the traffic signal for George's Field on the New Hampshire side and will land on the Vermont side, on Route 142, just south of the parking lot for 28 Vernon St., which was formerly known as the Marlboro College Graduate Center.

When asked why the old bridges shouldn't just be removed, J.B. Mack, Principal Planner for the Southwest Region Planning Commission, said as part of the environmental review for the replacement bridge, the Anna Hunt Marsh and the Charles Dana bridges were characterized as part of the historical heritage of the region.

"The state historical agencies also thought it made sense to keep the bridges," he said. "Some people disagree with that assessment, but that's a major milestone in getting federal funding for the project."

While there is debate on whether the bridges should stay or go, said Brattleboro Assistant Town Manager Patrick Moreland, the scope of the the Existing Bridges Committee "assumes the bridges will stay and people will have access to the island."

Moreland noted that the overall goals for the main project we were established in 2013. 

They included: Maintaining a transportation corridor between the two towns; fixing the safety, structural and functional deficiencies of the existing corridor; maintaining area social and economic relationships; preserving the integrity of area resources; and conserving fiscal resources.

Notably, "The Brattleboro/Hinsdale transportation corridor has numerous natural and cultural resources that contribute to the social, economic, environmental, and aesthetic qualities of the area," states "Purpose and Needs" statement for the project.

"To the extent possible," said Moreland, "we should consider minimizing the impact on the local taxpayers. There's going to be considerable money spent here, but it's important to consider ways in which to have it not hit the tax base in Brattleboro or Hinsdale."

The cost for the entire project is $50 million, with $26 million coming from New Hampshire, $4 million from Vermont and $20 million from either a federal grant or low-interest loans.

Robert Landry, the administrator for the Bridge Design Division of the N.H. Department of Transportation, said rehabilitating the existing bridges for pedestrian use is estimated to cost $8 million, which is included in the $50 million price tag.

How much it will cost to maintain the bridges and how long they will last after the new bridge is open is being studied and those numbers should be available soon, said Landry.

Who will be responsible for the upkeep of the bridges and any improvements to the island, which might include signage, lighting and some sort of low-impact event space, has yet to be determined. Currently, the state of New Hampshire owns the bridges and the island, and Hinsdale has shown little interest in assuming ownership, and the accompanying fiscal responsibility. The town of Brattleboro has also indicated it doesn't want to own the bridges or the island.

"There is an ongoing discussion, not at this committee's level, working toward resolving who ultimately will be the owners," said Moreland, adding how much it will cost to maintain the bridges and the island will help to inform that decision.

Nonetheless, said Moreland, it's important to define a vision for the space when the bridges are closed in 2023.

"What do we aspire to?" asked Mack. "The possibilities are wide open."

The committee spent some time discussing the recreational opportunities afforded by the bridges and the island, including access to hiking and biking trails on both sides of the Connecticut River and access to the river itself, such as for swimmers and kayakers and canoeists.

Moreland noted that whatever happens with the existing bridges, the community needs to know that the bridges and the island will be "clean and safe."

Currently, the island is being used as a camping spot for the homeless and for drug use. The Hinsdale Police Department has conducted sweeps on the island to clear the campers out, but they usually return.

"Actually making it a more active destination as opposed to people moving through the site like now, would be better for keeping it clean and safe," noted Kathy Urffer, a river steward with the Connecticut River Conservancy. "A little conscious development might go a long way, such as interpretive signs, maybe a deck or a vista where people could sit and paint or look at the river."

The committee agreed the area could be used for cultural events such as small concerts or museum exhibits, but expressed concerns about crowds that might be attracted to the island and the lack of bathrooms.

Porta poddies? NH is spending the most money on the project and the Brattleboro Refomer is soley allowing the Brattleboro losers to publically discuss these issues.   

"The number of people who can be on that island is limited," said Moreland.

"Pie in the sky," said Marion Major, a planner with the Windham Regional Commission, "we'd love to have events staged here with an extremely low-maintenance event space."

"It's important to keep it pretty broad," agreed Mack. "The next step is to get the public involved."

The committee discussed ways to get people involved and how best to solicit ideas from the public. They agreed that public meetings were a good starting point, but also suggested setting up information booths at local farmers markets and events.

The planning commissions also hope to set up an online commenting tool for people who can't make the meetings or don't necessarily like to speak in public.

Mack said if people want to submit ideas for the existing bridges or the island right now, or have any ideas on how to reach the committee can reach out to the public, they should email him at jbmack@swrpc.org.