IR 05000395/1999002
| ML20206P530 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/26/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20206P507 | List: |
| References | |
| 50-395-99-02, 50-395-99-2, NUDOCS 9905180252 | |
| Download: ML20206P530 (21) | |
Text
.
U. S. NUCLEAR REGULATORY COMMISSION REGION ll Docket No.:
50-395 License No.:
NPF-12 Report No.:
50-395/99-02 Licensee:
South Carolina Electric & Gas (SCE&G)
Facility:
Virgil C. Summer Nuclear Station Location:
P. O. Box 88 Jenkinsville, SC 29065 Dates:
February 14 - March 27,1999 Inspectors:
M. Widmann, Senior Resident inspector M. King, Resident inspector (In-Training)
G. Hutto, Robinson Site Resident inspector (In-Training) (Section O2.1)
W. Bearden, Reactor inspector, Ril (Sections M1.2, M2.1, M2.2, and E8.5)
P. Fillion, Reactor Inspector, Ril (Sections E1.1, E1.2, E1.3, E8.1, E8.2, E8.3, and E8.4)
Approved by:
R. C. Haag, Chief, Reactor Projects Branch 5 Division of Reactor Projects t
9905190252 990426 i
PDR ADOCK 05000395 S
O PM
,
,
ENCLOSURE
- _ _ _ _ _ _ _ _ _
r-
'
.
,
EXECUTIVE SUMMARY Virgil C. Summer Nuclear Station NRC Integrated inspection Report No. 50-395/99-02 This integrated inspection included aspects of licensee operations, maintenance, engineering, and p! ant support. The report covers a six-week period of resident inspection; in addition, it includes the results of announced inspections by regional reactor inspectors in the areas of maintenance and engineering.
Ooerations Operations personnel in the control room generally exhibited a high level of
professionalism with the exception of two isolated instances on backshifts which were discussed with management (Section 01.2).
Based on a detailed walkdown the Emergency Diesel Generators (EDGs) were found to
be properly aligned and in a standby condition per licensee procedures. Technical specification requirements for fuel oil and surveillance requirements were being met.
Several small EDG lube oilleaks were observed. The maintenance rule program properly monitored EDG performance (Section O2.1).
The licensed operator simulator requalification examination scenarios were challenging
and operators' performance met test objectives. Examination critiques were thorough and provided a comprehensive assessment of individual and crew performance (Section O5.1).
A Management Review Board (MRB) held to review the plant transient of January 3,
=
1999, provided valuable insights into the contributing factors and circumstances surrounding the event. Both the inspectors and the MRB recognized the need to better understand the circumstances surrounding and contributing factors to this event in a more timely manner (Section O7.1).
Maintenance Nine completed surveillance test and preventive maintenance packages demonstrated
acceptable test results for emergency core cooling system relief valves and check valves (Section M1.2).
_The A emergency diesel generator operability, slave relay and support system leak
surveillance tests were performed in accordance with established procedures and-demonstrated operability of the equipment in accordance with the Technical Specification surveillance requirements. Personnel conducting the tests demonstrated a good level of knowledge. The pre-job briefing was thorough (Section M1.3).
Review of leakage testing data indicated acceptable material condition for Reactor
Coolant System (RCS) isolation boundaries. No examples of inadequate maintenance were identified during this review. No problems were identified during the review of equipment history which would indicate an adverse trend or degradation of the material condition of RCS pressure isolation valves (Section M2.1).
N
_ _ _ _. _ _ _ _ _ _ -. _ _ _ _ _ _ _ _
[
l l-i
{
-
.
No problems were identified with setpoint testing of ASME Class 2 and 3 relief valves in
the chemical and volume control, safety injection, and residual heat removal systems (Section M2.2).
Corrective maintenance and corrective actions have been ineffective in preventing an
increased unavailability time for the meteorological tower during the last part of 1998 and 1999. The licensee had not established a system to actively track availability time to ensure that the Final Safety Analysis Report annual target of 90% data recovery is achieved (Section M2.3).
l Enaineerina Both trains of safety-related batteries have exhibited the early stages of post seal a
leakage. The licensee made a conservative decision to replace these batteries in the
1999 refuel outage. One non-safety-related battery is approaching end of usefullife and will also be replaced (Section E1.1).
The licensee has initiated a design basis document (DBD) improvement project to be
completed over a five year period. The licensee plans to prioritize reworking / replacing / initiating calculations, technical reports, etc., using maintenance rule program risk rankings. The emergency feedwater and component cooling water system DBDs were " improved" as trial examples to help define the detailed plan and illustrate
,
the need for the project (Section E1.2).
The licensee's program for refurbishment of 7.2 kV circuit breakers is being
aggressively implemented, and should help preclude failures similar to the circulating water pump breaker failure (Section E1.3).
Plant Support A non-cited violation was identified concerning failure to properly control an escorted
visitor in the protected area. A contributing factor was an informal turnover of escort responsibilities prior to the occurrence (Section S1.1).
