25th Annual Report, Volume II, Exhibit D.4, Diablo Canyon Independent Safety Committee Report on Fact Finding Meeting at DCPP on November 19–20, 2014 by Robert J. Budnitz, Member, and R. Ferman Wardell, Consultant

1.0 Summary

The results of the November 19–20, 2014 fact-finding trip to the Diablo Canyon Power Plant in Avila Beach, CA are presented. The subjects addressed and summarized in Section 3 are as follows:

  1. Pressurizer Weld Overlay Issue
  2. Containment Fan Cooler Unit Modifications/Issues
  3. Fire Doors Update
  4. Intake Concrete Inspection and Repairs
  5. Safety Systems Functional Failures Update
  6. Outage 2R18 Results
  7. Radioactive Waste Systems Review and Walkdown
  8. Equipment Qualification Program Update
  9. Steam Generator Performance and Inspections through Outage 2R18
  10. Radiation Monitoring System Long-Term Strategy
  11. Observe NSOC Summary Session
  12. Meet with NRC Resident Inspector
  13. Dr. Budnitz’ Meeting with DCPP Chief Nuclear Officer

2.0 Introduction

This fact-finding trip to the DCPP was made to evaluate specific safety matters for the DCISC. The objective of the evaluation was to determine if PG&E’s performance is appropriate and whether any areas revealed observations which are important enough to warrant further review, follow-up, or presentation at a Public Meeting. These safety matters include follow-up and/or continuing review efforts by the Committee, as well as those identified as a result of reviews of various safety-related documents.

Section 4—Conclusions highlights the conclusions of the Fact-finding Team based on items reported in Section 3—Discussion. These highlights also include the team’s suggested follow-up items for the DCISC, such as scheduling future fact-finding meetings on the topic, presentations at future public meetings, and requests for future updates or information from DCPP on specific areas of interest, etc.

Section 5—Recommendations lists specific recommendations to PG&E proposed by the Fact-finding Team. These recommendations will be considered by the DCISC. After review and approval by the DCISC, the Fact-finding Report, including its recommendations, is provided to PG&E. The Fact-finding Report will also appear in the DCISC Annual Report.

3.0 Discussion

3.1 Pressurizer Weld Overlay Issue

The DCISC Fact-finding Team (FFT) met with Mike Leger, Lead In-service Inspection Specialist, for an update on DCPP’s Pressurizer Weld Overlay Issue. The DCISC last reviewed this topic at its Fact-finding Meeting in September 2013 (Reference 6.1) and its Public Meeting in October 2013 (Reference 6.2). At its September Fact-finding Meeting the DCISC concluded the following:

DCPP’s root cause evaluation and resultant corrective actions for the failure to detect small flaws in the Pressurizer nozzle structural weld overlay appear satisfactory. The DCISC should follow up in mid-2014 when actions have been completed, in particular to review the results of any finite element modeling performed to assess the overlay.

An “Indication” is a flaw or crack inside the weld that can be detected by reflections during ultrasonic test (UT). The key safety question for such flaws is whether they are sufficiently small that they would not be expected to grow in size during service. Very small flaws do not grow and do not present a safety hazard. If a flaw is sufficiently large that it could grow, then normally the weld material with the flaw would be removed by grinding and the welding repaired.

DCPP had applied pre-emptive structural weld overlays (SWOLs) to the Unit 2 Pressurizer nozzles’ dissimilar-metal butt welds during Refueling Outage 2R14 in March 2008. The overlays were applied using a provision from the American Society of Mechanical Engineers (ASME) Section XI In-service Inspection Code known as a relief request. The purpose of the weld overlays, which have been used in other plants, too, was to provide structural reinforcement of the original Alloy 600 SE weld areas, which had experienced Primary Water Stress Corrosion Cracking (PWSCC) elsewhere in the industry. The Unit 1 Pressurizer nozzles do not use Alloy 600 and do not have this issue.

The Pressurizer weld overlays were originally inspected following the welding in March 2008 using conventional UT exams (using several discrete ultrasonic angle beams), and they were inspected again in Outage 2R15 in October 2009 with similar UT exams with the exception that low angle detection was not required. During subsequent inspections in Outage 2R17 in February 2013 using more advanced UT techniques (phased array techniques), several new indications (flaws) were discovered that were outside the ASME Code allowable screening size. These flaws were determined to involve single weld passes, which meant that a Code-required flaw analysis be done, which was performed by AREVA under contract to PG&E. Using conservative assumptions, this analysis found that the flaw sizes were sufficiently small that the structures would be expected to provide satisfactory performance for at least an additional operating cycle. Review of the AREVA report by the DCISC Fact-finding team revealed that the analysis was satisfactory to demonstrate that no additional growth of the detected flaws would occur and to support continued operation for another operating cycle. An independent Electric Power Research Institute (EPRI) analysis supported this conclusion.

DCPP initiated a Root Cause Evaluation (RCE) to determine the reasons for not detecting the indications originally in Outages 2R14 and 2R15. The root cause was identified as:

A mismatch exists between the conventional UT weld overlay inspection procedure and the Performance Demonstration Initiative qualification process. Although the qualification process successfully demonstrated the ability to detect flaws, the procedure instructions do not adequately constrain the zero-degree scan speed to assure that small cross-section, low angle flaws are consistently detected in the field.

Contributing causes were that inattentive errors were made by vendor examiners for the following reasons:

  1. Data indicate that 45-degree angle beam was able to detect indications in the weld overlays, yet the indications were not recorded.
  2. Examiners failed to adequately investigate indication responses to determine the actual length of the flaw.
  3. Examiners failed to recognize zero-degree angle ergonomic factors necessitating reduced scan speed to maintain optimum search unit coupling.

