IR 05000275/2018008
| ML18159A483 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/08/2018 |
| From: | Gerond George NRC Region 4 |
| To: | Welsch J Pacific Gas & Electric Co |
| George G | |
| References | |
| IR 2018008 | |
| Download: ML18159A483 (27) | |
Text
une 8, 2018
SUBJECT:
DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - NRC BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000275/2018008 and 05000323/2018008
Dear Mr. Welsch:
On May 3, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at the Diablo Canyon Power Plant, Units 1 and 2. The NRC inspection team discussed the results of this inspection with you and members of your staff. The results of this inspection are documented in the enclosed report.
The NRC inspection team reviewed the stations corrective action program and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for corrective action programs. Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety. However, the team identified some challenges in the area of problem identification in that issues were not always adequately identified by station personnel in a way that they could be promptly and fully addressed.
The team also evaluated the stations processes for use of industry and NRC operating experience information, and the effectiveness of the stations audits and self-assessments.
Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety. However, the team noted instances where lessons learned from industry and NRC operating experience had not been incorporated into station programs, processes, and procedures to prevent similar operational events.
Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment, and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, the team found no evidence of challenges to your organizations safety-conscious work environment.
Diablo Canyon Power Plant, Units 1 and 2 employees appeared willing to raise nuclear safety concerns through at least one of the several means available. NRC inspectors documented two findings of very low safety significance (Green) in this report, both of which involved violations of NRC requirements. Additionally, inspectors documented a licensee-identified violation that was determined to be of very low safety significance. The NRC is treating all of these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest these violations or their significance, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at the Diablo Canyon Power Plant, Units 1 and 2.
Likewise, if you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV and the NRC resident inspector at the Diablo Canyon Power Plant, Units 1 and 2.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Gerond A. George, Team Leader Inspection Programs and Assessment Team Division of Reactor Safety Docket Nos. 50-275 and 50-323 License Nos. DPR-80 and DPR-82
Enclosure:
Inspection Report 05000275/2018008 and 05000323/2018008 w/ Attachments:
1. Information Request 2. Supplemental Information Request
Inspection Report
Docket Number(s): 05000275, 05000323 License Number(s): DPR-80, DPR-82 Report Number: 2018008 Enterprise Identifier: I-2018-008-0002 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Power Plant, Units 1 and 2 Location: Avila Beach, California Inspection Dates: April 16, 2018, to May 3, 2018 Inspectors: E. Ruesch, J.D., Sr. Reactor Inspector (Team Lead)
J. Braisted, Ph.D., Reactor Inspector N. Okonkwo, Reactor Inspector J. Reynoso, Resident Inspector Approved By: Gerond A. George, Team Leader Inspection Programs and Assessments Team Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at Diablo Canyon Power Plant, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. Findings and violations being considered in the NRCs assessment are summarized in the table below.
A licensee-identified non-cited violation is discussed in report section 71152.
List of Findings and Violations Failure to Identify Diesel Generator Air Inlet Boot Seal Critical Characteristics Cornerstone Significance Cross-cutting Aspect Report Section Mitigating Green NCV 05000275/2018008-02; None 71152Problem Systems05000323/2018008-02 Identification and Closed Resolution A self-revealed, Green, non-cited violation (NCV) of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criteria VII and XV, occurred when the licensee failed to ensure materials intended for installation in safety-related applications conformed to procurement requirements or, if they did not, were adequately controlled and evaluated.
Failure to Promptly Identify and Correct Emergency Diesel Generator 1-1 Cardox System Inoperability Cornerstone Significance Cross-cutting Aspect Report Section Mitigating Green NCV 05000275/2018008-03 H.14 71152Problem Systems Closed Identification and Resolution An NRC-identified, Green, non-cited violation (NCV) of the licensees fire protection license condition occurred when licensee personnel failed to identify a trouble light lit on the Emergency Diesel Generator (DG) 1-1 cardox fire protection system panel. The light, which had been lit for 2 weeks before being identified by the NRC, indicated a condition that would have prevented the automatic fire suppression system from effectively suppressing a fire in the DG 1-1 room.
Additional Tracking Items Type Issue number Title Report Status Section URI 05000275/2018008-01; Emergency Diesel Generator 71152 Open 05000323/2018008-01 Mission Time for Operability Inspection Evaluations Results
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs)in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.
Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
OTHER ACTIVITIES - BASELINE
71152Problem Identification and Resolution Biennial Team Inspection
The inspectors performed a biennial assessment of the licensees corrective action program, use of operating experience, self-assessments and audits, and safety-conscious work environment. The assessment is documented below.
- (1) Corrective Action Program Effectiveness: Problem Identification, Problem Prioritization and Evaluation, and Corrective Actions - The inspection team reviewed the stations corrective action program and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for corrective action programs.
