IR 05000346/1996007
| ML20128G465 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/03/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20128G412 | List: |
| References | |
| 50-346-96-07, 50-346-96-7, NUDOCS 9610090041 | |
| Download: ML20128G465 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Docket No:
50-346
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License No:
NPF-3 Report No:
50-346/96007(DRS)
l Licensee:
Toledo Edison Company
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i Facility:
Davis-Besse Nuclear Power Station i
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Location:
5503 N. State Route 2
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Oak Harbor, OH 43449 i'
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j Dates:
July 29 - August 27, 1996
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j Inspectors:
R. Roton, Senior Resident Inspector, Zion
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R. Langstaff, Engineering Specialist
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l Approved by:
M. Ring, Chief Lead Engineers Branch
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j 9610090041 961003
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DR ADOCK 05000346 PDR
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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Station NRC Inspection Report No. 50-346/96007(DRS)
This inspection reviewed and evaluated various aspects of your engineering activities, particularly the effectiveness of the engineering organization to perform routine and reactive site activities, including the identification and resolution of technical issues and problems.
Engineering involvement in the resolution of technical issues was timely
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and provided effective corrective actions.
Plant engineers were knowledgeable of their systems and actively
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involved in ensuring the reliability of the system for which they had responsibility.
Plant engineers were knowledgeable of the various regulatory
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requirements and programs which affected their performance.
Facility snodifications were implemented in accordance with the
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applicable installation and testing requirements.
One unresolved item was identified in the performance of 10 CFR 50.59
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licensing evaluations.
Responses to industry events were usually thorough and demonstrated a
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good questioning attitude on the part of plant engineering.
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Report Details I. Enaineerina El Conduct of Engineering El.1 General Comments (37550)
Using Inspection Procedure 37550, the inspectors reviewed and evaluated the licensee's engineering activities, particularly the effectiveness of the engineering organization to perform routine and reactive site activities, including the identification and resolution of technical issues and problems.
E2 Engineering Support of Facilities and Equipment E2.1 Acoropriateness and Timeliness of Enaineerina's Involvement in the Identification and Resolution of Technical Issues a.
Inspection Scone (37550)
The inspectors revietod a number of Potential Condition Adverse to Quality Reports (PCAQRs) to determine the effectiveness of engineering in the identification and resolution of technical issues, b.
Observations and Findinas Most of the PCAQRs documented actions which addressed the problem identified.
The inspectors noted one exception in that PCAQR 96-0753 identified that a hydrogen analyzer pump had stopped running during a test and could not be restarted.
The initial resolution only stated that maintenance personnel could not find any problerrs with the pump and that they suspected that there was a problem with the test. When the inspectors questioned the resolution, engineering personnel stated that they were aware of the problem which had occurred and that the problem was likely caused by personnel error during the test.
The inspectors considered tae weak closecut of the PCAQR to be a documentation problem rather than inadequate corrective action. The test engineer supplemented the documentation in the PCAQR to reflect engineering's considerations.
c.
Conclusions The inspectors considered the resolution of technical issues to be good.
E2.2 Evaluation of Desian Chances and Plant Modifications a.
Inspection Scope (37550)
The inspectors reviewed several design changes and plant modifications to verify conformance with the applicable installation and testing requirements.
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Observations and Findinas
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The inspectors reviewed the following modifications:
Plant Limited Modification 93-0016, Pressure Locking Resolution
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for DH11 and DH13 Plant Limited Modification 94-0019, Addressing Pressure Locking
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Concerns for DH63 and DH64
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Temporary Modification 96-10, Lift Leads to 3 Fire Detection Zones
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Leak Sealing of Valve SP12B1
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Modifications 93-0016 and 94-0014 drilled holes in the upstream discs
for gate valves in the decay heat removal system.
Except for valve l
DH64, the modifications were perf:,rmed during the May 1996 outage. The
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licensee planned to perform the modification for valve DH64 while at power.
Temporary modification 96-10 deactivated fire detectors in three fire
@tection zones.
The fire protection engineer stated that the affected fire detectors were not required for fire protection or to meet a licensing commitment. The engineer also stated that the detectors would be removed by a future modification.