The Corrective Action to Prevent Recurrence (CAPR) for the root cause was to revise the In-service Inspection Program procedure to not permit the conventional UT technique to be used for weld overlays until the recommendations for the first contributing cause have been addressed. These recommendations are:

For the UT Qualification Process

  1. Assure that scan speed, length sizing, and any other essential variables used during qualification testing are conservatively reflected in the examination procedure.
  2. Expand the sample set to include Westinghouse pressurizer nozzle configurations.
  3. Include more realistic oriented fabrication flaws in the test set.

For the UT Procedure

  1. Add guidance on when to reduce scan speed
  2. Evaluate the need to increase sensitivity for zero-degree examinations
  3. Include instructions related to detection of low-angle flaws

Additionally, DCPP will recommend to EPRI to publish a communiqué to all qualified examiners to review the causes and contributors of the DCPP event.

DCPP had submitted to the NRC a single cycle ASME Code Request for Relief (RFR) based on the initial analysis. This was approved, but NRC would need an additional analysis to support a request for long-term operation, looking at lateral crack growth in addition to the original circumferential cracks. This analysis produced acceptable results and was approved by NRC in an October 14, 2014 letter for continued operation to the year 2045.

DCPP performed re-examinations of the weld overlay in Outage 2R18 in the fall of 2014. The techniques and results were essentially the same as in 2R17, i.e., no crack growth. Additionally, DCPP committed to performing phased array examinations during the next three ISI inspection periods.

The DCISC Fact-finding Team recommends that this issue be closed.

Conclusions:
DCPP has satisfactorily completed its analysis of the Pressurizer weld overlay cracking issue to support continued operation until 2045 as approved by the NRC. The DCISC Fact-finding Team believes that this issue can be closed.
Recommendations:
None

3.2 Containment Fan Cooler Unit (CFCU) Modifications/Issues

The DCISC Fact-finding Team met with Greg Porter, System Engineer for Ventilation Systems, and Lou Fusco, Manager of Mechanical Systems, for an update on the CFCUs. The DCISC last reviewed CFCUs at the June 2014 DCISC Public Meeting (Reference 6.3) and at a Fact-finding Meeting in April 2013 (Reference 6.4) when it concluded the following:

DCPP discovered a damaged coupling on the 2-5 Containment Fan Cooler Unit (CFCU) during Outage 2R17. The damage did not adversely affect the CFCU’s safety function. The coupling was replaced, and the unit was returned to service with a temporary modification to restrict its fan speed to low speed while the root cause of the problem is determined. The DCISC should follow up on this issue.

DCPP had added anti-rotation devices to each CFCU to prevent reverse rotation. Reverse rotation is a potential problem because, if it were to occur above a prescribed amount, a start-up of the CFCUs could result in loss of the motors due to over-current. Unit 1 CFCU anti-rotation devices were installed during 2010 with satisfactory performance. A Unit 2 device was installed by May 2011, and by June noisy operation was evident, resulting in replacement with a spare. Shortly afterward two more devices were found to be noisy (ratchet pawls dragging), causing DCPP to write a Prompt Operability Assessment (POA) for justification of operation only at low speed. Performing an Apparent Cause Evaluation (ACE), DCPP and the vendor determined the devices are rubbing due to machining tolerance issues. Through the end of 2011 all devices were refurbished.

During Refueling Outage 2R17, a routine PM (Preventive Maintenance) inspection of the CFCU 2-5 coupling/anti-reverse rotation device (ARRD), the fan side coupling struts were discovered to have failed and the tension struts had buckled. Even with damage, CFCU 2-5 was determined to still be capable of performing its safety function. No problems were apparent on the remaining Unit 2 CFCUs, and no problems were noticed from inspections of Unit 1 CFCUs in outage 1R17. Thus there was no common failure. Following vendor inspection and analysis, it was determined that this damage could only have occurred due to application of reverse torque. The CFCU 2-5 damaged coupling was replaced with a spare.

DCPP hired a consultant to perform a failure analysis. The consultant concluded that the coupling failed due to a tensile overload resulting from a torque applied in the reverse direction, which was most likely caused by a shift of the CFCU motor from High to Low speed while the fan was rotating at more than the low speed of ∼ 600 revolutions per minute (rpm). DCPP performed a temporary modification to restrict the 2-5 CFCU to low speed while the investigation continues into the cause of the damaging speed change. The CFCU safety function, cooling of Containment following a loss of coolant accident, uses low speed. High speed is used for normal Containment cooling, and compensatory measures have been taken to assure that function is maintained.

There is more work to be done on the CFCUs including adjusting the timing sequence to address the anti-rotation device problem. In the meantime the CFCUs are run only in low speed. Design changes are also required to the CFCU cooling coils to upgrade and replace the current coils. Along with replacing the cooling coils, the plant will implement design changes to the inlet dampers to the CFCUs to meet the requirements of the cooling coils.

The anti-rotation devices are currently working well. The fan/motor couplings are not designed for instant slowdown from 1200 to 600 rpm in going from high to low speed. A design change is being issued to improve the delay time for speed changes and to implement a sequencing scheme when on emergency power. These design changes are scheduled for completion by mid-2015.

Some CFCUs had experienced high vibration at higher speeds due to damper changes to reduce air flow to reduce the potential for Component Cooling Water (CCW) overheating. DCPP will replace the cooling coils and modify the dampers to accommodate the reduced airflows. Finally, DCPP has a CFCU coil replacement program due to aging and corrosion. The Unit 1 and 2 coils are scheduled for replacement in Outages 1R19 and 2R19, respectively.