The sample included review of over 200 notifications (SAPNs) and associated records, and an in-depth 5-year review of condition reports associated with the auxiliary feedwater (AFW) piping and seismic supports.
- (2) Operating Experience, Self-Assessments, and Audits - The team evaluated the stations processes for use of industry and NRC operating experience. The team also evaluated the effectiveness of the stations audits and self-assessments program. The sample included several industry operating experience communications and associated site evaluations.
- (3) Safety Conscious Work Environment - The team evaluated the stations safety-conscious work environment. The team interviewed 43 station personnel in 5 group interviews. These included personnel from operations, engineering, maintenance, security, fire protection, and radiation protection and chemistry. The team also interviewed employee concerns program personnel, reviewed employee concerns files, and reviewed the results of the most recent safety culture survey.
INSPECTION RESULTS
- OBSERVATIONS/ASSESSMENT Corrective Action Program Assessment 71152Problem Identification and Resolution Effectiveness of Problem Identification: Overall, the team found that the licensees identification and documentation of problems was adequate to support nuclear safety, though some challenges were noted. In particular, the team identified several equipment conditions in the plant that had not been identified by station personnel. These are described in observations, findings, and violations below.
Effectiveness of Prioritization and Evaluation of Issues: Overall, the team found that the licensees prioritization and evaluation of issues was adequate to support nuclear safety.
Effectiveness of Corrective Actions: Overall, the team found that the licensees corrective actions generally supported nuclear safety. However, the team noted some examples where corrective actions had not been timely or were not completed as originally designed. These examples are described in observations, findings, and violations below.
Observations on the Corrective Action Program (CAP) 71152Problem Identification and Resolution The teams 5-year review of CAP items related to auxiliary feedwater (AFW) piping and seismic supports found that material conditions were generally appropriately identified, adequately evaluated, and appropriately corrected. However, in reviewing the first quarter 2017 system/component walkdown checklist per Diablo Canyon Power Plant (DCPP), Units 1 and 2 system engineering program, dated April 17, 2017, the team noted that walkdown or assessment of the material conditions of piping inside the auxiliary building was not included.
Licensee personnel stated that because AFW lines 1 and 2 are complex piping that is exposed to the outside environment, and experience has shown significant thermal movements is experienced, they are walked down. In contrast, AFW lines 3 and 4 are located indoors in the auxiliary building and are shorter non-complex pipe runs; therefore, walkdowns are unnecessary. However, on February 10, 2015, while working on a pipe hanger, maintenance personnel noticed excessive wear (1/16 inch deep) into an AFW pipe wall caused by a misaligned shim plate resulting in excessive vibration. The shim plate was replaced and piping wear was determined not to impact AFW operability. Although the piping wear met the design basis allowable stress limits, this wear had been ongoing for a long time and likely could have been identified sooner if system engineering walkdowns required inspections of piping lines inside the auxiliary building. The licensee documented this observation in SAPN 50978723.
The team determined DCPP continued to identify adverse trends in document control timeliness and complete records. DCPP initially identified an issue with timely submittal of quality records into the record management system in December 2015. This issue was identified again in 2017, indicating corrective actions have not yet been effective. The associated cause evaluations point to a lack of knowledge and training. The breadth of document control issues includes security, engineering, maintenance, and operations. There have been several documented instances in which records were missing or not retained. The team reviewed the corrective actions to address document control which include contributing factors that involve human error, lack of training, workload, poor procedures, and lack of supervisory oversight. A review of the immediate and planned corrective actions by the team concluded that appropriate corrective actions have been initiated and the appropriated priority is being placed that will improve this long-standing issue. The licensee documented this observation in SAPN 50978721.
The team identified an event associated with changing light bulbs on safety-related electrical panels as another missed opportunity to review past operating experience. In 2009 an operator, while changing a light bulb on a safety-related 4kV panel, inadvertently unscrewed the light socket, causing an electrical short in the emergency diesel circuit, and resulting in DG inoperability. Corrective actions developed to prevent this event from occurring again included placing warning labels on high risk electrical panels. However, as of April 2018 these tags had not been installed as required by that action. Potentially as a result, on April 10, 2018, while attempting to change a light bulb on the startup transfer breaker on vital bus G the light bulb socket came loose in the very same manner as the event in 2009. The operator stopped soon enough to prevent another electrical short event, but extensive troubleshooting and maintenance resources were required to return the light socket to normal condition. The NRC identified this as a near miss of a repeat equipment event. The licensee documented this observation in SAPN 50978496.
Use of Operating Experience Assessment 71152Problem Identification and Resolution In general, the team found operating experience (OpE) information was being appropriately used at DCPP, Units 1 and 2, to ensure potential issues were promptly identified and corrected. However, the team noted some indications that screening of OpE information could be improved.