I
!
l
{
.
.
f l
Report Details i
l Summarv of Plant Status l
l The inspection period began with the unit operating at full power. On March 17, the licensee commenced a planned end-of-life coast down in pre.paration for refueling outage RF-11
- scheduled to commence April 2,1999. At the end of the inspection period the unit was at
!
approximately 92 percent power.
l. Operations l
Conduct of Operations 01.1 General Comments (71707)
The inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious. Specific events and noteworthy observations are detailed in the sections below.
01.2 Control Room Conduct a.
Insoection Scope (71707)
Inspectors observed frequent control room activities during the inspection period to assess operator performance.
I b.
Observations and Findinas As part of normal observation of control room activities the inspectors generally noted a high standard of professionalism and that the area was free from congestion and disturbances. Overall, the operators were appropriately focused on reactivity and important plant parameters. However, the inspectors did note during the inspection period two separate back-shift observations where control room conduct was not maintained at this high level of performance. These observations were discussed with plant management.
c.
Conclusions Operations personnel in the control room generally exhibited a high level of professionalism with the exception of two isolated instances on backshifts which were discussed with management.
Operational Status of Facilities and Equipment O2.1 Enaineered Safetv Feature (ESP) System Walkdown - Emeraency Diesel Generator (EDG)
l a.
Inspection Scope (71707)
The inspectors performed a detailed walkdown of the A and B EDGs and portions of the support systems including service water, air start, and fuel oil systems.
n
'
,
L b.
Observations and Findinas During the period of February 22-25, the inspectors performed a detailed walkdown of the EDGs and their support systems. The EDGs are risk significant and in the scope of 10 CFR 50.65, the Maintenance Rule. Overall, the inspectors determined that the EDGs and their support systems were operable and available and in a standby readiness state to' enable the EDGs to perform their intended safety function.
As part of this detailed ESF system review the inspectors verified that system valves for
- the service water, air start and fuel oil supply systems were in their required positions and properly labeled consistent with system drawings. Electrical breakers for the fuel oil l
transfer pumps and diesel generator output were verified in their proper position.
l Generator relay indications and local control panel settings were verified to be correct.
'
Starting air pressure, tube oil temperature and jacket water temperature were verified to be in their standby ranges. The EDG governor settings as well as governor oil level were verified consistent with system requirements. Control room switch positions were
!
verified to be consistent with established procedures. In addition, the correct revision of the annunciator response and emergency operating procedures were available in the control room and at the local control panel. Fire protection systems were checked, including fire extinguishers for current inspection certifications and charge pressure. No
fire protection system issues were identified.
The inspectors discussed with the system engineer various aspects of EDG maintenance rule trending and system monitoring and reviewed the EDG vendor manual and system engineer files as part of the maintenance rule review. At the time of the review, the EDG system was in 10 CFR 50.65 Maintenance Rule category a(1) due to the a(2) performance criteria for EDG unavailcbility and reliability being exceeded due to
!
problems with EDG governor controls.- The inspectors reviewed the scoping, risk classification, performance criteria, classification and recording of failures, the a(1)
evaluation, corrective action and goals monitoring of the EDGs performed under the maintenance rule program. The inspectors concluded that the EDG performance,
including the failures which occurred in November and December 1997, were being
'
properly monitored under the maintenance rule program. The licensee plans to place the EDG system into a(2) category following issuance of the March 1999 Monthly System Status Report if continued good availability and reliability are maintained.
The inspectors reviewed data from several completed surveillance test procedures (STP)-125.002, " Diesel Generator Operability Test," Revision 18 and observed a monthly A EDG operability test which are documented in Section M1.3 of this report, i
Performance of system operating procedure SOP-307, " Diesel Generator Fuel Oil System," Revision 9, was observed for delivery of new fuel Chemistry sampling results for the new fuel were reviewed and the inspectors verified that Technical Specification (TS) 4.8.1.1.2 requirements were met. The inspectors verified that TS minimum fuel oil levels in the EDG day tanks and the underground storage tank were met. On February l l'
22, the inspectors observed the filling of the A EDG fuel oil storage tank. During this process the level gauge manometer fluid blew out the top vent on the gauge when the fuel oil transfer pump was stopped. This rendered the gauge inoperable until fluid was
E
]
-
.
l
added to the gauge. Based on discussions with operations personnel, this failure had occurred previously. A maintenance work request was written to address this item.
d The inspectors assessed housekeeping and checked for combustible materials in the EDG rooms as well as assessed the general material condition. The inspectors found several small lube oil leaks on the EDGs. There were several small puddles of lube oil on the flooring underneath the diesels and indications of gasket leaks. Oil was standing on the camshaft cover near the number one cylinder of the B EDG and there was leakage around the engine driven lube oil pump on the A EDG. The inspectors were
!
concerned that the standing oil could be a potential combustible source around the j
EDGs. The maintenance manager and EDG system engineer were informed of these conditions.
l c.