The inspectors reviewed the fire hazard analysis report and did not identify a specific licensing i
commitment which required detectors in the affected fire detection zones. The fire hazard analysis report also showed that the combustible loading for the area was reasonably low. The inspectors determined that the modification was acceptable.
Valve SP12B1 was an instrument isolation valve off of a main steam line from a steam generator. The licensee documented identification of the leak from the packing area on PCAQR 96-0851, dated May 29, 1996.
The inspector verified that the repair was non-structural. The Maintenance Work Order (MWO) for the repair showed that an enclosure was required for the leak sealing.
The inspectors reviewed the calculation which demonstrated that the enclosure did not adversely affect seismic considerations and considered the calculation acceptable. The licensee had scheduled replacement of the valve for the next outage. The licensee controlled leak sealing activities by procedure DB-MM-09067,
"On-Line Leak Sealing," dated November 22, 1995, rather than a modification procedure.
The inspectors considered the licensee's actions appropriate and concluded adequate procedural guidance existed for controlling leak sealing activities.
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Conclusions The inspectors determined that facility modifications were implemented in accordance with the applicable installation and testing requirements.
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E2.3 Evaluation of Temporary Plant Modifications
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Inspection Scone (37550)
i The inspectors evaluated several ten.;:orary plant modifications to verify conformance with the applicable requirements.
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b.
Observations and Findinos The inspectors reviewed the licensee's use of data acquisition systems (DAS) on the Safety Features Actuation System (SFAS) and the Integrated Control System (ICS).
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The MWO for the DAS connected to the ICS indicated that the DAS had been
installed since 1988. The DAS was used to monitor a number of non-safety related signals including those used for main feedwater pump l
control, steam generator level, turbine bypass valves, and average
reactor temperature (Tave). No engineering review for potential impact i
on plant systems was documented for when the DAS was installed in 1988.
In response to a 1993 licensee audit finding, the licensee performed an engineering evaluation and a 10 CFR 50.59 screening. The engineering
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evaluation only addressed seismic considerations associated with the i
installation of DAS equipment. The evaluation did not address whether
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ICS could be affected from a functional standpoint due to electrical
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interactions.
The 10 CFR 50.59 screening concluded that the DAS i
equipment was not a change to the facility nor was the use of the equipment a test or experiment. Accordingly, the screening concluded
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that no formal safety evaluation was required.
The inspectors noted that if not properly isolated, the DAS could potentially affect balance
of plant systems and induce a plant transient. As such, the potential existed for increasing the probability of an initiating event, such as a
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i plant transient, and thereby increasing the probability of an accident.
No 10 CFR 50.59 safety evaluation was performed.
The MWO for the DAS connected to one channel of the containment
radiation part of SFAS showed that the DAS was in use from October 12, n
i 1995 to October 25, 1995. Control room logs for the period showed that
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bistable for the affected SFAS channel had conservatively been tripped when connecting and disconnecting the DAS unit.
However, the bista!:le i
had been reset shortly after connection and, as such, the DAS was connected to operable safety related equipment.
The licensee had
documented an engineering evaluation stating that the DAS would not
impact SFAS from either a functional or seismic ' tandpoint based on s
i engineering judgement.
No 10 CFR 50.59 safety evaluation was performed.
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The inspectors reviewed procedure NG-EN-00304, " Safety Review and Evaluation," dated May 17, 1995, with respect to tests and experiments.
The inspectors determined that the procedure was consistent with
industry guidance, NSAC-125, " Guidelines for 10 CFR 50.59 Safety Evaluations," issued June 1989, with regards to test and experiments.
However, the inspectors noted that a test or experiment which had the potential to induce a plant transient would be excluded from requiring a
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10 CFR 50.59 safety evaluation due to the licensee's definition of
" conduct a test or experiment not described in the safety analysis report." Such a test or experiment would increase the probability of an initiating event, such as a plant transient, and would thereby increase the probability of an accident previously evaluated in the safety analysis report. Although such a test or experiment would meet the definition of an "unreviewed safety question," the activity would likely not be identified as such because it did not meet the screening criteria for requiring a safety evaluation.
The inspectors questioned the lack of 10 CFR 50.59 safety evaluations for the ICS and SFAS DAS applications, and the appropriateness of the licensee's definition for conducting a test or experiment.