Observation on the Use of Operating Experience 71152Problem Identification and Resolution The team determined that gaps exist in the OpE program associated with the cause evaluation process since guidance requires a review and identification of OpE preventable events, but this process is not being appropriately implemented. Specifically, Procedure OM7.ID3, Root Cause Evaluation Process, and Procedure OM7.ID4, Cause Evaluation, provide guidance on looking at missed opportunities documentation into whether there were OpE preventable events; the documentation of these processes was inconsistent.
In at least one casewhich resulted in a performance deficiency that led to a charging pump failure, as described in NRC Inspection Report 2018001this represented a missed opportunity to identify a vulnerability and take actions to prevent an event. The team concluded that additional guidance or training may be required to ensure the CAP process properly assesses missed opportunities to prevent an event. One example of such a missed opportunity is described in the corrective action program observations above. The licensee documented this observation in SAPN 50978724.
Self-Assessments and Audits Assessment 71152Problem Identification and Resolution Based on the samples reviewed, the team determined that station performance in these areas adequately supported nuclear safety.
Safety Conscious Work Environment Assessment 71152Problem Identification and Resolution The team found no evidence of challenges to the safety-conscious work environment of station work groups. Individuals appeared willing to raise nuclear safety concerns through at least one of the several means available. However, the team found evidence of possible challenges to the safety-conscious work environment within some onsite Pacific Gas and Electric (PG&E) work groups whose reporting lines are outside the nuclear organization.
INSPECTION RESULTS
- ISSUES/FINDINGS Minor Violation 71152Problem Identification and Resolution Performance Deficiency: Failure to use the site corrective action program to track, trend, correct, and prevent recurrence of failures and deficiencies in the physical protection program, as required by Title 10 of the Code of Federal Regulations § 73.55, Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage.
On April 17, 2018, during a plant tour, inspectors identified a deficiency associated with the physical protection program and brought it to the attention of control room operators. On May 1, 2018, inspectors asked licensee personnel for a copy of the SAPN documenting the deficiency. None had been initiated. Further, the deficiency had not been logged in the security logs as required. The failure to log the issue was itself a loggable event. The licensee documented the deficiency and the failure to initially document it in SAPN 50978291.
Screening: The performance deficiency was minor because it would not have led to a more significant security concern and did not adversely affect the security cornerstone objective.
Enforcement:
This failure to comply with 10 CFR 73.55 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Minor Violation 71152Problem Identification and Resolution Performance Deficiency: Failure to install safety-related pressure transmitters (PTs) in accordance with engineering design documents, without documented authorization and prior approval for deviation from that design.
Unit 2 Steam Generator pressure transmitters PT-544A and PT-534A were not installed per design. The design called for mounting the PTs on independent unistruts but, contrary to this, the transmitters were installed on a common unistrut. Though the new mounting configurations were documented and analyzed in SAPNs 50881613 and 50881415, Work Order 68039185 which installed the PTs did not record the deviation from originally designed mounting configuration. The licensee attributed the failure to install per original design to human error and initiated SAPN 50976632 to evaluate it.
Screening: The performance deficiency is minor in that the current configuration was evaluated not to affect the seismic or structural qualification.
Enforcement:
This failure to comply with 10 CFR Part 50, Domestic licensing of production and utilization facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion III, Design Control, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Licensee-Identified Non-Cited Violation 71152Problem Identification and Resolution This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program. This is being treated as a non-cited violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR Part 50, Appendix B, Criterion III, requires that measures shall include provisions to assure that appropriate quality standards are specified and included in design documents, and that deviations from such standards are controlled.
Contrary to the above, from approximately February 2004 until August 2017, the licensee did not assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, the licensee had classified the seat o-ring used in Crosby and Lonergan pressure relief valves (e.g., RV-354 and RV-355) servicing safety-related back-up air/nitrogen applications as non-safety related when they should have been classified as safety-related. Consequently, the o-rings were procured as commercial grade (non-safety related), not dedicated as safety-related and installed in safety-related equipment.
Significance/Severity Level: This violation was more than minor because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, the use of non-qualified seat o-rings had the potential to cause excessive leakage past the seat, adversely affecting the fixed air/nitrogen volume required to operate safety-related equipment during a loss of normal air/nitrogen. Using IMC 0609, Appendix A, dated June 19, 2012, the team determined that this violation was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system or component, and operability was maintained.