Conclusions i
Based on a detailed walkdown the Emergency Diesel Generators (EDGs) were found to j
be properly aligned and in a standby condition per licensee procedures. Technical specification requirements for fuel oil and surveillance requirements were being met.
Several small EDG lube oilleaks were observed. The maintenance rule program I
properly monitored EDG performance.
Operator Training and Qualification
"
05.1 Simulator Examination Observations
)
a.
Insoection Scope (71707)
The inspectors observed three simulator exams being administered to licensed operator crews and attended the critiques of the examinations that followed.
b.
Observations and Findinas
On March 2 and 23, the inspectors observed licensed operator requalification simulator examinations. The simulator scenarios sufficiently challenged each licensed position.
'
'
Simulator scenario test objectives were met by the crews, and the operators passed the simulator portion of the requalification exam.
Overall, each crew appropriately responded to their event, took actions based on
appropriate prioritization commensurate with the safety significance; crews observed l
used or entered the correct annunciator response procedures, abnormal operating
,
l procedures, and emergency operating procedures; generally good oversight by supervision was evident including the ability to identify and implement appropriate TS actions; 3-way communications were generally performed with minor exceptions. Minor i
,
I deficiencies noted in the performance of an individual or crew (i.e. breakdown in i
I
_ communications, individual actions not in accordance with standards or expectations)
were appropriately captured by the examiners and discussed with the individual or crew.
!
The examination critiques observed were thorough and provided a comprehensive
,
assessment of individual and crew performance.
E
.
Also during the observations, the inspectors noted that physical differences in the configuration of the simulator and functioning of systems versus the actual control room were recognized by the training staff. These item were being tracked for resolution, c.
Conclusions The licensed operator simulator requalification examination scenarios were challenging and operators' performance met test objectives. Examination critiques were thorough and provided a comprehensive assessment of individual and crew performance.
Quality Assurance in Operations 07.1 Licensee Self Assessment Activi'ies - Manaaement Review Board (MRB)
a.
Inspection Scope (40500)
The inspectors observed the meeting of the MRB on March 18, to assess plant management review of the circumstances surrounding and the contributing factors to the plant transient of January 3,1999, b.
Observations and Findinas At plant management's request a MRB was conducted to review the failure of the moisture separator reheater pressure switch that caused power to be reduced to approximately 62% on January 3,1999. As a result of this meeting, good questions l
were generated and areas were identified where additional review and invesngation were needed.
l Based on presentations by the licensee staff, documentation provided to the inspectors, l
and discussions that occurred during the MRB meeting, the inspectors concluded that:
'
!
1) the root cause analysis performed on the failed pressure switch was appropriate; 2)
an effective reactivity impact review was completed by reactor engineering supervision; and 3) there was good use of simulator resources to aid in review of the event.
I The MRB recognized the need to initiate actions to review events earlier such as the i
one that occurred on January 3,1999. Areas the MRB identified from the January 3, 1999, event that needed additional review and investigation included TS 3.1.1, "Boration Control" applicability and verification of shutdown margin; the initial turbine load l
reduction rate selected to 0.5% was not appropriate; turbine control system response issue; development of formalized critiques for crew performance; and possible
'
enhancement of Abnormal Operating Procedure used during the transient. The MRB recommended that an incident Response Team (IRT) be implemented that would be responsible to review incidents or events for other than a plant trip, in a non-emergency plan situation, at the time the event occurred. The IRT would be composed of engineering and licensing personnel, operations staff, and an independent safety engineering group member.
t-I
-
.
c.
Conclusions A Management Review Board (MRB) held to review the plant transient of January 3,
!
1999, provided valuable insights into the contributing factors and circumstances L
surrounding the event. Both the inspectors and the MRB recognized the need to better l
understand the circumstances surrounding and contributing factors to this event in a
'
more timely manner, 11. Maintenance M1 Conduct of Maintenance M1.1 Observation of Work Activities (62707. 61726)
The inspectors observed all or portions of maintenance and surveillance testing i
'
activities listed below.
- ICP-130.004
" Boric Acid Blend Flow IFT00113," Revision 5, and Periodic Maintenance Task Sheets (PMTS) 9817752, loop calibration for
!
boric acid blending system flow transmitter, IFT-00113 i
- MWR 9702287 Condition Evaluation Report /Non-Conformance Notice 99-0131 for troubleshoot, repair, and temporary recorder hookup for XSW1DB05, Pressurizer Heater Backup Group 2 Breaker
- STP-128.024
"CO2 System Functional Refueling Test," Revision 8B, and Maintenance Work Request (MWR) 9715463, functional relay room CO2 fire system test
- STP 205.004
" Residual Heat Removal Pump and Valve Operability Test,"
Revision 4A
)
- STP 230.006A
" Emergency Core Cooling System / Charging Pump Operability Testing," Revision 3 j
- STP-360.002
" Fuel Handling Bridge Radiation Monitor (RM-G8) Operational Test," Revision 7 l'
The inspectors verified that work was performed with the work package present and actively referenced. All activities observed were conducted in accordance with established procedure instructions. Procedures provided sufficient detail and guidance i
for the intended activities. Technicians demonstrated that they were experienced and knowledgeable of their assigned tasks. Quality control personnel were present whenever required by procedure and when applicable. The inspectors noted that appropriate radiation control measures were in place. The inspectors concluded that
!