Information Notices 95-13 and 95-13, Supplement 1, both nott +. hat a 10 CFR 50.59 safety evaluation is required for using DAS on orrating systems. The issue of whether 10 CFR 50.59 safety evaluations were required has been referred to NRR for resolution (Unresolved Item 50-346/96007-01).
The inspectors also noted that there were few administrative controls associated with using the DAS units. Although the licensee issued guidelines in response to the 1993 audit finding, these guidelines primarily addressed whether a DAS unit was an appropriate means of data collection in a given application.
The guidelines did not require evaluation of how the equipment being monitored could be affected by the DAS unit. The inspectors considered the licensee to be relying heavily upon the experience and judgement of the individual engineer involved to ensure that operational equipment was not affected.
At the time of this inspection, the licensee was in the process of developing an engineering policy for using DAS in the field.
The licensee initiated this effort in response to the inadvertent starting of an emergency diesel generator (EDG) in May 1996 due to the connection of DAS equipment (reported to the NRC by letter dated July 17,1996).
In response to concerns raised by the inspectors, engineering management stated that they planned to implement a procedure regarding the use of DAS units which would require equipment functionality, seismic considerations, and a 10 CFR 50.59 screening to be addressed.
Installation of DAS units to equipment in service was technically adequate. However, administrative controls to ensure future installations of DAS units would be technically acceptable were lacking.
The issue of whether 10 CFR 50.59 safety evaluations were required remains under n view and is considered an unresolved item (50-346/96007-01(DRS)).
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Conclusions
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One unresolved item was identified.
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E2.4 Emeroency Diesel Generator (EGD) Rotor Replacement a.
Insoection Scope (37550)
The inspectors reviewed the licensee's resolution of degraded performance of the No.1 EDG (EDG 1).
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Observat4ons and Findinas During the eighth regular refueling outage (RF0-8), completed in April
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1993, the licensee identified that although EDG 1 reached rated speed in approximately 8 seconds, the EDG took 10.7 seconds to attain rated voltage. The EDG technical specifications require that the EDG accelerate to 900 RPM in s 10 seconds and the SFAS technical specifications require an EDG response time of 515 seconds. Although the EDG met all required technical specification acceptance criteria, plant engineering continued to troubleshoot and assess the condition of EDG 1.
(Note: EDG 2 would reach rated speed and voltage in 7.5 - 8.5 seconds.)
Through a systematic process of alimination, plant engineering determined that the cause of the EDG's performance problem centered around the larger power components, specifically the field windings on the generator and the exciter current-voltage transformers (CVTs).
Since the EDG exciter system is self-excited, plant engineers developed a troubleshooting guide to connect an external variable DC power supply to the generator field and a power resistor to the output of the exciter which modeled the resistance of the generator field. While monitoring system response, the DC supply was stepped-up at 5A increments and the results were charted and graphed. The results of this investigation indicated the existence of shorted-turns in the generator rotor. To confirm the existence of winding degradation, a generator pole-to-pole voltage drop test was completed at the beginning of RF0-10, with the intent to correct the condition during RF0-11. The results of the drop test not only confirmed the presence of shorted-turns, but due to the magnitude of voltage drops, the licensee altered the RF0-10 schedule and completed repairs to the rotor.
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Conclusions The inspectors consider the continued pursuit and resolution of the degraded condition of the EDG demonstrated a good questioning attitude and an unwillingness to accept the condition of the EDG, although the EDG met all technical specification acceptance criteria for operability.
E2.5 Evaluation of Containment Soray System Response Time a.
Inspection Scope (37550)
The inspectors reviewed the licensee reevaluation of containment spray system response time.
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Observations and Findinas In March 1996, the licensee identified that the SFAS integrated time response test (DB-SC-03114) acceptance criteria requirements for the containment spray system were not in accordance with the assumptions made in the USAR.
Specifically, USAR Section 6.2.2.3.9 states the accident analysis assumed that the containment spray pumps would cc.ne up to rated speed in approximately 39 seconds and then required an additional time delay of approximately 41 seconds during which the containment spray header was filled. However, while reviewing DB-SC-03114, plant engineering identified that the test procedure used 80 seconds as the acceptance criteria for both the containment spray pumps and overall system performance.
