IR 05000275/1994022
| ML16342C684 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 09/21/1994 |
| From: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML16342C683 | List: |
| References | |
| 50-275-94-22, 50-323-94-22, NUDOCS 9410030014 | |
| Download: ML16342C684 (44) | |
Text
APPENDIX U.S.
NUCLEAR'REGULATORY COMMISSION
REGION IV
Inspection Report:
50-275/94-22.
50-323/94-22 Licenses:
DPR-80 DPR-82 Licensee:
Pacific Gas and Electric Company 77 Beale Street, Room 1451 P.O.
Box 770000 San Francisco, California Facility Name:
Diablo Canyon Nuclear Power Plant, Units 1 and
Inspection "At:
Diablo Canyon Site,,San Luis Obispo County, California Inspection Conducted:
July 24 through August 31, 1994 Inspectors:
M. Miller, Senior Resident Inspector M. Tschiltz, Resident Inspector i
Approved:
Ins ection Summar irsc
,
i
,
eactor roJects rane a
e Areas Ins ected Units
and
Routine, announced, resident inspection of onsite followup of events operational safety verification, plant maintenance, surveillance observations, plant support activities, onsite engineering, quality oversight activities, followup of violations and in-office review of licensee event rep'orts.
Results Unit's 1 and
~oerations:
Strengths:
Alert control room operators noted unexpected trip bistables activated during troubleshooting of steam generator instrumentation inputs.
Prompt operator action prevented a potential unit trip (Section 5. 1).
Control room crews proactively reviewed the 'procedures and methods for mitigating a unit trip without offsite power after receiving a startup bus undervoltage condition'Section 2).
9410030014 940921 PDR ADOCK 05000275 Q
~
Nuclear Quality Services (NQS) identified that less than effective management of change was a weakness common to several operations staff problems noted in past reports.
Operations management evaluation of this concern evidenced a proactive and self-critical attitude (Section 9. 1).
Maintenance:
Strength:
~
The proactive and prompt issuance of a nonconformance report (NCR) when the conduct of a troubleshooting job was not consistent with expectations was a strength (Section 5. 1).
Weaknesses:
~
The lack of understanding by a technician and weakness in job planning during the troubleshooting of steam generator instrumentation resulted in test signals being improperly inserted into control and protection circuitry, making the unit more vulnerable to a reactor trip (Section 5.1).
~
Items were noted to be laying across surface contamination area (SCA)
boundaries after work on a job was interrupted at the end of shifts (Section 4. 1).
Strength:
~
Quality assurance audits evidenced intrusive and insightful involvement, problem identification, and willingness to elevate concerns to management when line organizations were not timely in resolving those concerns.
Additionally, quality organizations identified defensiveness in the engineering organization regarding resolution of some problems (Section 9).
Weaknesses:
Two failures by engineering organizations to properly address effects of hot leg streaming resulted in improper documentation of plant design and performance (Section 9.2),
Failure to aggressively coordinate site and corporate resources to resolve the emergency diesel generator (EDG) air flow inconsistencies resulted in inconsistencies between test and analysis data for several months (Section 6. 1).
-3-
~
Commercial grade dedication was not performed properly, resulting in a
lack of inspection to assure that all solder connections had been properly completed on printed circuit board cards (Section 6.2).
~
The system engineer walkdown had not identified containment ventilation system leakage at access port gaskets during monthly walkdowns, which indicated a need for improvement in the quality of walkdowns for this system (Section 5.2).
~
The failure of the Nuclear Regulatory Services'organization to update procedural requirements to clarify the need for a one-hour nonemergency report upon actuation of EDGs, due to startup transformer undervoltage, resulted in a failure to make a prompt report to the NRC, a violation of the
CFR 50.72 reporting requirements (Section 7),.
~P1 S
in Weakness:
Strength:
~
Emergency Planning and emergency response was strong regarding the August 1994 wildlands fire.
The self critical'ttitude of the EP or anization resulting in initiation of an NCR to record lessons learned the fire response was a strength (Section 7).
~
Housekeeping was not maintained consistent with the expectations of
'lant management, or NRC inspectors and management, and was in need of.
improvement (Section 8).
Summar of Ins ection Findin s:
Noncited Violation 323/94-22-01 was identified (Section 5. 1).
Inspector Followup Item 275/94-22-02 was.identified (Section 6.2).
Noncited Violation 323/94-22-03 was identified (Section 7).
Violations 275/94-07-01 and 323/94-01-01 were closed (Sectio'n 10).
Licensee Event Reports 275/92-27, Revision 0, 323/03, Revision 1, were closed (Section 11).
Attachments:
~
Attachment 1 - Persons Contacted and Exit Heeting
~
Attachment
DETAILS
PLANT STATUS (71707)
1.1 Unit
Unit 1 operated at 100 percent power for the entire report period with the exception of a reduction to about 93 percent power for main turbine valve testing.
1.2 Unit
Unit 2 operated at 100 percent power for the. entire report periods
ONSITE RESPONSE TO EVENTS (92701 and 93702)
F 1 Wild-Land Fire Outside of the Owner Controlled Area On August 12, a fire about 20 miles north of the site caused power failures and an undervoltage condition on'he Units
and 2 startup transformers, resulting in a start of the EDGs and partial loss of emergency notification siren capability.