,
{
routine maintenance and surveillance activities were satisfactorily performed.
l i
3
!
!
U
-
.
M1.2 Review of Maintenance and Test Packaaes for Emeroency Core Coolina System (ECCS) Comoonents -
a.
Insoection Scope (61726)
~ The inspectors verified that completed PMTSs and Surveillance Test Procedure (STP)
test packages involving ECCS check valves and relief valves satisfied the applicable requirements including referenced TS surveillance requirements.
b.
Observations and Findinos The inspectors reviewed test package documentation for recent performances of the following maintenance and surveillance testing activities:
- STP-215.008,"ECCS Check Valve Testing"
- PMTS 9702299," Charging /SI Pump A Relief Valve Setpoint Testing"
- PMTS 9702300,"Si Accumulator Tank A Relief Valve Setpoint Testing"
- PMTS 9702301,"Si Accumulator Tank C Relief Valve Setpoint Testing"
- PMTS 9702302,"Si Accumulator Tank B Relief Valve Setpoint Testing"
.
- PMTS 9702303," Cold Leg injection Header Train A Relief Valve Setpoint Testing"
- PMTS 9702304, " Cold Leg injection Header Train B Relief Valve Setpoint Testing"
- PMTS 9702305," Hot Leg injection Header Relief Valve Setpoint Testing" No problems were identified. Ah stated testing requirements including TS surveillance requirements had been satisfied. Completed test packages demonstrated acceptable i
test results.
c.
Conclusions Nine completed surveillance test and preventive maintenance packages demonstrated acceptable test results for emergency core cooling system relief valves and check valves.
M1.3 Emeraency Diesel Generator Surveillance Observations
- a.
Insoection Scooe (61726)
The inspectors observed and reviewed surveillance testing activities related to the A EDG.
L
R
f
!
!
b.
Observations and Findinas l
l On March 24 the inspectors observed performance of STP-250.008, " Emergency l.
Diesel Generator A Support System Leak Test," Revision 1, STP-105.016, " Charging -
l Pump and Diesel Generator Slave Relay Testing," Revision 7A, and STP-125.002, i
" Diesel Generator Operability Test," Revision 18D for the A EDG. The inspectors observed the pre-job briefing and concluded that the pre-job brief was thorough, in that, the precautions, initial conditions, communications and coordination of the three l
- surveillance procedures with the test personnel were appropriately covered.
l Discussions with control room and local operators during the EDG testing indicated that the operators had a good level of knowledge concerning the surveillance tests and the EDG system. The tests were conducted in accordance with established procedures and met all the required acceptance criteria. Performance of STP-250.008 demonstrated that the A EDG support systems (air start system, fuel oil system, cooling water ar:d exhaust subsystems) operated properly.
During A EDG operation the inspectors and the leak test personnel noted two small oil and water leaks. The licensee had previously identified these leaks and had written MWRs. During the surveillance test it was noted that fuel oil suction strainer differential pressure (DP) indicator IPl05417 was reading less than zero. Operators and maintenance personnel responded to the issue expeditiously and determined that the pressure Indicator needed to be vented and refilled to ensure that indicator was i
operated properly. Earlier questions concerning the strainer DP indication while the EDG was in standby conditions were raised in February 1999; however, the licensee's initial assessment was that no action was required. This work resulted in the EDG having to run approximately one hour longer than a normal EDG surveillance run.
c.
Conclusions The A emergency diesel generator operability, slave relay and support system leak surveillance tests were performed in accordance with established procedures and demonstrated operability of the equipment in accordance with the TS surveillance requirements. Personnel conducting the tests demonstrated a good level of knowledge.
The pre-job briefing was thorough.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Maintenance / Material Condition of Reactor Coolant System (RCS) Pressure Isolation ValvM a.
Inspection Scoce (62700)
.The inspectors reviewed the licensee's program for maintenance and testing of RCS pressure isolation valves (PlVs) to determine the adequacy of that program for maintaining the integrity of RCS isolation boundaries. The inspectors also verified that the licensee's program for periodic leak rate testing of selected isolation valves satisfied l
TS 3.4.6.2.f and 4.4.6.2.2 for verification of RCS PlV leakage.
l
!
I
\\
.
.
l b.