As a result of plant engineering's review, DB-SC-03114 was revised to reflect the USAR assumptions in the test's acceptance criteria. A review of the five most recently completed tests indicated that both pump and system performance meet the accident analysis, c.
Conclusions The inspectors considered the resolution of this issue to be appropriately conservative and reflected a good understanding of design basis on the part of plant engineering.
E2.6 Review of Peak Load Testina of EDGs a.
Inspection Scope (37550)
The inspectors reviewed the licensee's assessment of peak load testing of their EDGs.
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Observations and Findinas As follow-up to a recent (1994) NRC violation issued to - different utility, in 1995 the licensee reviewed the adequacy of ti ir surveillance testing to ensure the EDGs and the SB0DG could carry the design bases accident luads.
As a result of their review, the licensee determined that although existing surveillances met the operability requirements as specified in technical specifications, such testing would not verify the ability of the EDGs and SB0DG to carry the design bases accident loads. Although the EDGs had been factory tested to 110 percent of rated load (2,860 KW)
for two hours and the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating was 2,838 KW, there was no periodic testing conducted to ensure that they would attain a load higher then the continuous rating (2,600 KW).
The current load tables list show loading levels of 2,709.5 KW for EDG1 and 2,671.3 for EDG2 to meet a Loss of Coolant Accident coincident with a Loss of Offsite Power.
These peak loads occur during the first 25 seconds of the EDG loading sequence. As a result, the licensee changed the monthly surveillance test to verify EDG capability to handle these loads.
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l The inspectors verified EDG1 and EDG2 were tested in March 1996, during
their routine monthly surveillance tests, to a level of approximately
2,750 KW for five minutes. Additionally, the licensee is revising their semi-annual EDG surveillance test to demonstrate the ability of the EDGs
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Conclusions The inspectors considered the licensee's response to industry events to
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part of plant engineering.
E2.7 Thermocraohv Proaram
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Inspection Scope (37550)
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The inspectors reviewed and obrerved various portions of the licensees thermography program.
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Observations and Findinas
The inspector observed the performance of thermographic surveillances
and toured the plant with the plant engineer responsible for the
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licensee thermography program. Additionally, the inspector reviewed the
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results and history of various adverse conditions which had been identified and resolved through the use of thermographic techniques.
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Conclusions The inspector considered the licensee's thermographic program to be very strong. As a predictive tool, the licensee has been very successful in using thermographic techniques in the identification and correction of
deficient conditions before those conditions impact component or system
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performance.
E4 Engineering Staff Knowledge and Performance a.
Inspection Scope (37550)
The inspectors interviewed and conducted tours with various plant engineers to evaluate the relative capabilities of the engineering organization with regard to experience and understanding of responsibilities.
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Observations and Findinas The inspectors reviewed system status with several cognizant engineers.
In-plant tours and discussions with engineers indicated that overall they had good knowledge of their systems with respect to significant maintenance performed, identified problems, modifications, and application of the Maintenance Rule (10 CFR 50.65).
Based on a limited review of engineering products (i.e., PCAQR resolutions, modifications,
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and testing activities), the inspectors considered plant and design i
engineering performance to be good.
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The inspectors.onsidered the Davis-Besse engineering staff, as
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reflected by those individuals interviewed, to be kne; iedgeable and
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overall engineering performance appeared to be good, t
E7 Quality Assurance in Engineering Activities a.
Inspection Scope (37550)
The inspectors reviewed various audits and reports in order to evaluate J
the overall effectiveness of the independent safety engineering group and the quality assurance organization.
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Observations and Findinas Review of audit AR-96-DESIN-01:
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The inspectors reviewed audit AR-96-DESIN-01, dated July 12, 1996, which
focused on design engineering and configuration management. The
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inspectois noted that the audit identified issues with respect to system i
description, drawing, and USAR discrepancies.
The inspectors considered
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the level of detail described by the issues indicative of an in-depth
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review by the quality assurance (QA) organization.