The licensee made a formal report of the loss of the siren capability to the NRC.
The NRC resident inspector responded to the site.
As a prudent measure, after the loss of startup bus voltage, the control room crews performed a thorough review of the methods and procedures to be used to mitigate a plant trip without offsite power.
The fire caused faults in the electrical distribution grid, resulting in a load rejection at the Horro Bay (nonnuclear)
power plant, which is one of the four sources of offsite power to Diablo Canyon.
The Horro Bay plant load rejection caused multiple power outages in the area, and resulted in two other sources of offsite power, the Hesa and Hidway 230 KV lines, to lose voltage, thus causing the undervoltage on Units
and 2 startup transformers.
The assessment of the licensee emergency response organization's response to this event is discussed in Section 7 of this report.
The fire burned approximately 50 thousand acres of grassland and forest, and approached to within a few miles of the emergency operations facility (EOF),
within about 10 miles of the main 500 KV transmission lines, and within a few miles of Black Butte, where major communication and siren control facilities are located.
The fire was contained several miles from the Diablo Canyon site, and was extinguished by August 19.
Conclusion The fire did not at any time pose a significant threat to the safe operation of the facility.
The control room crew briefing on mitigation of a plant trip after loss of offsite power was a noteworthy strength.-
OPERATIONAL SAFETY VERIFICATION (71707)
3. 1 Ins ector Walkdowns The inspectors performed several walkdowns of the plant systems, and found the systems to be in appropriate condition with exception of houseke'eping concerns noted later in this report.
Several areas of the plant were found to have locked valves appropriately positioned, and system conditions appeared satisfactory.
The inspectors noted very few minor discrepancies other than housekeeping.
The preparations for the Unit 2 outage were underway, and materials were found to be stored appropriately, and scaffolds were found to be well constructed with respect to seismic requirements.
Conclusion The plant systems were found in good condition with respect to locked valves and system integrity, i.e,,
obvious leakage.
The outage preparations for Unit 2 were well controlled'
PLANT NAINTENANCE (62703)
During the inspection period, the inspector observed and reviewed selected documentation associated with maintenance and problem investigation activities listed below, to verify compliance with regulatory requirements, compliance with administrative and maintenance procedures, required quality assurance/quality control department involvement, proper use of safety tags, proper equipment alignment and use of jumpers, personnel qualifications, and proper retesting.
Specifically, the inspector witnessed portions of the following maintenance activities:
Unit
~
Annunciator response for high ambient temperature compensatory measures for EDG air flow (both units, multiple observations, PK 15-05, PK 15-09)
~
Inspection of battery charger printed circuit boards (both units)
Unit 2 Scaffolds associated with Thermo-Lag removal in the component cooling water (CCW) heat exchanger room Pyrocrete Installation on structural members of the component cooling water room walls Installation of Eagle 21 protection system wiring, raceways and supports in the 115 foot elevation
~
Unit 2 reactor coolant pump seal injection transmitter replacement and calibration t
~
Main Steam Valve MS-2-FCV-151 actuator yoke replacement
~
EDG 2-2 fan drive alignment check Conclusion With the exception of the observations noted below, all maintenance activities were performed according to procedures, personnel appeared very knowledgeable of requirements, and activities appeared to have been planned and performed in a knowledgeable and professional manner.
Outage times for equipment were minimized according to their safety significance, particularly the fan drive alignment check for the EDGs.
4.1 Unit 2 Reactor Coolant Pum Seal In 'ection Transmitter Re lacement and Calibration r
The inspectors and several members of NRC Region IV management toured the Radiological Control Area (RCA) after routine working hours, noting that this particular job appeared to have been temporarily secured at the end of the shift in a less than orderly manner.
The job was in progress about a week, during the day shift.
On several occasions, the inspectors noted that items in the work area (a surface contamination area (SCA)) were left at the end of the shift laying across the SCA boundary.
After informing the health physics access senior staff and maintenance management upon finding each of the repeated observations, items were still observed laying across SCA boundaries.
This work area was not brought into an orderly condition until after the conclusion of the job.
,Maintenance and health physics management clarified that items laying across SCA boundaries were not consistent with management expectations.
Conclusion Although the work on the transmitter appeared to have been accomplished in a satisfactory manner, the cleanliness of the work area and potential to spread contamination by items lying across SCA boundaries was not appropriate,
SURVEILLANCE OBSERVATIONS (61726)
Selected surveillance tests required to "be performed by the Technical Specifications (TS) were reviewed on a sampling basis to verify that:
(1) the surveillance tests were correctly included on the facility schedule; (2)
a technically adequate procedure existed for performance of the surveillance tests; (3) the surveillance tests had been performed at a frequency specified in the TS; and (4) test results satisfied acceptance criteria or were properly dispositioned.