Observations and Findinas
[
The inspectors reviewed machinery history and leak testing data for selected RCS PlVs
!
i to evaluate the adequacy of the program for maintaining the integrity of those RCS lsolation boundaries and to verify that TS 3.4.6.2.f and 4.4.6.2.2 requirements had been l
satisfied. Valves selected for review consisted of isolation valves, including check valves, which if failed could result in an interfacing system loss of coolant accident (ISLOCA). The inspectors reviewed the surveillance procedures for periodic leak rate j
testing of PlVs and as-found leakage test data for selected valves from testing
!
performed during RF-09 and RF-10 refueling outages. Specific leakage test packages
!
reviewed were listed in Section M1.2. The inspectors also reviewed selected maintenance procedures used by the licensee for disassembly and inspection of check valves as required by the inservice testing (IST) program.
The inspectors noted that each of the leakage testing procedures required that a corrected value for valve leakage be calculated for RCS normal operating pressure of 2215 psig. This corrected leakage value was required to be used rather than the actual observed leakage values anytime testing involved a lower test pressure.
The inspectors verifieo wt the program for maintenance and testing of PlVs had satisfied the TS requirer mts. The inspectors determined that no as-found leakage testing failures of RCS Pivs had occurred during the previous two refueling outages.
No examples of inadequate maintenance were identified during the review. No problems were identified during the review of machinery history which would indicate adverse trends or degradation of material condition of any RCS PlVs.
The licensee has experienced some operational backleakage from RCS PlVs. This backleakage has resulted in a portion of the RHR discharge piping being pressurized to approximately 544 psig. Potential backleakage paths have been identified but the exact source of the leakage is unknown. There have been no as-found failures for any of the PlVs in those potential flowpaths and previous leakage data taken during refueling outages does not show any adverse trends. Overall RCS unidentified leakage has remained approximately 0.2 gpm which was within TS limits. The licensee has estimated
'
the amount of leakage required to be approximately 0.01 gpm based on the rate of repressurization of the affected RHR piping. This value is well below the leakage criteria for any of the associated check valves in the potentialleakage paths. Based on information from Westinghouse, the licensee concluded that this is not an unexpected condition.
Inconsistent licensee design information resulted in the licensee evaluating this issue as a potentially deficient condition. Although the affected piping segment was designed for 600 psig and the original piping analysis was performed for 600 psig, the Gilbert piping specification listed the normal design pressure as below 535 psig. The licensee
'
interpreted this to mean that the piping segment had been in an upset condition. The l
piping specification further stated that the cumulative duration for all upset conditions was less than 1% of the total expected 40 year design life of the piping. Since the j
affected piping had no permanent pressure indication, the licensee had not been able to determine the actual amount of time the segment had been affected. Therefore, the licensee concluded that the affected piping had already been exposed to upset L
l.
.
conditions for greater than 1% of the 40 year design life. The inspectors discussed this issue with licensee management and members of the site engineering organization.
)
The inspectors were informed that the licensee had subsequently evaluated the affected piping as not being in a degraded condition and that an operability concern had not existed. The licensee's basis for this conclusion considered that overall RCS leakage
,
was well below TS limits and the original piping analysis was performed for 600 psig for 100% of the 40 year life. Based on information provided by the licensee, the inspectors concluded that the existing RCS operational backleakage had not represented a significant safety concern. The licensee's evaluation of this issue was continuing and was expected to include considerations of various options to possibly correct the
,
backleakage, additional review of industry information, and clarification of the apparent conflicting design information.
c.
Conclusions Review of leakage testing data indicated acceptable material condition for RCS isolation boundaries. No examples of inadequate maintenance were identified during this review.
No problems were identified during the review of machinery history which would indicate an adverse trend or degradation of the material condition of RCS PlVs.
M2.2 Testina of ASME Section XI Class 2 and 3 Relief Valves a.
Inspection Scoce (62700)
The inspectors reviewed results of lift setpoint testing of ASME Class 2 and 3 relief valves in the chemical and volume control (CS), safety injection (SI), and residual heat removal (RHR) systems performed during the two most recent refueling (RF) outages.
Verification of correct lift setpoints for these relief valves was necessary to ensure proper operation of ECCS systems and because of the potentialimpact of improper lift setpoints on a postulated ISLOCA event. Additionally, the inspectors reviewed corrective actions for as-found failures on relief valves and verified established relief valve setpoints were proper for piping design pressure values.
b.
Observations and Findinas ASME Section XI Class 2 and 3 relief valves included a number of smaller relief valves in various systems such as CS, Si, RHR, and other systems. The inspectors reviewed documentation for selected ASME Class 2 and 3 relief valves in the CS, Si and RHR systems that had been tested during the RF-09 and RF-10 outages. Specific relief valve test instructions reviewed were documented in Section M1.2. Some as-found lift setpoint failures had occurred for relief valves during this period but, in each case, lift setpoints had been readjusted whenever the as-found setpoint exceeded +/- 3% of nominal as required by ASME/ ANSI OM-1987, Part 1, " Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." The inspectors reviewed documentation for resetting lift setpoints for selected relief valves which had out-of-tolerance as-found lift setpoints. The inspectors compared lift setpoints from test procedures to plant drawings and verified that established relief valve setpoints were proper for piping design pressure values. Additionally, the inspectors reviewed L
e
-
.