Corrective Action Reference(s): SAPNs 50935776 and 50970247 Unresolved Item Emergency Diesel Generator Mission Time for 71152Problem (Open) Operability Evaluations URI 05000275/2018008-01; Identification and 05000323/2018008-01 Resolution
Description:
The team identified an unresolved item (URI) related to diesel generator (DG) mission time for operability evaluations. On December 3, 2016, an operator discovered during rounds that the air inlet boot seal on DG 1-2 had degraded, and subsequently, an inspection of the other diesel generators (DGs) revealed that the DG 2-2 boot seal was also degraded. The licensee performed an operability evaluation and concluded that the DGs were operable based on a mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee then performed a past operability evaluation, concluding that the DGs had remained able to perform their safety function for this stated 24-hour mission time despite the deficiency; therefore no licensee event report was required by 10 CFR 50.73.
The team requested information related to the basis of the 24-hour mission time. The licensee provided a non-controlled reference document, Engineered Safety Feature (ESF)
Equipment Mission Time, to the licensees operability determination Procedure OM7.ID12.
The document listed the mission time for the DGs as 7 days (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). The 6 and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> values depend on the particular accident sequence and electrical power recovery time, and were from a letter sent to the NRC related to the licensees Individual Plant Examination of External Events (IPEEE), which is a plant-specific probabilistic risk assessment (PRA). The 7-day value is related to the required diesel fuel oil storage volume as discussed in Technical Specification Bases 3.8.3. The document also states that the licensee has no defined post-accident operation / mission times because such times are not mandated by regulation or recommended by NRC guidance.
The team noted, however, that IPEEEs do not typically evaluate accidents past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and furthermore, IMC 0326, Operability Determinations and Functionality, states that the use of PRA or probabilities of occurrence of accidents or external events is not consistent with the assumption that the event occurs, and is not acceptable for making operability decisions.
Additionally, Procedure OM7.ID12 defines mission time as the duration of structure, system, or component (SSC) operation that is credited in the current licensing bases for the SSC to perform its specified safety function; however, as documented above by the licensee, there is no design or licensing basis mission time for the DGs. The licensees definition of mission time is essentially the same as described in IMC 0326.
The inspectors performed a brief review of documents related to mission times. Technical Specification Limiting Condition for Operation 3.8.3, Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist, requires verification of diesel fuel oil level to satisfy a 7-day fuel oil storage requirement. Additionally, NUREG-1407 discusses an Electric Power Research Institute approach that defines and evaluates the capacity of those components required to bring the plant to a stable condition (either hot or cold shutdown), and maintain that condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Also, the ESF equipment mission time document referenced several 30-day mission times for SSCs that would require emergency power from either offsite power, if available, or the DGs. The team also performed a search of previous NRC findings at the DCPP, Unit 1 and 2, and found one reference to a 7-day mission time for the DGs in NRC Pilot Engineering Inspection Report 2006005.
The inspectors also reviewed NEI 97-04, Design Bases Program Guidelines, Revised Appendix B, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases. The Appendix describes how the 10 CFR 50.2 design bases of a facility are a subset of the current licensing basis and are required pursuant to 10 CFR 50.34(a)(3)(ii) and
- (b) and 10 CFR 50.71(e), to be included in the updated Final Safety Analysis Report (FSAR). Title 10 CFR 50.2 design bases consist of design bases functions and design bases values. Design bases values are the values or ranges of values of controlling parameters established as reference bounds for design to meet design bases functional requirements. In other words, the 10 CFR 50.2 design bases include the bounding conditions under which SSCs must perform their design bases functions and may be derived from normal operation, or any accident or events for which SSCs are required to function.
Because 10 CFR 50.71(e), IMC 0326, and Procedure OM7ID.12 indicated that DG mission time should be part of the design and licensing bases, and documented in the FSAR, but a DG mission time design and licensing basis does not appear to exist at DCPP, Units 1 and 2, the inspectors could not determine that an appropriate mission time was used for a past operability determination. Therefore, the team could not conclude that the licensee had not missed a 10 CFR 50.73 event report because of a potentially incorrect assumption about DG mission time.
This is applicable to both units.
Planned Closure Action(s): In order to resolve this issue, the NRC needs to determine whether or not the basis for the 24-hour DG mission time is appropriate by determining which standard or standards apply to mission time at DCPP, Units 1 and 2.
Licensee Action(s): Because the licensees position is that the DG mission time is not a part of their current licensing or design basis, they maintain that the 24-hour mission time used in the past operability determination was adequate to provide reasonable assurance of operability and, therefore, no event report was required. However, prior to this inspection and because of other uncertainties in determining mission times, the licensee generated Notification 50832335 to reassess the mission times associated with the ESF equipment.
The intent is to develop the bases for ESF equipment mission time in a controlled document.
However, this effort is not yet complete and, as such, the mission time for the DGs has not been evaluated under this notification.
Corrective Action Reference(s): Notifications 50832335, 50882125, 50882140, and 50882498.