Conclusions:
DCPP appears to have satisfactory solutions to problems with its Containment Fan Cooler Unit Fans. The DCIAC should continue to follow this issue after each refueling outage.
Recommendations:
None

3.3 Fire Doors Update

The DCISC Fact-finding team met with Dave Hampshire, Fire Protection Supervisor; Alex Arsene, Appendix R Program Engineer; and Al Clark, Civil Engineering Door Systems Engineer, for an update on impaired fire doors. The DCISC last reviewed Fire Doors in March 2014 (Reference 6.5), concluding the following:

The DCISC learned in December 2013 that 16 impaired fire doors would not be repaired or replaced until 2017 due to funding deferrals and found this unacceptable. Following up in March 2014, the DCISC found that six doors had been repaired or replaced, and the remaining ten were the highest priority on the Plant Door Life Cycle Management Plan. The ten impaired doors are compensated for by fire watches, which, while acceptable, are not desirable. This is an acceptable start, and the DCISC should follow up on this issue near the end of 2014.

At the March 25, 2014 Plant Health Committee meeting the Appendix R Program Manager reported that this fire protection program health was Red, unsatisfactory, due to the following:

Excessive Critical Component Failure/Adverse Equipment Trend (one or more critical component failures without an action plan) because of 16 impaired fire doors for several years due to financing deferrals. The impaired doors require fire watches, an unsatisfactory long-term substitute for fully functional fire doors. DCPP has an action plan to replace/repair these doors, but funding has been deferred through 2016. This deferral was a concern to the DCISC, and the earlier Fact-finding Team recommended that the DCISC look further into the deferrals.

There are a total of 94 doors needing replacement. Of these, nine are Appendix R fire doors with compensatory measures in place consisting of roving fire watches. An additional 31 doors are in the DCPP Equipment Control Guidelines (ECGs) as doors which cannot be repaired and require replacement. The funding for these doors in the original Door Replacement Program had been deferred from 2012 until 2017, which appeared unacceptable to the DCISC. Six of these 16 doors have now been repaired or replaced, leaving 10 doors needing resolution. These ten remaining doors have been included as highest priority in the Plant Door Life Cycle Management Plan.

A new “Power Block Door Project” was presented in July 15, 2014 to the Project Review Committee for funding. This Project included replacement of all 94 doors in the Power Block because they had outlived their useful life, i.e., they had degraded to the point where they can no longer be repaired to meet the design safety function.

The Project Review Committee, in its July 15, 2014 meeting, approved including the 2015 Power Block Project scope in the DCPP Five Year Plan and review additional funding in the future.

Conclusions:
The DCISC concern regarding the needed, but delayed, replacement of fire doors and other safety function doors has been somewhat alleviated by DCPP funding for the new Power Block Project high-priority doors for 2015 and consideration of additional funding for future years. The DCISC notes that 6 of the 16 highest priority fire doors have been replaced. The DCISC should continue to monitor the replacement of DCPP fire and other safety function doors.
Recommendations:
None

3.4 Intake Concrete Inspections and Repairs

The DCISC Fact-finding team met with Behrooz Shakibnia, Civil Engineering Supervisor; John Fonturbel, Civil Engineer; and Tom Voss, Consulting Concrete Inspector, for an update on DCPP Intake Structure concrete inspections and repairs. The DCISC last reviewed Intake Concrete in June 2013 (Reference 6.6), concluding the following:

DCPP’s concrete repair procedure and repairs of concrete in the Intake Structure appeared satisfactory.

Because of the saltwater environment, the concrete intake structure can deteriorate when corrosion of rebar occurs, which causes swelling of the rebar and concrete spalling, which then further exposes the steel reinforcing bar, causing the degradation to accelerate. This reduces structural integrity. DCPP has a program to inspect and repair the damage to assure structural integrity. During Outage 2R17, the repair work consisted of the following:

The PG&E Applied Technology Services Group performs inspections and soundings of the concrete, identifying areas needing repair. The Saltwater Structural Engineering Group makes determinations of the soundness of structures. The Intake Repair Program Group makes the repairs specified by the other two groups.

DCPP’s Procedure MIP C-7.0, “grouting and Repair of Concrete Defects,” Revision 3 governs the repair process. In addition to providing definitions and responsibilities, the procedure specifies the following:

The DCISC found the procedure to be comprehensive and detailed.

In this November 2014 visit the DCISC was interested in concrete inspections and repairs during Refueling Outages 1R18 and 2R18. In these outages work was focused on the common walls between Unit 1 and 2 Circulating Water Conduits and the Auxiliary Saltwater (ASW) System Pump bypass walls. In 1R18 89 square feet of concrete was repaired, but repair of a section of 29 square feet was not currently necessary and was deferred to the future. In 2R18 similar inspections were made, but no repairs were necessary. The DCISC reviewed the resulting inspection and repair reports, which were satisfactory.

Conclusions:
DCPP’s inspections and repairs of the Intake Structure concrete appeared appropriate to assure continued operation of the structure for normal and emergency functions.
Recommendations:
None

3.5 Safety System Functional Failures Update

The DCISC met with Tom Baldwin, Manager of Regulatory Services, for an update on DCPP Safety System Functional Failures (SSFFs). The DCISC last reviewed SSFFs in March 2014 (Reference 6.7), concluding the following:

DCPP’s performance on reducing or eliminating Safety System Functional Failures (SSFFs) has not improved despite implementation of a corrective action plan. This is a DCISC concern. A new plan has been developed, and the DCISC should review this item in the third quarter of 2014.

A Safety System Functional Failure (SSFF) is defined as “The failure of or the loss of the ability of a system safety function to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive materials, or mitigate the consequences of an accident.” Therefore, a safety system may meet a Technical Specification (TS) limiting condition for operation (LCO), but exhibit an SSFF at the same time.