Specifically, portions of the following surveillances v,ere observed by the inspector during this inspection period:
Unit
~
Radiation Monitor for Containment Vent (STP-I39-R44.8A)
~
Containment Penetration Isolation Valve Leak Testing (STP-V-633)
~
Emergency Signals and Communications Systems Functional Test (STP-I-29)
Unit
~
Condensate Storage Tank Level Indicator 2-1 Calibration (STP-I-16-L40)
~
Channel Calibration Steam Generator Feed Flow, Steam Flow and Steam Pressure Channels (STP-I-12B)
5. 1 STP-I-12B1 Channel Calibration of Steam Generator Feed Flow Steam Flow and Steam Pressure Channels i
L
~Back round Buring the last report period, the licensee identified that, during the July 1994 performance of the surveillance test STP-I-12Bl, the input signal for steam generator feedwater flow was not properly removed from service.
This resulted in a small change in steam generator level, which was identified by operators.
The licensee took corrective action to revise the procedure to place greater emphasis on and clarify.the proper actions to remove feedwater circuits from service during calibrations.
Current Concern On August 24, 1994, during troubleshooting which used the revised surveillance test STP-I-12Bl, to evaluate steam generator pressure and steam flow analog channels, and associated alarm, protection and safeguards comparators, only the steam generator feedwater channel was removed. from service prior to inserting simulated inputs into the circuits, although the procedure indicated that several other circuits should be removed from service Since the procedure directed that the other circuits, which had been improperly removed from service, should receive test inputs, 'reactor protection set bistables were activated.
Control room operators, aware of the ongoing testing, observed the bistables changing to the trip position, and took prompt action to stop the testing and restore systems to normal.
During an improperly performed surveillance, documented in the last report, this was the same point of confusion for the technician, who misunderstood the applicability of the step for the feed flow instrument, since the instruments which were being calibrated were steam generator pressure and feed flow.
In this case, the misunderstanding was exacerbated by the work order referring to removing the channel from service, while the tailboard discussion and discussions between the foreman and the work planner had concluded that the entire rack should be removed from service.
Many of the same individuals involved in the issue discussed in the last report were involved in this occurrence.
The Director of Instrumentation and Controls informed the NRC that the improper performance of the troubleshooting had occurred, and also concluded that an NCR should be initiated, and that
corrective action should include detailed discussions to determine the level of involvement and understanding of each person involved with the surveillance.
Conclusion Prompt operator response to unexpected bistable trip indications prevented an improperly performed surveillance from significantly impacting plant operation.
Although the procedural step which removed the instruments from service was previously revised, due to a failure to remove feed flow instrumentation from service, the surveillance was performed in a different manner.
The I&C supervisor signed the work order designating that the channel, rather than the previously agreed upon rack, be removed from service, and improper work was later performed.
The 18C director took appropriate action to initiate an NCR, inform the NRC, and to identify and address the sources of the errors which led to this improper performance of troubleshooting work.
The failure to properly plan and perform the maintenance work is a violation of TS 6.8. 1., which states, in part, that written procedures shall be established, implemented, and maintained covering applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978'.
Appendix A of Regulatory Guide 1.3, Revision 2, recommends procedure covering surveillance testing of safety-related systems, Contrary to this requirement, on August 26, 1994, several instruments associated with Unit 2 steam generator instrumentation were not removed from service prior to testing as required by Surveillance Test STP-I-12B1.
Since this violation was identified by the licensee, and other criteria of Section VII.B(2) of the Enforcement Policy were satisfied, this violation was not cited (323/94-22-01).
5.2 Containment Penetration Isolation Valve Leak Testin STP-V-633 The inspectors observed the testing of the containment penetration valve leakage following valve operation during a containment venting evolution.
The engineers performing the test appeared very knowledgeable of the requirements of the test, the methodology of test performance, and the procedures and theory involved in the operation of the test equipment.
The surveillance procedure was properly followed, signed step-by-step as required, and proper time intervals were followed for measuring equipment to reach equilibrium.
The measuring equipment calibration dates had been checked, and were within calibration intervals.
While in the area of the test, the inspectors walked down portions of the containment ventilation system.
The system in the area of the containment vent penetration appeared to be in appropriate condition with respect to valve positions, seismic restraints, and cleanliness.
The inspectors noted that, on the containment vent line, two gasketed access ports downstream of the second isolation valve evidenced significant air leakage.
The inspectors noted this to the licensee, and pointed out that system engineer walkdowns should identify these types of concerns.
Safet Si nificance The licensee identified that these ports had leaked several years ago, and that a safety analysis had determined that this leakage
was insignificant, since it was downstream of the containment isolation valves.
This appeared to be an appropriate conclusion.
Conclusion The surveillance tests were performed well.
The leakage from the containment ventilation line was not significant.
However, the system engineer walkdown had not identified this leakage during monthly walkdowns, which indicated a need for improvement in the quality of walkdowns for this
.
system.
5.3 Condensate Stora e Tank Level Indicator 2-1 STP-I-16-L40 On August 26, 1994, the inspector noted that an electrical meter (Fluke)
was present under one of the Unit 2 control room consoles, with an electrical line attached to one of the emergency response facility data system (ERFDS)
multiplexer cards.
The meter was not listed on jumper logs, did not have a
tag identifying its purpose, and, according to the control room operators, had been present in the control room since the crew turnover that morning at 7 a.m.,
presumably since the afternoon before.