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The inspectors identified that the audited organization had not provided i
a written response to four of the audit findings within 30 calendar days j
as required. Specifically, Section 17.2.18.5 of the quality assurance program description (QAPD) specifies that "a written response to each audit finding is required from the audited organization within thirty
(30) days following issuance of the audit finding, except when a shorter
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period is specified on the audit finding." The audit findings had been documented on PCAQRs using the licensee's routine corrective action program rather than using audit finding reports (AFRs). Although the timeliness of the responses did not meet QAPD requirements, the
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inspectors noted that the timeliness was commensurate with the safety significance of the items identified. The licensee performed a review of audit findings and identified that, during 1996, 10 of 16 audit findings documented on PCAQRs had not met the 30 day requirement. The licensee issued an internal memo which instructed the QA staff to document audit findiags on AFRs. The failure to provide a written response to the audit findings within 30 days is contrary to the requirements of the licensee's QAPD and 10 CFR 50, Appendix E, Criterion II, " Quality Assurance Program." This failure constitutes a violation of minor significance and is being treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforcement Policy (50-346/96007-02(DRS)).
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Review of Audit SR-96-ENGRG-02. "Ocerability of H. Recombiner Resultina
from a Review of Operatina Experience Report OE 7672:"
The inspectors reviewed audit SR-96-ENGRG-02, dated April 4, 1996. This
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audit reviewed the licensee's 1990 assessment that steam condensation in the hydrogen recombiner piping low spots could create water traps that
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might block the flow to the recombiner and affect its operability.
The engineering evaluation prepared as part of the 1990 assessment concluded that condensate would not develop as long as the recombiner was not used
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directly after the operation ci the containment spray system. However, based on a recent problem at the Oconee Nuclear Power Station, QA
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conducted a review of thi: existing assessment.
In their review, QA
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determined that several weaknesses existed in the licensee's initial
dispostioning of this concern.
For example, the 1990 assessment
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concluded that due to the temperatures of the rooms, in which the
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hydrogen recombiner piping was located, being higher than the dew point i
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of the air drawn in the recombiner from containment, condensation would
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not form in the piping. However, a field walkdown by QA located piping in areas of low temperatures which could result in the formation af condensation in the recombiner piping. A PCAQR was generated requesting
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engineering's further evaluation of this issue.
The inspectors considered this to be a good effort on the part of QA which detonstrated a good understanding of the issue and of plant design.
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Review of Audits AR-95-CORAC-0'. c !a-95-CORAC-02:
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These two audits discussed QAs review and assessment of the licensee's corrective action program. Both audits determined that the overall
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corrective action process had improved over the previous year. The audits appeared to be of sufficient scope and appeared to provided
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Conclusions The inspectors considered that quality assurance audits were of
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appropriate depth and scope.
However, one non-cited violation was identified because the licensee failed to ensure that audit findings were responded to within the timeliness requirements specified by their i
qualit.y assurance program.
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Manaaement Meetincs X1 Exit Meeting Summary
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The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on August 27,
1996. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary
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information was identified.
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j PARTIAL LIST OF PERSONS CONTACTED
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Licensee
J. Wood, Vice President - Nuclear J. Lash, Plant Manager i
T. Myers, Director - Nuclear Assurance
J. Feels, Manager - Regulatory Affairs
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K.- Tyger, Manager - Quality Assurance D. Geisen, Supervisor - E/C Systems J. Rogers, Manager - Plant Engineering N_RC S. Stasek, Senior Resident Inspector, Davis-Besse K. Sellers, Resident Inspector, Davis-Beste
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INSPECTION PROCEDURES USED
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IP 37001 10 CFR 50.59 Safety Evaluation Program, issued December 29, 1992
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j IP 37550 Engineering, issued March 17, 1995
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IP 37700 Design Changes and Modifications, issued September 17, 1990 i
Part 9900 10 CFR Guidance, 10 CFR 50.59, Changes to Facilities, Procedures
and Test (or Experiments), issued January 1, 1984
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Part 9900 10 CFR Guidance, 10 CFR 50.59, Interim Guidance on the
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Requirements Related to Changes to Facilities, Procedures and i
Tests (or Experiment), issued April 9, 1996
Part 9900 Technical Guidance, Assessing On-Line Leak Sealing of ASME Code l
Class 1 & 2, issued October 11, 1994
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LIST OF ACRONYMS USED PCAQ Potential Condition Adverse to Quality PCAQR Potential Condition Adverse to Quality Report MWO Maintenance Work Order ICS Integrated Control System SFAS Safety Features Actuation System QAPD Quality Assurance Program Description EDG Emergency Diesel Generator 13