Junk Plant Hope Creek: Stupidly Damaged Fuel By Not Following Procedure

Update Nov 19

Basically, this is the kind of mistake you should take their license away from them for 6 months to a years.

***This can't come from a seasoned and well trained crew.  This is like getting into you car without keys. This is basically a very simple procedure. Basically up around 70% power, they drastically slowed down the increase in power to 100% and are required more reactor testing. This is so stupid, it has to go across all the crews. It was maliciously stupid....

November 14, 2018

EA-18-044 

(Here is the link) 08038    SUBJECT: HOPECREEK GENERATING STATION UNIT 1 – INTEGRATED INSPECTION REPORT AND EXERCISE OFENFORCEMENT DISCRETION 05000354/2018003

Inadequate Procedures for Fuel Conditioning Results in Multiple Fuel Leaks Cornerstone Significance Cross-Cutting Aspect Report Section Reactor Safety – Barrier Integrity

 Green  NCV 05000354/2018003-03 Closed H.6 – Human Performance – Design Margins 71153 (a.1) The inspectors documented a self-revealing Green NCV of TS 6.8.1, Procedures and Programs, when PSEG did not maintain adequate procedures for fuel conditioning.  Specifically, PSEG’s procedure for selecting the appropriate fuel PCI rules, NF-AB-440, BWR Fuel Conditioning, did not provide adequate guidance for protection of the fuel during restart from the April 2018 refueling outage (RF21).  As a result, PSEG’s selection non-conservative PCI rules resulted in three PCI fuel leaks.  Description:  HCGS is currently in operating Cycle 22 with a modified control cell core design strategy.  During RF21, a large number of GE14 fuel assemblies were replaced with new GNF2 fuel assemblies.  Each of the four Group 10A banked position withdrawal sequence control cells are at the center of two twice-burned GE14 (new fuel in Cycle 20) fuel assemblies and two once-burned GNF2 (new fuel in operating Cycle 21) fuel assemblies that form a control cell.  During Cycle 21, prior to RF21, the GE14 fuel assemblies were in the second row from the periphery (lower power), or outer edge of the reactor core.  In RF21, some of these fuel assemblies were moved from the outer edge, inward toward the core center (higher power) into the cells surrounding the four Group 10A control cells.  This type of movement is known to create a configuration that may reduce margin to pellet-clad interation related failures….

Just remember, the NRC would not be so fixed on the recent spate of problem, if I didn't the  problems much earlier.

Review of Equipment Issues Associated with the ‘H’ SRV and SRV Discharge Line:

The inspectors performed an in-depth review of PSEG's evaluation and corrective actions associated with multiple equipment issues experienced on the ‘H’ main steam safety relief valve (SRV) and SRV discharge line.  Specifically:

1. ‘H’ SRV Main Seat Leakage August 2014 (NOTF 20659947; ACE 70168360) documented loud cyclic banging noises coming from the TORUS area.  PSEG determined that there was significant leakage past the ‘H’ SRV main seat due to the existence of cold spring in the tailpipe during installation of the valve (NOTF 20661387 and NCV 05000354/2014005-01);

2. ‘H’ SRV High Tailpipe Temperature April 2018 and May 2018 (NOTF 20789878, 20794091 and 20794237) documented that during down power for and the start up from RF21, the ‘H’ SRV tailpipe temperature spiked up to 220 degrees Fahrenheit which is indicative of potential SRV main and/or pilot valve leakage;

3. ‘H’ SRV Vacuum Breaker Failure April 2018 (NOTF 20792630 and ERE 70199676) documents that one of the ‘H’ SRV discharge line vacuum breakers (F037H) failed open due to a missing locknut and damage caused by high vibrations and poor maintenance practices from item #1 above; and,

4. ‘H’ SRV Pilot As-Found Lift Test Failures May 2018 (NOTF 20794371, 70200658, and LERs 05000354/2018-002-00 and -01) documented the ‘H’ SRV pilot as-found setpoint testing.  Eight of HCGS’s fourteen SRV pilots lifted high (above the 3 percent TS limit).  The ‘H’ SRV pilot was the only valve that lifted high on the first and second as-found lift testing (8.3 and 3.3 percent).  

[Note that the 2-stage SRVs, manufactured by Target Rock, of which HCGS has 13 2-stage and 1 3-stage SRVs, have been subject to setpoint drift, typically in the increased setpoint direction at a number of boiling water reactor nuclear power plants, and that the specific setpoint drift issue will be addressed by the unresolved item (URI) opened in NRC Inspection Report, URI 05000354/2018001-02, Concern Regarding As-Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Limit.]