The inspector noted that the meter's calibration was not expired.
Operators were in the process of determining the purpose of the meter.
Since the time was close to 11 a.m.,
the inspector expressed concern, to both the shift foreman and the maintenance manager, that sensitivity to seismic requirements and control room ownership may be less than appropriate.
Within 15 minutes, the control room determined that the meter was in use in performing STP-I-16-L40, which had carried over from the previous day.
Haintenance management and control room staff expectations require that measuring equipment be removed.from safety-related equipment if a surveillance is secured until the next day.
The meter was promptly removed until the surveillance was continued at a later time.
Safet Si nificance The presence of the meter was not a concern to safety-related equipment since the area of the meter included only emergency response facility data system circuitry, with extensive seismic supports.
The circuit to which the meter was attached was an indication circuit only.
Conclusion The sensitivity of control room staff and maintenance staff was less than optimum regarding measuring equipment being left in the area of seismically qualified equipment.
ONSITE ENGINEERING (37551)
6. 1 Insufficient Radiator Air Flow to EDGs
~Back round As discussed in an earlier report, the licensee identified that the technique to measure radiator air flow had been revised by the industry, which resulted in application of a significant penalty to compensate for uncertainty in EDG radiator air flow measurement.
The licensee operability evaluation concluded that the EDGs were currently operable based on compensatory measures involving opening of doors to provide additional air flow if ambient air temperatures rose above 71'F, and implementation of security compensatory measure Recent Information During this report period, post calibration data was obtained for the air flow measuring equipment, which indicated that the error band was significantly wider than earlier anticipated.
The licensee promptly made a one-hour non-emergency report to the NRC stating-that the air flow
.
through the radiators may be significantly less than expected, and that the compensatory measures still were considered adequate to provide design basis cooling.
Com ensator Measure Concerns The NRC questioned if the licensee had considered the effect of the compensatory measures (opening doors and stationing security guards at the doors during high ambient temperatures)
on the security during a design basis security event or a design basis reactor event resulting in a release on site.
The licensee stated that, although an analysis was not documented, both a design basis security threat response and design basis potential exposure to personnel bounded these effects on the licensee personnel performing compensatory measures for EDG operability during high ambient temperature.
Ins ector Concerns The inspector identified that the coordination between onsite engineering and corporate engineering did not appear to have been optimum, since the air flow calculations performed in the general office had been done with minimal involvement of onsite engineering, and with minimal use of past EDG performance data.
The engineering organizations were scheduled to meet
'and resolve apparent contradictions between air flow data and other performance indications on September 2,
1994.
Several contradictions in the analysis were resolved during this meeting.
Sensitivit to Wind S eed and Direction The inspector also identified that the EDGs appeared to be sensitive to both wind speed and wind direction in addition to ambient air temperature.
Preliminary licensee calculations had identified this potential, but compensatory measures were conservatively based only on ambient air temperature, and appeared to the inspector to conservatively bound the wind concern, but had not been formally addressed by the licensee.
Conclusion The concerns appeared bounded by the compensatory actions, and therefore, were not an urgent safety issue.
However, the effective characterization of the EDG air flow performance has not been accomplished in an optimum manner, since effective coordination of engineering resources associated -with the EDGs has not occurr'ed, and some analysis efforts resulted in contradictory conclusions.
The licensee's compensatory actions appear to have ensured adequate cooling to the EDGs.
Further NRC review of this issue will be concluded during the review of the Licens'ee Event Report (LER),
6.2 Commercial Grade Procurement Concerns 6.2. 1 Replacement of Insulator in 4160 V / 480 V St;pd )wn Transformer
~Back round During the last Unit l outage, the licensee identified that cracks may occur in the ceramic insulators which secure the 480 V bus bars in the
(I safety-related 4160 V / 480 V stepdown transformers.
The insulators were inspected and a cracked insulator was identified and replaced with a glass-epoxy insulator of similar configuration.
A calculation was performed which concluded that, if cracks were to occur in the remaining ceramic insulators, the seismic forces would not result in failure of the insulators to perform their safety-related function.
Review of the Calculation The inspector reviewed the calculation.
The calculation used data from an earlier shaker table test as input to a finite element analysis to determine stress on the insulators.
After discussion with the engineers, the inspector concluded that the assumptions of forces acting on the insulators appeared reasonable, and bounded by the worst case expected
,acceleration.
Commercial Grade Dedication Concern The inspector identified that the structural characteristics of the replacement (epoxy-glass)
insulator such as tensile, compressive strength, and creep, as well as electrical characteristics of the insulator, had been verified by the licensee only at room temperature, and had not been verified at operating temperature.
The licensee stated that the characteristics had been verified at 130'elsius (C)
by a laboratory.
The licensee accepted the conclusions of the vendor that these characteristics were considered by the vendor to be valid up to 140 C,
and additionally took as assurance that these insulators were in widespread use in the electrical industry.
The licensee had not performed an evaluation of the validity of the vendor's assertions.
En ineerin Anal sis Concern The inspector identified that, since the Westinghouse documented design hot spot in the transformer is a temperature of 150'C, and the vendor's maximum recommended service temperature of the epoxy-glass insulator is 140'C, the insulator material may not be suitable for the application.