I j
maintenance work packages and post maintenance test documentation for completed l
work on selected relief valves. No problems were identified during this review.
The inspectors noted that the licensee's pro em for testing ASME Section XI Class 2 and 3 relief valves was being revised to cW ct inconsistencies between ASME Code
- -
requirements and implementing procedureo. These problems had previously been l
identified by the licensee and were documented as NCV 50-395/98005-03. Corrective i
actions in this area were still in progress. The inspectors were informed that required revisions to test procedures would be issued prior to the upcoming refueling outage.
No problems were identified during this review of selected relief valve test packages and other documentation reviewed by the inspectors.
c.
Conclusions
'
l No problems were identified with setpoint testing of ASME Class 2 and 3 relief valves in
the chemical and volume control, safety injection, and residual heat removal systems.
Licensee corrective actions associated with previously identified deficiencies associated with ASME/ ANSI OM-1987, Part 1, requirements for ASME Section XI Class 2 and 3 relief valves was stillin progress.
M2.3 MeteoroloaicalTower Availability a.
Insoection Scooe (62707)
i
'
The inspectors reviewed corrective maintenance initiated as a result of equipment being out of service frequently, b.
Observations and Findings During the inspection period the inspectors reviewed the technical specification tracking log for equipment out of service. The inspectors noted that the meteorological tower (i.e., met tower) was out of service 23 out of 42 days during the inspection period.
Further review of available data by the inspectors indicated that the met tower had been
.
available approximately 65% of the time during calendar year 1999. In addition, the i
inspectors noted that the met tower has had 18 MWRs written and worked over the last l
15 months that resulted in approximately 941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br /> of out of service time. In particular
'the inspectors noted that corrective maintenance has been performed on the 10-meter
. temperature detector six times since November 1998.-
On page 3A-12 of the Final Safety Analysis Report (FSAR), the licensee l' dicates that n
after the first year of commercial operation, " Efforts for ninety percent annual data
recovery shall be continued after this period only for wind speed 10M (meters), wind j
direction 10M, and differential temperature (10-61M). Technical Specification 3.3.3.4,
!
" Meteorological Instrumentation," states that when the met tower is inoperable for more g
l than seven days a special report is required to be submitted outlining the cause of the malfunction and the plans for restoration. The licensee during this inspection period issued one special report that stated no further corrective actions were required.
However, the day the report was issue the met tower again failed. The licensee revised the special report to reflect the additional failure.
<
_
.
The inspectors discussed the system availability and the maintenance work history with the system engineer, engineering department supervisor, and the Instrument and Controls (l&C) supervisor responsible for repairs to the system. Discussions with the system engineer revealed that the licensee had not established a system to actively track availability time to ensure that the annual 90% data recovery is achieved. The system engineer and l&C supervisor were aware of the numerous system problems, but were unable to fully explain the causes. Design changes and component upgrades are being considered for future implementation by the licensee in an effort to make the met tower more reliable and available on a consistent basis.
The system is important during events to indicate wind speed, direction and temperature to aid in determining actions required during a radiological emergency, in the event that data from the met tower is unavailable, an Emergency Plan implementing procedure states that wind speed and direction can be obtained from the national weather service at the Columbia airport; however, temperature shall be obtained manually at or near the met tower. No event has occurred at Virgil C. Summer station that has required the licensee to obtain weather data _while the primary met tower was out of service.
c.
Conclusions Corrective maintenance and corrective actions have been ineffective in preventing an l
increased unavailability time for the meteorological tower during the last part of 1998 I
and 1999. The licensee had not established a system to actively track availability time to ensure that the Final Safety Analysis Report annual target of 90% data recovery is achieved.
Ill. Enaineerina L
E1 Conduct of Engineering E1.1 Analysis and Resolution of Batterv Dearadation a.
Insoection Scope (37551)
The inspectors reviewed the battery degradation problems that have occurred on both safety-related and non-safety-related batteries.
b.
Observations and Findinas The system engineer responsible for the batteries explained the licensee's evaluation and resolution of battery degradation problems. The two safety-related batteries are 125 VDC batteries comprised of LCR-31 cells _ manufactured by C&D Company. The l
batteries were installed in May 1990. The battery posts are starting to exhibit post seal j
leakage. A leaking post seal allows small amounts of electrolyte to come out of the l
- battery and causes corrosion on the connections. The manufacturer's representative came to the site and inspected the batteries. After examining the batteries, he concluded that the problem would become worse over time. The licensee decided to replace the batteries with the same style. The replacement will take place during the 1999 refueling outage. The manufacturer stated that the post seal had been redesigned i
l
-
.
after 1990, and therefore the new batteries should be less prone to post seal leakage.