Failure to Identify Diesel Generator Air Inlet Boot Seal Critical Characteristics Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green NCV 05000275/2018008-02; None 71152Problem Systems05000323/2018008-02 Identification and Closed Resolution A self-revealed, Green, non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criteria VII and XV, occurred when the licensee failed to ensure materials intended for installation in safety-related applications conformed to procurement requirements or, if they did not, were adequately controlled and evaluated.
Description:
On December 3, 2016, an operator discovered during rounds that the air inlet boot seal between the air filter intake piping and the inlet to the turbocharger on DG 1-2 had degraded to the point that the boot seal was cracked around the entire perimeter with sections missing.
One piece of the boot seal was visible in the inlet of the turbocharger. Subsequently, an inspection of the other DGs revealed that the DG 2-2 boot seal was also torn.
The licensee replaced the degraded boot seals on the date of discovery and performed a material chemical analysis of the boot seals on December 8, 2016. The chemical analysis found the boot seals were not of neoprene as required. The same analysis also indicated a high likelihood that the boot seals were of Styrene-Butadiene Rubber.
The licensee performed an apparent cause evaluation of the issue. The evaluation concluded that the Replacement Part Evaluation (RPE) of the boot seal, RPE 8*1743, was inadequate since they had failed to detect the improper material during the dedication process of the boot seals purchased as commercial grade. Specifically, the RPE did not list material type as a critical characteristic and, therefore, did not have a requirement to perform material validation testing for neoprene (during dedication activities when the boot seal was procured commercial grade). Boot seals were most recently procured under Work Order 3500988221 and the dedication activities were performed between February 2, 2014, and March 19, 2014.
As stated in 10 CFR Part 21, when applied to nuclear power plants licensed pursuant to 10 CFR Part 50, critical characteristics are those important design, material, and performance characteristics of a commercial grade item that, once verified, will provide reasonable assurance that the item will perform its intended safety function. Furthermore, dedication is an acceptance process undertaken to provide reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety function and, in this respect, is deemed equivalent to an item designed and manufactured under a 10 CFR Part 50, Appendix B, quality assurance program.
Corrective Action(s): In response to this issue, the licensee replaced the affected boot seals in operation, removed a nonconforming boot seal from storage, updated RPE 8*1743 to include material type as a critical characteristic, added a requirement to verify material type during dedication activities, and increased the boot seal inspection frequency.
Corrective Action Reference(s): Notifications 50882125, 50882140, and 50882498.
Performance Assessment:
Performance Deficiency: The failure to identify critical characteristics of the DG air inlet boot seal and verify its acceptability by inspections, tests, or analyses, as required by 10 CFR Part 21, was a performance deficiency.
Screening: The performance deficiency was more than minor because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, an unacceptable boot seal could degrade to the point that pieces of the boot seal, or other foreign materials, would get ingested into the engine of a DG and adversely affect the availability and reliability of that engine.
Significance: Using IMC 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system, or component, and operability was maintained.
Cross-Cutting Aspect: No cross-cutting aspect was assigned to this finding because the team determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, requires that measures be established to assure that purchased material, equipment, and services conform to the procurement documents. Criterion XV, Nonconforming Materials, Parts, or Components, requires that nonconforming items shall be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.
Contrary to these requirements, prior to approximately March 2014 the licensee failed to ensure purchased material conformed to procurement documents, and failed to review and accept, reject, repair, or rework nonconforming items in accordance with documented procedures. Specifically, the licensee failed to identify that the received DG air inlet boot seals were not made of neoprene as specified in the procurement document and, subsequently, did not review and accept, reject, repair, or rework the nonconforming items.
The licensee failed to identify the boot seal was not of neoprene because it did not list material type as a critical characteristic of the item in its commercial grade dedication document and, therefore, the material type was not verified. Consequently, the licensee did not successfully complete the commercial grade dedication process for the boot seals as described in 10 CFR Part 21.
Disposition: This violation is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Promptly Identify and Correct Emergency Diesel Generator 1-1 Cardox System Inoperability Cornerstone Significance Cross-cutting Report Section Aspect Mitigating Green H.14 71152Problem Systems NCV 05000275/2018008-03 Identification and Closed Resolution An NRC-identified, Green, Non-cited Violation of the licensees fire protection license condition occurred when licensee personnel failed to identify a trouble light lit on the DG 1-1 cardox fire protection system panel. The light, which had been lit for 2 weeks before being identified by the NRC, indicated a condition that would have prevented the automatic fire suppression system from effectively suppressing a fire in the DG 1-1 room.