The recent history of this issue began in 2001 when the Nuclear Regulatory Commission (NRC) changed the significance of a SSFF event when it established a new Reactor Oversight Program (ROP) that, among other things, uses performance indicators for key parameters, including SSFFs. Depending on the number of SSFFs that a plant experiences, the plant will receive a varying level of regulatory oversight. For, example, if a plant experiences five SSFFs within a rolling four quarter period, the plant will move into the White regulatory response column and receive greater NRC oversight.

Between Ju1y 1, 2010 and August 31, 2011, DCPP Units 1 and 2 experienced a combined total of 12 SSFFs. Of these 12 SSFFs, four were common to both units. There was considerable variety in the nature of the SSFFs. Some examples include the following:

DCPP’s Root Cause Evaluation (RCE) Report of these SSFFs further notes that, beginning with the discovery of incorrect open limit switch settings on motor-operated Emergency Core Cooling System (ECCS) sump suction valves in 2009, “DCPP experienced multiple events that resulted in the loss of a system safety function to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive materials or mitigate the consequences of an accident.”

DCPP’s examination of this issue in its Root Cause Evaluation (RCE) was extensive and detailed, and included reviews of operating experience within the industry. The examination concluded that DCPP lacked clear standards for risk assessment, risk evaluations, and risk mitigation activities that could, and did, result in SSFFs. It further concluded that, when reviewing evaluations, the station had a tendency to justify and accept the evaluations rather than to provide a healthy challenge to them. It also noted that opportunities had been missed to reinforce high standards, that resolutions of identified risks were sometimes incomplete, and that there was sometimes no means or expectation for identifying risk significant activities. A contributing cause identified by the station was that “station personnel had insufficient understanding of the definition of an SSFF, resulting in failure to recognize that adherence to station procedures and plant Technical Specification action requirements does not prevent SSFFs.”

To address the root and contributory causes of this adverse trend in SSFFs, DCPP developed 30 planned actions, which collectively comprise one of the eight areas for improvement in a broader “Regulatory Excellence Action Plan.” The first major component of the Action Plan to address Safety System Functional Failures involved completing the RCE which resulted in its March 7, 2012 Action Plan, which contained 30 major and supporting actions.

The purpose of the March 2014 fact-finding visit was to assess DCPP’s progress on reducing the number of SSFFs. The RCE effectiveness evaluation concluded that the corrective actions in the Action Plan were not effective because of an increased number of SSFFs. Prior to the RCE, all SSFF events were designated as preventable. Following the RCE, five of nine events were designated as preventable, and the remaining four would have been preventable had the corrective actions been effectively implemented. DCPP found no commonality of causes. DCPP has taken new, augmented corrective actions to the Corrective Action Review Board, which contained processes to preclude SSFFs from happening from initiating events. These new corrective actions include the following:

  1. Update the applicable procedure to include all modes of operation and expand the list of Single Failure Vulnerable Systems to include shared portions of systems that create a single point vulnerability.
  2. Establish risk mitigation actions for any condition, which reduces vulnerability to SSFF to loss of a single component, power supply, or train.
  3. Establish the Station Focus Area that includes the top five human performance error prevention tools (the “High Five”).
  4. Post Systems, Structures, and Components (SSCs) where the loss of the component would result in loss of SSF, and revise the Operations Policy to reflect this standard.
  5. Require Outage Scope Review Team (OSRT) identification of SSFF vulnerabilities and establishment of risk-commensurate mitigations when repair will be delayed or deferred.
  6. Develop and proceduralize a clear standard for evaluations of conformance to licensing basis and SSFF vulnerability to be implemented in Operating Experience Assessments, plant modifications, design and licensing basis reviews, NRC communications, and Licensing Basis Verification Program processes.
  7. Educate station Senior Reactor Operators, managers, senior leadership team, and engineers such that they can recognize a SSFF or potential SSFF challenge.
  8. Communicate to station to achieve plant-wide recognition of DCPP SSFF Performance Indicator vulnerability, including:
  1. Current station SSFF performance
  2. Bottom industry decile standing
  3. How to recognize an SSFF vulnerability
  4. Expectations to reduce risk of SSFF events
  5. Broad range of situations whereby plant staff can create a possible SSFF event.

With the relatively high number of SSFFs recently, DCPP needs strong, effective correction action to reverse the degrading trend of SSFFs. The DCISC should follow this issue closely with quarterly reviews to assess the effectiveness of corrective action.

The trend of SSFFs for the last two years at DCPP is as follows:

Quarter Unit 1 SSFFs Unit 2 SSFFs
1Q13 3 3
2Q13 3 4
3Q13 3 4
4Q13 3 3
1Q14 4 2
2Q14 5 2
3Q14 3 1
4Q14 3 2

NRC’s four-quarter Performance Indicator for DCPP’s SSFI is currently Green based on the following data:

Unit
No. of SSFIs NRC White Threshold DCPP Goal
1 4 > 5 0
2 2 > 5 0

Conclusions:
It appears to the DCISC Fact-finding Team that there has been little improvement in DCPP System Functional Failures (SSFFs) since July 2010, when originally reviewed. This has been and is still a concern to the DCISC. The DCISC should review the SSFF status in mid-2015.
Recommendations:
It is recommended that DCPP review again the causes of its Safety System Functional Failures (SSFFs) and develop and implement corrective action which will eliminate or significantly reduce the number of SSFFs.
Basis for Recommendation:
Since March 2010, when the DCISC began its review of DCPP SSFFs, DCPP has made little progress in reducing the numbers of its SSFFs as shown below for the last two years:

Quarter Unit 1 SSFFs Unit 2 SSFFs
1Q13 3 3
2Q13 3 4
3Q13 3 4
4Q13 3 3
1Q14 4 2
2Q14 5 2
3Q14 3 1
4Q14 3 2

Although the numbers are small, the DCISC believes the number should be close to zero. DCPP is working on its second action plan; however, there have been little or no positive results to date.