The licensee initiated a prompt operability assessment (POA),
and plans to resolve the concern by either removing the insulator, or providing an engineering analysis.
The initial assessment concludes that the design hot spot is located in the 4160 V transformer core, a distance from the insulator.
The convection cooling in the transformer cabinet would be expected to reduce the temperature near the insulator to a temperature significantly lower than 150'C.
Conclusion The inspector concluded that, since the insulator material was a
commercial grade dedication item, and the analysis required specific minimum structural and electrical characteristics at operating temperature, acceptance of a commercial grade vendor's operating temperature test data may not be adequate for items installed in safety-related applications.
This concern, as well as the following commercial grade dedication issue in the next section, will be examined further upon completion of the licensee's evaluation (50-275/94-22-03).
Regarding the glass-epoxy insulator maximum temperature being less than the analyzed high temperature in the transformer, the licensee failed to formally address the potential concern during the engineering analysis.
The licensee
-12-has since initiated appropriate corrective actions, and the failure to resolve and document the resolution of the temperature environment concern appears to be of minimal safety significance.
6.2.2 Printed Circuit Boards in Battery Chargers Hissing Solder at Connection Points During the Unit 1 Outage in Hay 1994, the licensee noted that the safety-related 1-1 battery charger voltage was higher than expected.
Investigation did not identify the source of the problem, and the problem later ceased; Later, in July 1994, a higher voltage again occurred, and investigation identified that one of the connections on a printed circuit board in the battery charger was missing solder.
The wire attached to that connection pad on the circuit board was not making proper electrical contact, causing higher voltage.
A prompt operability assessment was written, determining that solder is required for seismic qualification of the batt ry charger, and that if a second case of missing solder was identified in safety-related battery chargers, a formal report to the NRC should be made.
The licensee initiated a
schedule to inspect all warehouse stock, all spare battery chargers, and all in-service battery chargers.
Inspections identified two-other cases of missing solder in warehouse stock, and one additional case of missing solder in safety-related service.
The defective boards were replaced.
The condition was promptly reported as a 4-hour, nonemergency report, in accordance with
CFR 50.72.
The licensee is continuing inspections, and will evaluate the safety significance of the as-found conditions upon conclusion of the inspections, approximately mid-September 1994.
Commercial Grade Dedication Activities Since the circuit boards are procured as commercial grade items, the licensee performs dedication of the items to safety-related service.
The dedication activities had not required verification that solder was present on solder circuit connection pads.
The licensee initiated steps to inspect solder pads in future commercial grade dedication activities, and to identify other printed circuit boards being procured by commercial grade dedication.
Conclusion The two examples of less than adequate commercial grade dedication activities (glass-epoxy insulator and circuit board solder connections)
are still under evaluation by the licensee and the inspectors.
These commercial grade dedication issues will be examined further during a future inspection (50-275/94-22-02).
EMERGENCY RESPONSE (71750)
The license response to the wildland fire was observed-by the resident inspectors to determine if the response was appropriate, and if the licensee considered the effects of the fire relative to the ability of the emergency response organization to respond to events.
Effectiveness and Involvement of the Emer enc Res onse Or anization The licensee manned the EOF soon after the fire caused a loss of electrical power
-13-for most of San Luis Obispo county.
Although the Diablo Canyon site was not threatened, the fire approached within a few miles of the EOF, the siren and pager communication transmitter on Black Butte, and the offsite power transmission lines.
The licensee's emergency response organization coordinated closely with the county emergency staff, and with the NRC, to ensure flow of accurate information concerning the fire and associated damage and identification of critical equipment to be protected from the fire.
Availabilit of the Emer enc Res onse Or anization When the transmitter on Tassajara Peak was destroyed by fire, the licensee concluded that about one-fifth of the emergency response staff would not be available by pager, although telephone service was still available to the affected area.
The licensee estimated that the remaining four-fifths of the emergency responders would be sufficient to respond to a design basis event at the site.
After questions by the NRC inspectors, the licensee concluded a formal evaluation about 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> later that the loss of. the transmitter would not significantly affect response to a design basis event.
Overall Effectiveness of Emer enc Res onse Or anization The licensee manned the EOF and continued to be involved with the NRC, county and state to provide assessments and updates of the fire damage and fire fighting priorities.
Throughout the observations by the inspectors, only a few minor coordination delays occurred.
The emergency -response organization, however, issued an NCR to accumulate and track lessons learned from the fire response.
Re ort of Unantici ated EDG Start In the area of reportability of engineered safety feature (ESF) actuations, the licensee did not'report the start of the six EDGs upon startup transformer undervoltage until several hours later.
The procedure which provided guidance concerning the reporting requirements specifically addressed the condition of an EDG start under these circumstances as not reportable.
The shift supervisor followed this guidance.
Licensee Mana ement Involvement and Corrective Actions After later review by licensee management, the licensee reported the start of the six EDGs as an unanticipated ESF actuation, and as a late 4-hour nonemergency actuation.
Earlier communication with the NRC had clarified that this type of unanticipated start of EDGs should be repor'ted as an ESF actuation.