Description:
On April 18, 2018, during performance of cardox testing for the DG 1-1 room, licensee personnel identified an unexpected light configuration on the test panel. As documented in SAPN 50976405, the Test Engineer contacted the WCSFM [work control shift foreman] and both agreed that acceptance criteria was [sic] not affected. The WCSFM then authorized the test engineer to complete the test and write a notification. Later in the test, a step could not be completed because of another light remaining lit unexpectedly. The test engineer described the condition in the SAPN: When the panel door is closed this light shines through the window labeled Wiring/Links. This is likely related to the issue described above. This condition did not impact any of the acceptance criteria. The WCSFM advised the Test Engineer to continue the test. Ultimately, the test was completed with all acceptance criteria met. Following completion of the test, an on-shift Senior Reactor Operator declared the cardox system functional at 1743 on April 18, 2018.
On May 1, 2018, during a tour of the Unit 1 DGs, inspectors noted the Wiring/Links light illuminated on the DG 1-1 room cardox fire suppression system panel. A notification tag was attached to the panel that did not describe the observed condition. The inspectors informed the shift manager, who requested an engineering evaluation.
On May 2, 2018, engineering completed its review of the condition and determined that the illuminated light indicated a configuration in which the fire suppression system would be unable to perform its credited function: Per Fire Protection, automatic actuation of the west rollup doors is credited for Cardox function. With this light ON, actuation of the west door ETLs [electro-thermal links] is disabled. The previous functionality determination had been incorrectthe system had been nonfunctional since April 18, 2018.
Corrective Action(s): On May 2, 2018, the licensees fix-it-now team reperformed two steps of the test procedure, which reset the condition and restored the systems functionality. The licensee is conducting an investigation into the conduct of the April 18, 2018, test.
Corrective Action Reference(s): SAPNs 50976405 and 50978342
Performance Assessment:
Performance Deficiency: The failure to promptly identify and correct an adverse fire protection condition as required by the facility operating license and station procedures was a performance deficiency.
Screening: This performance deficiency was more than minor because it was associated with the protection against external factors 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.
Significance: Using IMC 0609, Appendix F, dated September 20, 2013, the team determined that this finding was of very low safety significance (Green) because the impact of the fire finding would be limited to no more than one train of equipment important to safety.
Cross-Cutting Aspect: This finding had a conservative bias (H.14) cross-cutting aspect in the human performance cross-cutting area because the test engineer and the work control shift foreman failed to use decision making practices that emphasized prudent choices when deciding that the as-left condition of the DG 1-1 cardox system was acceptable.
Enforcement:
Violation: License Condition 2.C.(5) (Unit 2 is 2.C.(4)), Fire Protection states that PG&E shall implement and maintain all provisions of the approved fire protection program that comply with 10 CFR 50.48(a). Title 10 CFR 50.48(a)(4)(b)(1)(i), states that fire protection features proposed or implemented by the licensee have been accepted by the NRC staff as satisfying the provisions of Appendix A to Branch Technical Position (BTP) APCSB 9.5-1 reflected in NRC fire protection safety evaluation reports issued before the effective date of February 19, 1981. In Section 9.5 of DCPP, Units 1 and 2, FSAR Update, Appendix 9.5B, Table B-1, Comparison of DCPP, Units 1 and 2, to Appendix A of BTP APCSB 9.5-1, Section C, Quality Assurance Program, the licensee states the following: Policies governing corrective measures relative to fire protection failures, malfunctions, deficiencies, deviations, defective components, uncontrolled combustible material, and nonconformances are addressed in administrative procedures. Procedure OM5, Quality Assurance Program, Revision 12, states that the fire protection quality assurance program will meet the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, which requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.
Contrary to this requirement, from April 18, 2018, through May 2, 2018, the licensees established measures failed to assure that a condition adverse to quality was promptly identified and corrected. Specifically, after performing maintenance on the cardox automatic fire suppression system for the DG 1-1 room, the system was left in a configuration in which it would have failed to achieve the required cardox concentration to suppress a fire. Licensee personnel failed to identify and correct this condition prior to it being identified by NRC inspectors.
Disposition: This violation is being treated as a NCV, consistent with Section 2.3.2. of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
On May 3, 2018, the inspectors presented the inspection results to Mr. J. Welsch, Vice President, Nuclear Generation and Chief Nuclear Officer, and other members of the licensee staff. The inspectors confirmed that any proprietary information reviewed was controlled to protect from public disclosure.