3.6 Outage 2R18 Results

The Fact Finding team met with Matt Coward, Outage Manager, to review the results of DCPP’s 2R18 Refueling Outage, which began on September 28 and ended slightly ahead of schedule on October 31, 2014. The DCISC last reviewed refueling outages in May 2014 (Reference 6.8), when it concluded the following:

DCPP’s Outage 1R18 results were positive with the one exception of temporary loss of the Unit 1 Spent Fuel Pool Cooling Pump due to an electric grid disturbance. Operators restarted the pump, and there were no safety consequences of the event.

Outage goals and results were as follows:

Performance Category Goal Actual
Recordable & Disabling Injuries  0  0
Nuclear Safety Events  0  0
Human Events Clock Resets  0  0
Outage Duration (days) ≤ 33  32.4
Dose Goal (Person-Rem) 32 30.37 
Significant Foreign Material Events (FME)  0  0
Power Ascension (days) ≤ 5 4.7
Reliable Run at 100% (days) ≥ 90  TBD

Major Reliability Scope items include the following:

Conclusions:
DCPP’s 2R18 Refueling Outage met essentially all goals and was considered a success by DCPP. The DCISC considers it a success from a nuclear safety perspective.
Recommendations:
None

3.7 Radioactive Waste Management Systems Review and Walkdown

The DCISC Fact-finding Team met with Clint Miller, Solid and Liquid Radwaste System Engineer, and Surendra Sabharwal, Gaseous Radwaste System Engineer, for a review of the status of DCPP’s Radioactive Waste (Radwaste) Management Systems. The DCISC has not reviewed these systems recently.

The DCISC Fact-finding Team (FFT) reviewed the design and operation of the three radwaste systems and toured the accessible portions of the liquid and gaseous systems. The systems appeared to be in good working order, and the plant conditions in these areas appeared to be acceptable.

Gaseous Radwaste System (GRWS) Status

The GRWS collects and processes radioactive gases from various plant systems and includes two large waste gas decay tanks which hold radioactive gases for time to decay to low levels of radioactivity. The GRWS then discharges small amounts of gaseous radioactivity to the environment via the Plant Vent System. These discharges are monitored by a pre-set radiation detector, and, if too high, the discharges are automatically terminated. Annual reports of radioactive gas discharges have all shown that discharges are very small fractions of DCPP Technical Specifications and NRC regulatory limits.

The GRWS is healthy (White) according to the System Health Report. The major issue is Waste Gas Sampling Subsystem inoperability, which is due to obsolescent system components preventing continuous monitoring of oxygen levels upstream of the gas compressor when the compressor is shutdown, because the oxygen analyzer is located downstream. Manual samples are taken as a compensatory measure. Funding has been on hold until January 2015 due to budget constraints. It is expected that design, procurement and installation will commence in 2015 and be complete by the end of 2016.

Liquid Radwaste System (LRWS) Status

Liquid radwaste is processed and reduced by way of filters, demineralizers and evaporators. The clean water is recycled back into plant systems, and the spent resins and filters are input to the Solid Radwaste System. Small remaining amounts of radioactive liquids are diluted, measured for radioactivity, and discharged into the Pacific Ocean via the Auxiliary Saltwater System discharge. Annual reports of liquid radioactive releases have all shown that releases are very small fractions of DCPP Technical Specifications and NRC regulatory limits.

The LRWS is healthy (White) according to the System Health Report. The major issues are the nitrogen supply to the Reactor Coolant Drain Tank and level detectors for the Spent Resin Storage Tanks (SRSTs). New level probes for the SRSTs are scheduled to be installed during 2015.

Solid Radwaste System

The Solid Radwaste System (SRWS) collects and processes (decontamination, drying, compaction and packaging) solid radioactive materials for eventual shipment to licensed burial facilities. The solids are mostly spent resins, used filter media, and miscellaneous paper, cloth and other solids. The radioactive solids are handled remotely in shielded facilities. There are no solid radioactive wastes discharged into the environment.

SRWS health is White—healthy. There are a variety of issues needed for Green status. These are typically crane parts, lights, cooling fans, shielding, etc., whose repair, upgrading, or replacement is not central to achieving safe operation.

Conclusions:
The DCPP radioactive gas, liquid, and solid waste management systems are all healthy (White) each with minor issues which are being addressed.
Recommendations:
None

3.8 Equipment Qualification Program Update

The DCISC Fact-finding Team met with Kyle Millenaar, Engineer in Instrumentation & Controls Engineering and Equipment Qualification Program (EQP) Coordinator, for an update on the DCPP Equipment Qualification Program. The DCISC last reviewed the EQP in November 2012 (Reference 6.9) when it concluded the following:

The DCPP Environmental Qualification (EQ) program appears to be healthy. The self-assessment of the program conducted during the third quarter of 2012 was extremely thorough and found no maintenance deficiencies that challenge the environmental qualification of equipment. Minor deficiencies identified in the self assessment are being addressed through DCPP’s Corrective Action Program. Although no significant problems exist, the number of open Notifications has been increasing in recent years, and the expected attrition of knowledgeable individuals could aggravate this situation, along with having a potentially negative impact on the Program. The DCISC should continue examining the EQ program at least every two years.