The licensee had initiated actions to change the procedure to require a report, but had not yet updated the procedure.
After the event, the licensee updated the procedure to include the requirement to formally report this type of EDG start to the NRC.
Conclusion Regarding the effectiveness of the emergency response organization, the inspectors concluded that the response to the fire had been proactive and effective, and that the licensee had shown strong cooperation and coordination with the county.
Additionally, the licensee evidenced a
strong self-critical approach by initiating an NCR to identify improvements and track lessons learned from the response to the fir )
-14-Regarding the failure to promptly report the EDG starts, the inspectors concluded that the lack of a prompt report of the EDGs was a violation of the requirements of 10 CFR 50.72, which requires that inadvertent ESF actuations be reported to the NRC within one hour.
Since the violation is of very low safety significance, and since the inspectops have verified the completion of corrective actions, in accordance with Section VII.B of the Enforcement Policy, this violation was not cited (275/94-22-03).
PLANT SUPPORT HOUSEKEEPING (71750)
During several tours of all areas of the power block, the inspectors and several NRC visitors from the Region IV office observed that plant housekeeping had declined and was not in keeping with typical levels in reactor plant equipment areas.
Several observations of dirt, trash, radiation protection gloves and bags of clothing were noted.
The licensee has since initiated efforts to identify and correct the level of cleanliness in the plant.
Conclusion The licensee's housekeeping declined to a lower level than previously maintained.
The observations by the NRC were not safety significant concerns.
NUCLEAR OUALITY OVERSIGHT ACTIVITIES (40500)
The inspector evaluated the activities and effectiveness of several of the nuclear quality oversight groups.
The inspector evaluated each audit to determine if appropriate planning and preparation for the audit was apparent, if performance based findings had been identified, and if,the audit had been directed toward areas of activity which were safety significant.
The documentation of findings on quality problem identification documents was verified, and timeliness of resolution of the problems was evaluated for a sample of audit findings.
9. 1 ualit Performance Assessment Re ort PAR Second uarter 1994 The (PAR was issued quarterly by NgS and evaluated each area of plant management.
The assessment identified several well directed evaluations of licensee organizations.
Evaluations and conclusions appeared to have been substantiated with several observations.
Among other observations, the SPAR noted that several of the personal performance problems experienced in Operations Services had a common element of a change in processes.
The SPAR recommended that the Operations Services management closely evaluate change management to reduce the instances of personnel error.
Operations Services management accepted this observation and has initiated efforts to'etter manage changes controlled by the operations staff.
The (PAR also documented that Nuclear Engineering Services had evidenced a need to improve in the area of procedural compliance and recommended a continued e apnasis to establish a
cultural mind-set that fosters problem accountability and resolution as opposed to a defensive response to identified problem Conclusion The inspector concluded that the SPAR was a. credible effort and provided valuable insights to effect improvement.
NgS identified a weakness common to several operations staff problems not'ed in past reports, i.e., less than effective management of change.
Operations management evaluation of this concern evidenced a proactive and self-critical attitude.
9.2 Audits Performed b
N S
The inspector reviewed several audits performed by the NgS.
Audits reviewed by the inspector included the following:
Engineering Self-Assessment (Preliminary Briefing)
equality Operational equality Assessment (Audit 940091)
'R6 SSOMI - Installation Assessment (Audit 940151)
Solid Radwaste Management Program, Including the Process Control Program (Audit 94021I)
Radiation Protection Audit, Humboldt Bay Power Plant (Audit 940281)
Emergency Plan and Implementing Procedures (Audit 940221)
Annual Radiation Protection Audit (Audit 940171)
Solid Radwaste Management Program, Including the Process Control Program (Audit 94021I)
Generic Letter 89-10 Program Audit (Audit 940161)
1R6 Maintenance guality Assessment (Audit 94010I)
1R6 SSOMI - Design Assessment (Audit 930471)
En ineerin Issues Identified b N
S Audits Two separate audits identified a
weakness in the engineering organizations.
The first found that the Unit
hot leg streaming effects, caused by moving the Loop 2 T(hot) resistance temperature detectors (RTDs) closer to the core, were not adequately analyzed or documented by the corporate engineering organization in the design change analysis.
The second weakness involved the effects caused by the closer RTD location, which were later not appropriately dispositioned by the site reactor engineers during power range testing, when deviations between RTDs in Loop
were higher than the acceptance criteria.
In both cases, quality assurance auditors noted inappropriate resolution of the concern, as well as resistance to address the quality assurance issues by the respective engineering organizations.
The inspector considered this to have been a series failure of the line organizations, since the RTDs indicated higher deviations (1.2'F)
than the acceptance criteria (1.0'F),
as a result of being closer to the core than the generic Westinghouse analysis recommended.
The licensee stated that
-16-revision of the Unit 1 safety analysis would be pursued, and that there was minimal safety significance in this deviation since about 4'F margin existed between T hot and reactor trip setpoints.
The inspector questioned whether proper analysis would be performed for the similar Unit 2 design change, which the licensee agreed would be accomplished.
~Summar In addition to the two audit findings discussed above, each of the other audits reviewed were applied to safety related areas, and appeared to have been well directed to identify problems.