The inspectors examined both of the safety-related batteries and observed that post seals on a few of the cells were slightly swelled and deformed. The inspectors observed that the cell to cell connections did not have any noticeable corrosion. The inspectors observed that there was virtually no sedimentation and that the color of the plates was normal.
One of the non-safety-related batteries is a 125 VDC battery comprised of LCW-49 cells. The battery was installed in 1985. This battery has been exhibiting erratic individual cell voltages (ICV) on a few of the cells. The manufacturer stated that there had been a certain limited number of manufacturing batches where an unsuitable type of plate separator was used. This separator material promoted a lead crystalline growth that forms a conductive path between plates. The path forms, burns off and reforms, which explains the erratic ICVs. The inspectors examined the non-safety-related LCW-49 battery and observed that the plates were severely bowed. There was no significant sedimentation and the color of the plates was normal. The licensee plans to replace this battery with the same style during the 1999 refueling outage.
c.
Conclusions Both trains of safety-related batteries have exhibited the early stages of post seal leakage. The licensee made a conservative decision to replace these batteries in the 1999 refuel outage. One non-safety related battery is approaching end of useful life and will also be replaced.
E1.2 Review of Desian Basis Document (DBD) Imorovement Project Plan a.
Insoection Scope (37550)
The inspectors held discussions with the cognizant design engineers and reviewed documents to gain an understanding of the scope of the licensee's new design basis document improvement project.
b.
Observations and Findinas i
The licensee has defined a three-phase plan to consolidate control of the Design Basis information into one centrallocation. The existing DBDs were prepared by the original Architect Engineer and the NSSS supplier according to their scope of supply. The licensee is conducting the improvement project in-house for the Architect Engineer
, portion and the NSSS supplier was contracted to work on their portion. One of the main activities of the improvement project is to locate and file a hard copy of all the pertinent referenco material in a special DBD file it is expected that references will be added and subtracted from the original DBDs during the course of the improvement project. If certain design parameters, upon which the final design was based, do not have a supporting calculation, the missing calculation will be generated. The inspectors were informed that the pilot DBD improvement effort on the emergency feedwater (EFW) and component cooling water (CCW) systems resulted in new calculations being generated.
Also, the project work will include looking for and resolving any discrepancies among l
And r, i
'
.
!
13 documents. The 37 DBDs will be " improved' over a five year period. The priorities of the individual systems were to be taken from the maintenance rule program risk ranking.
c.
Conclusions The licensee has initiated a design basis document improvement project to be completed over a five year period. Priorities for reworking / replacing / initiating calculations, technical reports, etc., will be set in accordance with Maintenance Rule risk ranking. The EFW and CCW systems DBDs were " improved as trial examples to help define the detailed plan and illustrate the need for the project.
E1.3 Review of Circuit Breaker Failure a.
inspection Scope (37551)
The inspectors reviewed the recent failure of a non-safety-related 7.2 kV circuit breaker to trip on demand.
b.
Obsentations and Findinas-On March 3,1999, the 7.2 kV circuit breaker which controls the B circulating water pump (non safety related) failed to trip on demand. The trip signal was initiated by the control switch at the main control room. Trouble shooting conducted under a work request determined that the trip coil was burnt out. As indicated in condition evaluation report (CER) 99-0178, the cause of the burnt out trip coil was a seized bearing on the trip latch roller. When the trip signal was initially applied, the trip coil operated as designed. The seized trip latch roller prevented the circuit breaker from opening. There were probably several attempts to trip the breaker which caused the trip coil to remain
,
energized for times beyond its design energization time which caused the coil to
!
overheat. As indicated in CER 99-0178, the same problem occurred in 1997.
According to the system engineer, the 1997 problem was assumed to be a weak trip coil. The coil was replaced, and the breaker appeared to function correctly. The breakers are Magna Blast type manufactured by General Electric Company.
Since the seized bearing on the trip latch roller could have been a direct result of lack of breaker refurbishment, the inspectors inquired as to the licensee's refurbishment program for these breakers. The system engineer and the electrical maintenance supervisor described the refurbishment program. Most of the safety-related breakers have been refurbished. The breakers at one of three non-safety-related buses plus one additional reactor coolant pump breaker have been refurbished. Many breakers will be refurbished during the spring 1999 refuel outage. Af ter the outage, the breakers for only one non-safety-related bus will need refurbishment, and that will be done during the following outage. The refurbishments described above are the first refurbishments since initial plant startup. The licensee plans to perform refurbishments on a ten year interval thereafter, alternating with the five year regular preventive maintenance.
The inspectors also inquired whether there had been any problems with the breakers that were recently refurbished. The system engineer explained that two problems had occurred with the breaker that controls the safety-related charging pump. The problems I
.