The EQ Program is part of the Electrical Engineering Department. It is an industry-wide program; and at DCPP it is controlled by Procedure CF3.ID3, “Environmental Qualification (EQ) Program,” which implements Title 10 of the U.S. Code of Federal Regulations, Part 50.49 (10CFR50.49). This requires the generation and maintenance of evidence to ensure that electric equipment important to safety will operate when required to meet system performance requirements when subjected to expected environmental conditions. This includes mostly electrical equipment located where environmental conditions could be harsh during normal or postulated accidents, such as high temperature, high radiation, water spray, steam, conditions, etc. The procedure specifies the design bases for environmental conditions in various locations of the plant, the EQ Master List, applicable departmental procedures, deficiency identification and resolution, documentation requirements, and records retention. The procedure lists responsibilities for Engineering, Operations, Maintenance, Procurement, Learning Services, Document Services, and Quality Verification personnel for their parts of the program.

The EQ Procedure includes the following:

The DCISC Fact-finding Team reviewed the current revision of the procedure and found it appropriate for the task.

The DCISC learned that the only current I&C (Instrumentation and Controls) engineer qualified for EQ determinations had retired and his replacement had also left, and a new engineer was in training to become fully qualified prior to the retirement. Mr. Millenaar shared with the Team the two following training documents:

  1. Task Qualification Guide “Perform Tasks Associated with Performing Environmental Qualification (EQ) Related Engineering Activities”
  2. Task Qualification Guide “Perform EQ Maintenance Activities”

The guide includes all aspects of EQ, e.g., EQP scope, EQ Master List, requirements for various equipment, vendor qualification, EQ-related calculations, and EQ files. These two guides, which included both training and mentoring by a qualified engineer, appeared comprehensive and appropriate.

Some current activities underway include the following:

  1. Testing and evaluation is being conducted to qualify switchgear in the 4 kV Switchgear Room for High Energy Line Break (Main Steam Line Break) conditions, primarily steam and high humidity. This is a legacy issue.
  2. Rosemount Transmitters are being replaced because they are near their end of life. DCPP is qualifying the new transmitters for particular environmental conditions.
  3. As part of the Life Extension Program, Raychem splices are being reviewed for replacement. Although their end-of-life is not close, they are being replaced along with the particular end-of-life component which incorporates them.

The Unit 1 EQ life extension review has been completed and report being written. The DCISC should review the report in a future fact-finding meeting.

The EQ Program requires the EQ Program Coordinator to prepare a self-assessment (S-A) report following each Unit 2 refueling outage. The most recent report dated March 18, 2014 covers the period October 4, 2012 through January 13, 2014. The S-A serves as the program “Health card.” The report focused on significant work items “Replaced/reworked due to corrective maintenance” rather than recurring items such as transmitter calibrations and scheduled EQ component or equipment replacements such as position switches and solenoid valves. The S-A concluded that

“…there were no identified adverse trends in the qualification or in the maintenance of EQ equipment.” The S-A also concluded that the program complies with the NRC EQ regulation 10CFR50.49 regulation. The following major issues were identified:

DCPP believes that they will need to develop a “Field Guideline for EQ Inspection and Walkdown” for the younger, less experienced engineers as the experienced engineers are retiring. Developing the field guide will be resource intensive.

One significant challenge for EQ involves the Containment Fan Cooler Unit (CFCU) cooling coils. The CFCU’s function is to cool the Containment during normal operation and accident conditions. The coils will be replaced between the 2R18 and 2R20 refueling outages. Replacement of the coils will change the Containment environment temperature, which may require re-qualification of instruments inside Containment. This work will be performed by a contractor.

Also, the CFCU fan motors need replacement because their mechanical capability is considered poor. Qualified motors are available but will require an economic and engineering evaluation and an executive decision to move forward.

The following are S-A identified and reviewed EQ devices replaced or reworked due to corrective maintenance:

  1. Reactor Vessel Vent Valve RCS-2-8078C&D exhibited a failed seat leak test during a shop test and was repaired.
  2. Reactor Coolant System (RCS) Loop 1 Temperature Monitor TM-413A failed low reading, and a temporary modification was issued until the Resistance Temperature Detector was replaced in Outage 1R18.
  3. RCS Loop 3 temperature element TE-433A exhibited an erratic reading, and a temporary modification was issued until TE-433A was replaced in Outage 1R18.
  4. Rebuild of the valves on RCS-SOV-5 spool piece was performed because the valves failed their stroke test. A repaired spool piece was installed in Outage 2R17.

The S-A concluded that the above activities were properly performed.

Conclusions:
The DCPP Equipment Qualification Program appeared satisfactory. Because of an upcoming retirement, a new engineer is being qualified for the process.
Recommendations:
None

3.9 Steam Generator Performance and Inspections through Outage 2R18

The DCISC Fact-finding team met with John Ahar, DCPP Steam Generator (SG) System Engineer, for an update on SG performance and health and results of inspections in refueling outages 1R18 and 2R18. The DCISC last reviewed SGs in August 2013 (Reference 6.l0), when it concluded the following:

DCPP has established high performance goals for feedwater and steam generator chemistry and appears to be exercising effective control of feedwater and steam generator water chemistry. A few recent issues related to Unit 1 Steam Generator sulfates and Feedwater iron appear to have been effectively addressed. This topic continues to be a reliability issue rather than a safety issue. Results in DCPP’s new steam generators indicate no impact on reliability. Unless problems emerge in this area, the DCISC should defer its next review of this topic until at least mid-2015.

The four DCPP SGs per unit were replaced in outages 2R14 (Unit 2) in 2008 and 1R15 (Unit 1) in 2009 and have been performing as expected. One of the most important SG parameters is the integrity of the 4444 0.75-inch diameter Alloy 690 tubes in each SG. The tubes serve as the pressure boundary between the Reactor Coolant and the Main Steam and Feedwater Systems. Visual and Eddy Current Testing (ECT) inspections of 100% of the tubes have been performed in refueling outages 2R15 and 1R16 with only one tube in each unit showing minor indications of cracks. Inspections of 100% of the tubes in outages 1R18 and 2R18 resulted in 15 tubes showing minor indications. After evaluation, all were left in place. The next inspections are required to be in 1R21 and 2R21.