The problems identified were insightful, and findings were performance based.
Each problem identified during an audit was documented on a formal problem identification document, and resolved in accordance with problem resolution schedules.
The audit of the Eagle 21 installation and the audit of the 1R6 Design Assessment included several insightful findings which would not have been possible without effective leadership of the audit team and highly knowledgeable auditors.
Most problems identified during audits were promptly addressed by the line organization.
However, the inspector noted that, fo, some engineering problems, the licensee's line organization continued to dispute the significance or validity of findings.
Therefore, the quality organization continu'ed to elevate the problem to higher levels of management for resolution, as required by N(S procedures and policies.
Examples of these were concerns associated with hot leg streaming, weaknesses in recent changes to the implementation of the design control program, and fire protection circuit separation.
Safet Si nificance The findings of NgS were of varying significance.
They have been or are being resolved consistent with their safety. significance.
None of the findings have been dropped, and all have been elevated to higher levels of management for situations where the line organization has not provided a satisfactory or timely resolution.
Conclusion The audits reviewed were of significant strength, and identified several noteworthy Findings.
Many problems identified by audits have been promptly resolved.
The NRC will continue to follow the licensee's resolution of issues which are elevated to management based on lack of timely resolution within the line organization.
Regarding the series failure by the engineering organizations to properly disposition the effects associated with hot leg streaming, the issues have been elevated to management levels by the N(S.
Since the safety significance is low, the inspector will follow the quality organization's resolution of the issue.
9.3 Inde endent Safet En ineerin Grou ISEG The inspector reviewed minutes of the monthly ISEG meeting,. and found them to be intrusive in several areas of plant safety activities.
Areas evaluated included the effectiveness of maintenance investigations and operability evaluations, foreign material exclusion, operations procedure compliance, outage safety processes, EDG air flow, and design basis performance evaluations.
The evaluations included identification of specific and
-17-programmatic concerns, as well as root cause evaluations and evaluations of corrective actions.
The evaluation recommended areas of increased focus and management action.
Conclusion The ISEG involvement in plant processes was probing, and provided an excellent snapshot of issues for which ISEG involvement resulted in plant safety improvements.
The ISEG was effective in identifying problems, particularly in areas requiring a broad perspective.
The NRC will continue to follow licensee management involvement in ISEG identified issues.
9.4 ualit Assurance Surveillance A quality assurance surveillance is an audit of abbreviated scope to quickly assess an area, and which may result in additional resources being applied to that area of inspection.
The inspector reviewed each of the following surveillances by NQS.
DCPP-Plant Material Condition/Field Observations (Surveillance Report SQA-94-0058)
Status of FME Action Requests Written During 1R5, 2R5, and 1R6 (Surveillance Report SQA-94-0063)
Identification of Plant Problems (Surveillance Report SQA-94-0062)
Quality Performance and Assessment (Surveillance Report SQA-94-0040)
Timeliness of Oil Addition to Plant Equipment (SQA Surveillance Report SQA-94-0060)
'The surveillances appeared appropriate in scope and were performance based.
The auditors were prepared and identified performance based findings.
The auditors documented the findings, and recommended followup actions where necessary.
Conclusion The surveillance program is a flexible tool to quickly assess areas not traditionally required for full scale audits.
These surveillances have provided valuable assessments in a timely manner, and are a quality program strength.
9.5 ualit Control C
Activities The inspector reviewed several instances involving QC inspector involvement in performance of plant work.
In all cases, the QC inspector understood the requirements of the procedure step involving the QC hold point (such as the measurement of clearances),
the inspector verified the item to be inspected, and placed the QC inspection stamp properly on the procedure line documenting the observation.
The QC personnel understood -the requirements to b'e verified, and personally verified these requirement '
-18-
, Conclusion The gC inspectors observed appeared effective and knowledgeable in their area of expertise.
The (jC inspectors appeared to have an excellent understanding of their function regarding independent verification.
9.6 Nonconformance Re orts NCRs The inspector reviewed a sample of ten NCRs.
In all cases, the reports addressed the problem and, For cases where the root cause was not clearly
=
evident, multiple corrective actions were implemented.
Some of the NCRs were not yet closed; however, many corrective actions had already been implemented.
'he inspector noted a wide variation in timeliness of resolution of the NCRs.
In each case, the problem had not recurred before the report was closed.
Management involvement in the more significant NCRs was evident.
Conclusion The NCR process is,effective, flexible, and appears to address problems appropriately, although evidencing wide variation in timeliness of corrective action.
The inspectors concluded that the nuclear oversight organizations were providing substantial contributions to the improved effectiveness of activities at Diablo Canyon and plant safety.
FOLLOWUP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702)
10.1 Closed Violation 50-323 94-01-01:
Stora e of Combustible Materials in an Ina ro riate Area The inspector identified that a combustible storage permit had been authorized for an area which was designated as a no combustible storage area.
The "no combustibles" sign had been located near a handrail which led personnel to believe that the area of concern was within the handrail rather than the entire room.
The licensee responded to the violation by retracting the combustible storage permit, removing the combustibles, initiating an NCR (NCR DC0-94-SS-N004), clarifying the ambiguous
"no combustibles" label and painting stripes on the floor for clarification, and further inspecting areas requiring no combustible storage.