DOCUMENTS REVIEWED
Notifications (SAPNs)
276617 50672807 50860769 50868530 50882377 50915907 50959757
286030 50672808 50860840 50869181 50882498 50926817 50960226
50478663 50678944 50861002 50869475 50882634 50934271 50961430
50543796 50687004 50861050 50871484 50886666 50934650 50962228
50543797 50710432 50861097 50872056 50886703 50934855 50962275
50595631 50710790 50861161 50872133 50887238 50935077 50962645
50634825 50809593 50861200 50872733 50888276 50935776 50964666
50634990 50813807 50861706 50874462 50888383 50939069 50976326
50652094 50813849 50862239 50874890 50888976 50939087 50976395
50660171 50817678 50862530 50875802 50890598 50939302 50976632
50662101 50818595 50862531 50876600 50891169 50939305 50977343
50663322 50820956 50862731 50876610 50893239 50939391 50977345
50663604 50823081 50862811 50876729 50903456 50939394 50977676
50672438 50824398 50863138 50878466 50904953 50939582 50978179
50672439 50825789 50863710 50881414 50905315 50939608 50978308
50672490 50828441 50864453 50881415 50906259 50940114 50978338
50672491 50830829 50864563 50881612 50906321 50943781 50978496
50672492 50832335 50864636 50881613 50907166 50945776
50672804 50847067 50864895 50881653 50907301 50945777
50672805 50856300 50868502 50882125 50907664 50945779
50672806 50857023 50868506 50882140 50913334 50945780
Work Orders
60027837 60082568 60097574 60110300 64121123 64170447 68039185
60076901 60093023 60098162 60110505 64128758 68030083 68039186
60082567 60096050 60102679 64103044 64130174 68037723
Procedures
Number Title Revisions
AD1.ID2 Procedure Process Control 48B
AD2.ID1 Procedure and Work Plan Use and Adherence 26
AD7.ID2 Daily Notification Review Team and Standard Plant 29
Priority Assignment Scheme
AD9.ID7 Receipt Inspection and Acceptance Testing 8
B-5 Operation Section Policy 6
CF2.ID12 Cyber Security Assessment Team 0
CF3.ID13 Replacement Part Evaluation and CITE 30
CF4.ID7 Temporary Alteration 28
ER1.ID1 Equipment Reliability Process 7
MA1.DC54 Conduct of Maintenance 15
Procedures
Number Title Revisions
MA1.ID17 Maintenance Rule Monitoring Program 31
MA1.ID27 Preventive Maintenance Program 4
MP M-21.14 Fuel Control Linkage Assembly Routine Minor 4
Maintenance
MP M-21.40 Mechanical Maintenance Procedure 4-6
OM15.ID10 Performance Monitoring Program 0A
OM15.ID4 "Self-Assessment and Benchmarking" 14A
OM15.ID5 DCPP, Units 1 and 2, Performance Improvement 9F
Program
OM4.ID13 Nuclear Power Generation Internal Auditing 27
OM4.ID14 Notification Review Team 30, 30A
OM4.ID3 Operating Experience Program 27A, 28B, 29
OM6.ID12 Electrical Safety Program 27A
OM7.ID1 Problem Identification and Resolution 51
OM7.ID11 10 CFR Part 21, Reportability Review Process 4
OM7.ID12 Operability Determination 36
OM7.ID13 Technical Evaluations 7
OM7.ID3 Root Cause Evaluation 46A
OM7.ID4 Cause Evaluations 36A
OM7.ID7 Emerging Issue and Event Investigation 19
OM7-ID12 Operability Determination 37
OP1.DC3 Operator Routine Plant Equipment Inspections 14
OP1.DC40 Operations Equipment Deficiency and Adverse 10
Condition Monitoring
QCP 10.1 Receipt Inspection Program 17
STP M-78A Snubber Visual Inspection 23
STP M-9A3 Diesel Engine Generator 1-3 Routine Surveillance 10A
Test
STP M-9A3 Diesel Engine Generator 2-3 Routine Surveillance 11B
Test
STP P-AFW-FF23 Full Flow Exercise of Motor Driven Auxiliary 1
Feedwater Pump 2-3 Check Valve
TS5.ID1 Aging Management/System Walkdown Standard 23
TS5.ID1 System Engineering Program 27, 29
Drawings
Number Title Revision(s)
SK-H-017-07 Auxiliary Feedwater lines 1-4 6
System 3B Auxiliary Feedwater Maintenance Rule Scope 5
2-K16-568-6 Data Z-Snubber 2
054174, Optional Transmitter Mounting Plate Detail 65
Sheet 22B
054174, Fig 919B Optional Transmitter Mounting Plate Detail 66
Sheet 22B (Typical All Makes)
054174, Fig 919H, Mounting Details for Rosemount 1
Sheet 22J Transmitters PT-534A and PT-544A (Unit 2 Only)