Sludge lancing of mineral build-up on the tubes resulted in a very small (∼ three pounds) amount of material per unit.

DCPP’s Condition Monitoring Assessments, required following each outage SG inspection, had the following conclusions:

The condition monitoring (CM) assessment concluded that, based on the results of the 2R18 inspections, none of the SG performance criteria were exceeded since the last ECT inspection in 2R15. That is, the three cycle operating period between the start of the Unit 2 Cycle 1 and the end of Unit 2 Cycle 18. The operational assessment (OA) concludes that there is reasonable assurance that operation of the DCPP Unit 2 SGs until the next scheduled ECT inspection in 2R21 (three operating cycles) will not cause any of the SG performance criteria to be exceeded.

There was a similar assessment written for Unit 1 following outage 1R18.

Conclusions:
The DCPP Steam Generators (SGs) have been performing as expected since their replacement in 2008 and 2009. The most important SG parameter, tube integrity, has been shown to meet all criteria as a result of visual inspection and Eddy Current testing.
Recommendations:
None

3.10 Radiation Monitoring System Long-Term Strategy

The DCISC Fact-finding Team met with Kevin O’Neil, I&C Supervisor and Radiation Monitoring System Engineer, for a review of the DCPP Radiation Monitoring System health and Long-Term Strategy. The DCISC learned about the long-term strategy at a Plant Health Committee meeting in December 2013 (Reference 6.11) as follows:

The Radiation Monitoring System health is White (satisfactory) for Unit 1 and Yellow (unsatisfactory) for Unit 2 due to equipment reliability problems due to the age of components; however, obsolescence is not considered a problem because spare parts are readily available. Unit 2 additionally has had operability problems with the Plant Vent and Containment air particulate monitors. An integrated system asset replacement initiative will be performed concurrent with the DCPP Unit Relicense period; meanwhile, DCPP is developing a plan to manage and improve system health in the interim. A long-term strategy is scheduled for presentation to the PHC in mid-2014. [It was learned subsequently that the PHC approved funding.]

The existing Radiation Monitoring System (RMS) consists of 101 channels of radiation detectors and associated electronic components, and wiring located all around the plant. The system components come from four manufacturers. The system ranges in age from the 1970s to the 1990s and consists of both analog and digital components. Although there is a good supply of spare parts, there have been enough maintenance and reliability and availability problems for DCPP to develop a long-term radiation monitoring strategy. DCPP believes the performance of the system is acceptable, and the system is rated Healthy (White). With corrective actions both the reliability and availability improved noticeably in the fourth quarter of 2013 and have been very good during 2014.

The DCPP long-term RM strategy is under way in Engineering with an April 1, 2015 completion date. The purpose is to improve reliability and reduce the maintenance burdens. A presentation to the Plant Health Committee is set for July 1, 2015. The DCISC should review the strategy in mid-2015.

Conclusions:
The DCPP Radiation Monitoring System, consisting of both analog and digital components dating back to the 1970s, 1980s, and 1990s, has had availability and reliability problems up until the fourth quarter of 2013, when corrective actions resulted in noticeable improvements. For sustained improvements DCPP Engineering is developing a Long-Term Radiation Monitoring Strategy scheduled for completion in mid-2015. The DCISC should review that strategy at that time.
Recommendations:
None

3.11 Nuclear Safety Oversight Committee (NSOC) Summary Meeting

The DCISC has an agreement with DCPP to maintain NSOC information confidential, thus only limited information is presented here.

The DCISC Fact-finding Team attended the DCCPP NSOC summary session with plant management. The summary session followed three-and-a-half days of NSOC subcommittee meetings with plant personnel and observations of plant activities. The DCISC last attended an NSOC meeting in January 2011 and reported the following at its February 15, 2012 Public Meeting (Reference 6.12):

Dr. Budnitz reported he attended the January 19, 2011 NSOC meeting and commented there was a good deal of discussion about security questions which are outside the purview of the DCISC. Dr. Budnitz stated he was impressed by the thoroughness and quality of the NSOC team and he remarked he has been acquainted with two current members of NSOC for 35 years. Mr. David reported that the day prior to a NSOC meeting is devoted to meetings of the NSOC subcommittees, which include DCPP director and manager level personnel.

The NSOC subcommittees consisted of the following functional areas:

NSOC Functional Areas

The DCISC fact-finding team found that attending the NSOC review meeting was useful to the DCISC, because several of the issues that the NSOC reviewed are similar to issues that the DCISC reviews.

Conclusions:
Attending NSOC meetings is an excellent way for the DCISC to learn about various plant issues, and therefore the DCISC should plan to attend them regularly.
Recommendations:
None

3.12 Meeting with NRC Resident Inspector

The DCISC Fact-finding Team met with John Reynoso, NRC Resident Inspector at DCPP. The DCISC last met with the NRC in April 2014 (Reference 6.13) and concluded the following:

DCISC meetings with the NRC Resident or Senior Resident Inspector continue to be beneficial with regard to sharing information and to understanding issues important to the NRC and DCPP.

The discussion centered on the following subjects:

Conclusions:
DCISC meetings with NRC resident inspectors continue to be useful for sharing concerns and for reporting the results of reviews and activities.
Recommendations:
None

3.13 Dr. Budnitz Meeting with DCPP Chief Nuclear Officer

Dr. Budnitz met with DCPP Chief Nuclear Officer, Ed Halpin, to discuss items from this fact-finding and other items of mutual interest.