The corrective actions included briefing personnel concerning combustible storage rules, and clearly labeling areas in which no combustibles were allowed.
The actions by the licensee appeared to have been appropriate.
10.2 Closed Violation 50-275 94-07-01:
Inadvertent Level Chan e of the Reactor Coolant S stem RCS Durin Draindown The Operations staff did not properly follow procedure requirements for draining the RCS at the beginning of the last Unit 1 outage'his failure to follow procedures resulted in the equalizing valves between the reactor vessel head and loops not being opened until well into the d~'ai down.
When the valves were opened, the level between the RCS loops and 'ihe reactor vessel head equalized, resulting in an indicated level change in the loops of over 6 feet.
The licensee initiated an NCR, reviewed the occurrences leading up
-19-the level change, revised the procedure to clarify the need to open the
'qualizing valves before continuing with draindown, and has discussed this instance with the control room crews and other licensee staff to emphasize the need to fully understand the effect of various valve manipulations.
Several other operations procedures were reviewed to determine if similar clarification was required.
The licensee's actions appeared to have been appropriate and properly implemented.
IN-OFFICE REVIEW OF LICENSEE EVENT REPORTS (LER)
(90712)
The following LERs were closed based on in-office review:
~
275/92-27, Revision 0:
Containment Vent Isolation TS 3.3.7 Not Met
~
323/93-03, Revision 1:
Nonconservative Low Temperature Over Pressure Protection Setpoint Analysis
0'
PERSONS CONTACTED 1. 1 Licensee Personnel ATTACHMENT 1 G.
H.
Nucl
- W. H.
- R. P.
T.
L.
J.
S.
- G. H.
- S G
- W. G.
S.
R.
- B. W.
J.
D.
T. L.
- J J
- C. R.
C.
D.
J.
A.
- R.
W.
- J.
R.
- K. A.
J.
C.
H.
E.
J. J.
- D. B.
- H. D.
P.
T.
S.
R.
B.
H.
- D P
D. R:
D. A.
- D. H.
1.2 NRC Personnel Rueger, Senior Vice President and General Manager, ear Power Generation Business Unit Fujimoto, Vice President and Plant Manager, Diablo Canyon Operations Powers, Manager, Nuclear guality Services Grebel, Supervisor, Regulatory Compliance Bard, Director, Mechanical Maintenance Burgess, Director, Systems Engineering Chesnut, Reactor Engineer Supervisor Crockett, Manager, Technical and Support Services Fridley, Director,'perations Giffin, Manager, Maintenance Services Grammer, Engineer, Systems Engineering Grebel, Supervisor, Regulatory Compliance Griffin, Engineer, Onsite Nuclear Engineering Services Groff, Director, Plant Engineering Harbor,'ngineer, Systems Engineering Hays, Director, Onsite guality Control Hess, Assistant Director, Onsite Nuclear Engineering Services Hinds, Director, Nuclear Safety Engin'eering Hubbard, Engineer, Regulatory Compliance Kelly, Mechanical Group Leader, Nuclear Engineering Services Leppke, Assistant Manager, Technical Services HcCann, General Foreman, Instrument Maintenance Hiklush, Manager, Operations Services Nowlen, Director, Instrumentation and Controls Nugent, Engineer, Regulatory Compliance Ortore, Director, Electrical Maintenance Patton, Director, Technical and Support Services Sisk, Senior Engineer, Regulatory Compliance Stermer, Engineer, Systems Engineering Taggart, Director, Onsite guality Assurance Oatley,'irector, Materials Services
- M. Miller, Senior Resident Inspector H. Tschiltz, Resident Inspector
- Denotes those attending the exit meeting September 2,
1994:
In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
EXIT MEETING An exit meeting was conducted on September 2,
1994.
During this meeting, the inspectors reviewed the scope and findings of the report.
The licensee
acknowledged the inspection findings documented in this report.
The licensee did not identify as proprietary any information provided to, or reviewed by, the inspector ~
'
'
,CC CCW CVCS DDE DFWCS EDG EP EOF ERFDS ESF FCV FHB KV LCO LER LTOP M8(THE MSSV NCR PCV PGA POA PORV IQC RF RPS RCP RHR RWP SCA SFP SI SSE SSER SSFAR SSPS TS UFSAR ATTACHMENT 2 ACRONYMS
auxi 1 i ary feedwater Action Request centrifug'al charging (high head injection)
component cooling water chemical and volume control system double design earthquake digital feedwater control system emergency diesel generator emergency preparedness emergency offsite facility emergency response facility data system engineered safety feature flow control valve fuel handling building kilovolt Limiting Condition for Operation Licensee Event Report low temperature over pressure protection system measuring and testing equipment main steam safety v'alve Non-conformance Report pressure control valve peak ground acceleration prompt operability assessment power operated relief valve quality control radio frequency reactor protection system reactor coolant pump residual heat removal radiation work permit surface contamination area
. spent fuel pool safety injection safe shutdown earthquake Supplemental Safety Evaluation Report safety system functional audit and review solid state protection system Technical Specification Updated Final Safety Analysis Report
~ F