59549 Unit 1 Area Drawing 29
2021, Sheet 5 Combustion Air and Exhaust System 58
106703 Unit 1 Auxiliary Feedwater Piping 81
108004 Auxiliary Feedwater Steam Supply Unit 2 86
433151, Sheet 1 Electrical Schematic Diagram, Control Room 16
Pressurization System
437529, Sheet 1 Electrical Single line Meter and Relay Diagram, 48
Generation Excitation Main and Auxiliary Transformer
437530, Sheet 1 Electrical Single Line Meter and Relay Diagram 12KV 39
Start-up System
437548 Electrical Schematic Diagram, Generator and Main 45
Transformer
437592, Sheet 1 Schematic Diagram, Residual Heat Removal Flow 38
Control Valves
441228 Single Line Meter and Relay Diagram 4160V System 16
Bus Section D and E
441229, Sheet 1 Single Line Meter and Relay Diagram 4160V System 20
Bus Section F DG 23
441308, Sheet 1 Schematic Diagram, Containment Spray System Motor 1B
Operated Valves
441309, Sheet 1 Schematic Diagram, Residual Heat Removal Pumps 24
441310, Sheet 1 Schematic Diagram, Residual Heat Removal Motor 32
Operated Valves
441317, Sheet 1 Schematic Diagram, Safety Injection System Motor 20
Operated Valves
441570, Sheet 1 Electrical Wiring Diagram, 4KV Switchgear Bus 19
Section F Cell 12
Drawings
Number Title Revision(s)
445096 Electrical Schematic Diagram, Control Room 3
Ventilation System
445097 Diagram of Connections Control Room Ventilation 1
Panel
449299 Auxiliary Feedwater Hanger Supports 15
460131 Unit 2 Pipe Restraint Modifications 3
500623 Diagram of Connections, Safeguard Relay Board Bus 12
Section H
501166, Sheet 1 Electrical Diagram of Connections Control Room 10
Pressurization Panel train F Auxiliary Relay
663312, Mechanical Pressure Relief Valve RV-353, 355, 356, 2
Sheet 187 358, 359, and 360
663312, Mechanical Pressure Relief Valve OMNI 900-Valve 2
Sheet 197 Outline Model 951280ME
Miscellaneous
Documents Revision or
Number Title Date
Diablo Canyon Power Plant, Units 1 and 2, Backup April 1, 2017
Air/Nitrogen System Health Reports - June 30,
2017
22-0041-RPT- Estimation of Minimum Installation Torque for DEG Fuel 0
001 Pump Inlet to Fuel Header Cap Screws
110000000053 Design Specification for Anderson-Greenwood/Crosby 9
Safety Relief and Relief Valves
8000000772 DCI, Relief Valves, OMNI 900 Series, Systems 14 and 25, 0, 8, 9
Anderson-Greenwood/Crosby
AFW-03B Maintenance Rule Scoping Functions Map 17
C21 D-23-126 Clearance for 1-23-E-S-EJPA Unit 1, train A Cont. Room April 30,
Press System, Xfer Switch 2018
CE Manual Cause Determination Manual May 18,
2017
Commitment Review of Newly Initiated ARs for POA Determination July 26, 2012
T35567
DCP-1-25418 Remove Emergency Core Cooling System Motor Operated 0
Valve External Limit Switch Interlock
Efficiency System Health Reporting Efficiencies December 2,
Bulletin 16-33 2016
Miscellaneous
Documents Revision or
Number Title Date
FCT 7*3930 Field Correction Transmittal for PT-534A/PT 544A 0
Mounting As-Built
IB 7.4.1.7-2 ABB Instructions Manual for Three Phase Under Voltage E
and Phase Sequence Relays
IL.41-766.5B Westinghouse Instruction Manual for Type SSV-T and B
SSV-T Relays for Class IE Application
MI-11188 ALCO, Engine Air Filter
MI-11209A ALCO, Turbocharger, Type 165 May 1973
MISC-TP16-1-112 Recommendation to Resolve Flowserve 10 CFR Part 21, 4
Anchor Darling Disc Gate Wedge Pin Failures
SAQH - ETR 2016 Maintenance Rule Periodic Assessment January 3,
2016 2017
SAQH- Maint 2016 2016 Maintenance Self-Assessment March 16,
2016
T-161010 Valve Test Results, Anderson Greenwood Crosby May 13,
2016
TU164R1 2016 ESP Continuing Training -Extent of Condition 0
INFORMATION REQUESTS
SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword: NRC-002
By: GAG Yes No Publicly Available Sensitive
OFFICE RIV/DRS RIV/DRS RIV/DRP RIV/DRP RIV/DRS RIV/DRS
NAME JBraisted NOkonkwo JReynoso MHaire ERuesch GGeorge
SIGNATURE /RA-e/ /RA-e/ /RA-e/ /RA/ /RA/ /RA/
DATE 6/1/2018 6/1/2018 6/1/2018 6/4/2018 6/5/2018 6/8/2018