The inspectors reviewed associated documents and interviewed personnel to assess the adequacy of PSEG’s actions.  The inspectors also reviewed SRV main and pilot testing results, tailpipe temperature, main steam vibration records, and acoustic monitoring data.  The inspectors found the following issues during their review of the events listed above:

The inspectors found that PSEG had an extended timeline (6 months) and a lack of prioritization and ownership of the disassembly of the ‘H’ SRV pilot due to it lifting high twice (NOTF 20799218*).  Based on the inspector’s questions regarding timeliness, PSEG initiated a NOTF and actions to disassemble and inspect the pilot four months ahead of its original schedule.  As a result of the disassembly, PSEG’s determined that the pilot disc and valve body were severely steam cut and worn, with unknown impurities on the valve pilot disc.  PSEG initiated work group evaluation (WGE) 70200658 to evaluate these unexpected conditions;

WGE 70200658 was completed on September 21, 2018, for the ‘H’ SRV failed setpoint lift test high twice in which PSEG determined that the first high test lift was due to corrosion bonding, and the second high test lift was due to pilot valve wear between the disc and liner caused by steam cutting from a pilot leak during the last operating cycle.  PSEG’s WGE found that some of the unknown impurities were cobalt and nickel oxide due to the corrosion bonding experienced by the valve.  The WGE did not determine the source of the lead (Pb) in the impurities but pointed to the valve material test report that cites 0.5 percent of the total valve disc material being from ‘OTHER’ material.  The inspectors reviewed PSEG’s conclusion and discussed with PSEG on September 27, 2018, that the site is still awaiting feedback from the vendor and BWROG about the potential source of the lead in the impurities.


The inspectors determined that there was insufficient information provided by PSEG in licensee event report (LER 2018-002) for the as-found testing results of the SRV pilots, specifically, no information on the ‘H’ SRV pilot lifting high twice was reported As a result, PSEG initiated NOTF 20799025* and took corrective actions (70201546) to change their process for LER reviews to include a technical validation team review prior to submittal to the NRC;

The inspectors found that PSEG’s procedure for SRV removal and installation, HC.MDCM.AB-0006, was not revised in accordance with their causal evaluation (70168360) to include a step to unpin the spring can after installation of the SRV.  PSEG initiated a NOTF with actions to revise the procedure and review all completed SRV work packages to ensure all pins were removed (NOTF 20801471*, 20803451*, and 70202115).  As a result, PSEG’s review found that three SRVs replaced in RF20 did not have any documentation that their spring cans had been unpinned.  PSEG has created actions to conduct follow-up inspection of these SRVs (‘J’, ‘K’, and ‘R’) during the next refueling outage;

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The inspectors found that PSEG’s NOTF 20661387 and 70169063-0010 never validated a questionable spring can setting for the ‘H’ SRV due to a lack of understanding the issue.  Because of this, inspectors also questioned the validity of PSEG’s causal evaluation (70168360) conclusions based on the as-found cold spring being expected because of the piping configuration.  The inspectors determined that during the development of the evaluation, PSEG did not consult the appropriate resources knowledgeable in pipe stress analysis.  As a result, PSEG took action to validate that the spring can setting was correct and initiated NOTF 20803213* with a recommendation from engineering to review the causal evaluation’s conclusions based on the inspector’s questions and an independent engineering assessment.  As of September 12, 2018, this recommendation was not supported by PSEG because the condition on the ‘H’ SRV is no longer present and there is no perceived value in performing the action.  The inspectors noted that as of the end of this inspection period, PSEG initiated NOTF 20806034 on October 1, 2018, for degrading conditions associated with the ‘H’ SRV main seat leakage increasing from ~155 pound mass per hour (lbm/hr) to approximately 323 lbm/hr since H1R21 (June 2018), which is similar to the conditions that occurred on the ‘H’ SRV in August 2014, and were the subject of PSEG’s causal evaluation (70168360).

The inspectors found that PSEG’s WGE 70173184 had not determined a basis for what amount of displacement is considered unacceptable.  In addition, PSEG had not performed trending of SRV piping misalignments as discussed in the WGE for RF19 (2015) and RF20 (2016).  PSEG initiated NOTFs 20803211* and 20803212* to address the inspector’s concerns and plans to perform extent of condition reviews of all SRV main replacements over the last few outages.

The inspectors evaluated all of the issues above in accordance with the guidance in IMC 0612, Appendix B, “Issue Screening,” and Appendix E, “Examples of Minor Issues,” and determined the issues were of minor significance because the inspectors did not identify any condition adverse to quality that were not appropriately corrected or scheduled for correction in a reasonable period of time as a result of PSEG’s administrative delays, lack of prioritization, and insufficient information.  Consequently, these issues are not subject to enforcement action in accordance with the NRC’s enforcement policy.