4.0 Conclusions

4.1
DCPP has satisfactorily completed its analysis of the Pressurizer weld overlay cracking issue to support continued operation until 2045 as approved by the NRC. The DCISC Fact-finding Team believes that this issue can be closed.
4.2
DCPP arrears to have satisfactory solutions to problems with its Containment Fan Cooler Unit Fans. The DCIAC should continue to follow this issue after each refueling outage.
4.3
The DCISC concern regarding the needed, but delayed replacement of fire doors and other safety function doors has been somewhat alleviated by DCPP funding for the new Power Block Project high-priority doors for 2015 and consideration of additional funding for future years. The DCISC notes that six of the highest priority 16 fire doors have been replaced. The DCISC should continue to monitor the replacement of DCPP fire and other safety function doors.
4.4
DCPP’s inspections and repairs of the Intake Structure concrete appeared appropriate to assure continued operation of the structure for normal and emergency functions.
4.5
It appears to the DCISC Fact-finding Team that there has been little improvement in DCPP System Functional Failures (SSFFs) since July 2010, when originally reviewed. This has been and is still a concern to the DCISC. The DCISC should review the SSFF status in mid-2015.
4.6
DCPP’s 2R18 Refueling Outage met essentially all goals and was considered a success by DCPP. The DCISC considers it a success from a nuclear safety perspective.
4.7
The DCPP radioactive gas, liquid, and solid waste management systems are all healthy (White) each with minor issues which are being addressed.
4.8
The DCPP Equipment Qualification Program appeared satisfactory. Because of an upcoming retirement, a new engineer is being qualified for the process.
4.9
The DCPP Steam Generators (SGs) have been performing as expected since their replacement in 2008 and 2009. The most important SG parameter, tube integrity, has been shown to meet all criteria as a result of visual inspection and Eddy Current testing.
4.10
The DCPP Radiation Monitoring System, consisting of both analog and digital components dating back to the 1970s, 1980s, and 1990s, has had availability and reliability problems up until the fourth quarter of 2013, when corrective actions resulted in noticeable improvements. For sustained improvements DCPP Engineering is developing a Long-Term Radiation Monitoring Strategy scheduled for completion in mid-2015. The DCISC should review that strategy at that time.
4.11
The Nuclear Safety Oversight Committee summary meeting with DCPP management was focused on achieving excellence. The NSOC concerns and recommendations appeared appropriate.
4.12
DCISC meetings with NRC resident inspectors continue to be useful for sharing concerns and for reporting the results of reviews and activities.
5.0 Recommendations:
5.1:
It is recommended that DCPP review again the causes of its Safety System Functional Failures (SSFFs) and develop and implement corrective action which will eliminate or significantly reduce the number of SSFFs.
Basis for Recommendation:
Since March 2010, when the DCISC began its review of DCPP SSFFs, DCPP has made little progress in reducing the numbers of its SSFFs as shown below for the last two years:

Quarter Unit 1 SSFFs Unit 2 SSFFs
1Q13 3 3
2Q13 3 4
3Q13 3 4
4Q13 3 3
1Q14 4 2
2Q14 5 2
3Q14 3 1
4Q14 3 2

6.0 References
6.1
“Diablo Canyon Independent Safety Committee Twenty-Third Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2012—June 30, 2013”, Approved October 9, 2013, Volume II, Exhibit D.3, Section 3.11, “Pressurizer Weld Overlay Indication Update.”
6.2
Ibid., Volume II, Exhibit B.63, “Pressurizer Weld Overlay Indication Update.”
6.3
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2013—June 30, 2014”, Approved October 22, 2014, Volume II, Exhibit B.9, “Containment Fan Cooler Unit.”
6.4
“Diablo Canyon Independent Safety Committee Twenty-Third Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2012—June 30, 2013”, Approved October 9, 2013, Volume II, Exhibit D.8, Section 3.6, “Containment Fan Cooler Unit Anti-Rotation Couplings.”
6.5
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2013—June 30, 2014”, Approved October 22, 2014, Volume II, Exhibit D.7, Section 3.5, “Fire Door Issues.”
6.6
Ibid., Exhibit D.1, Section 3.13, “2R17 Intake Concrete Work.”
6.7
Ibid., Exhibit D.7, Section 3.7, “Safety System Functional Failures.”
6.8
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2013—June 30, 2014,” Approved October 22, 2014, Volume II, Exhibit D.9, Section 3.4, “Outage 1R18 Performance Results.”
6.9
“Diablo Canyon Independent Safety Committee Twenty-Third Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2012—June 30, 2013”, Approved October 9, 2013, Volume II, Exhibit D.4, Section 3.3, “Environmental Qualification (EQ) Program Update.”
6.10
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2014—June 30, 2015,” Approved October 20, 2015, Volume II, Exhibit D.2, Section 3.9, “Feedwater Chemistry and Steam Generator Health.”
6.11
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2013—June 30, 2014,” Approved October 22, 2014, Volume II, Exhibit D.5. Section 3.1, “Observe Plant Health Committee Meeting.”
6.12
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2010—June 30, 2011,” Approved October 22, 2011, Volume II, Exhibit B.6, “Report on Attendance at NSOC Meeting.”
6.13
“Diablo Canyon Independent Safety Committee Twenty-Fourth Annual Report on the Safety of Diablo Canyon Nuclear Power Plant Operations, July 1, 2013—June 30, 2014,” Approved October 22, 2014, Volume II, Exhibit D.8, Section 3.8, “Meeting with NRC Senior Resident Inspector.”

For more information contact:

Diablo Canyon Independent Safety Committee
Office of the Legal Counsel
857 Cass Street, Suite D, Monterey, California 93940
Telephone: in California call 800-439-4688; outside of California call 831-647-1044
Send E-mail to: dcsafety@dcisc.org.