?
l
were _a wrong shim in the spring charging mechanism and an incorrect adjustment of the operating mechanism. Both these problems caused failures to close on demand. The problems occurred after the breaker had correctly operated at least 55 times (i.e. 50 i
l times at the refurbishment facility and five times at the plant). The li::ensee plans to add steps to their regular preventive maintenance procedure to specifically look for these two problems. Other problems, considered minor by the licensee, have been detected in the post-refurbishment (pre-operational) checks of the circuit breakers.
!
c.
Conclusions l
The licensee's program for refurbishment of 7.2 kV circuit breakers is being aggressively implemented, and should help preclude failures similar to the circulating
!
' water pump breaker failure.
E8 Miscellaneous Engineering lasues (92700, 92903)
E8.1 (Closed) Violation (VIO) 50-395/97014-03: failure to correctly prepare safety evaluation screening questions. This violation involved a change to the FSAR that was not recognized as a change in the 50.59 screening process. As stated in NRC inspection
Report No. 50-395/97-14, when the safety evaluation was performed after identification of this problem by the NRC, it correctly concluded that the change was allowed to be made under 50.59 and was technically acceptable. The inspectors reviewed the corrective actions stated in the violation response having to do with training of personnel, and found that they were adequately completed. Concerning the longer term
)
resolution to replace the broken resistance temperature detector, the inspectors reviewed Work Order 9719699 and the outage schedule, and was satisfied that the replacement would take place in the spring 1999 refuel outage barring unforseen
'
problems.
E8.2 (Closed) VIO 50-395/98003-01: failure to follow procedures. The inspectors reviewed Engineers Technical Work Record LC-13109 documentation as to the impact assesement for an out-of-tolerance test instrument. The impact was negligible. The inspectors reviewed the training records to confirm that the training discussed in the response to the violation had taken place for both the cut-of-tolerance and 50.59 problems. The inspectors'also discussed the training with one individual who participated in the training. The inspectors confirmed that routine instruments were transferred from the Test Unit to a Field Standard, as discussed in the response. A Field Standard instrument requires periodic calibration and entry of calibration date on the test data sheet, and is less susceptible to the cause of this violation.
E8.3 (' Closed) Licensee Event Reoort (LER) 50-395/98-04-00 and -01: unanalyzed condition
. for equipment qualification during a secondary system line break outside containment - -
emergency feedwater system. The problem in this LER was the loss of ability to isolate emergency feedwater flow to a faulted steam generator for secondary side breaks with a loss of offsite power (LOOP) plus failure of A train DC power. For equipment qualification considerations, the faulted steam generator must be isolated within 10 minutes. The cause of the loss of ability to isolate was related to how the control cucuitry responded to this scenario. The corrective action was to change the type of f
control switch which controls the steam supply valve for the turbir - driven emergency
,
i
p
.
.
,
l
'
feed water pump (TDEFW). The inspectors reviewed the modification package for the new switch and examined the new switch on the main control board. Based on this
. review the inspectors concluded that the problem was resolved. The inspectors L
concluded that identification of this subtle type single failure in the control circuitry represented good failure modes and effects analysis by the licensee. The problem was dispositioned as NCV 50-395/98003-02 previously.
E8.4 (Closed) LER 50-395/98-09-00 and -01: unanalyzed condition for speed controller on j
turbine driven emergency feedwater pump. The inspectors reviewed the safety
'
evaluation for removal of the instrument air from the l/P transducers at RHR valves 603A,603B,605A and 605B (50.59 Log 1998-0010). The inspectors reviewed the 50.59 screening for removing the instrument air from the l/P transducer for the TDEFW pump governor (1SYO2034). The inspectors also verified that the valves and control
,
l room switches had the proper caution tags for these changes. The NRC had determined that the design basis problem discussed in the LER was documented in a previous NRC inspection Report as NCV 50-395/98008-01.
E8.5 (Closed) Unresolved item (URI) 50-395/98007-02: Incorrect pipe displacements shown on the pipe support drawings. During review of pipe support drawings, the inspectors
' identified that predicted pipe movement information had been incorrectly transcribed-from the piping analysis to pipe support drawings. The licensee had subsequently issued CER 98-0728 to address the problem. The inspectors reviewed the status of corrective actions in this area. Based on this review, the inspectors concluded that adequate progress toward completion of corrective actions had occurred. Remaining corrective actions which included some additional review of pipe support drawings and issuance of corrected drawings was scheduled to be completed by March 31,1999.
The inspectors determined that the incorrectly transcribed pipe movement information previously available on pipe support drawings had not been used during the licensee's design change process. Based on the above review, the inspectors concluded that the incorrectly transcribed pipe movement information did not adversely affect activities important to plant safety such as plant modifications or engineering safety evaluations.
IV. Plant Support
~S1
. Conduct of Security and Safeguards Activities S1.1
_ Control of Escorted Visitors a.
Ingrection Scope (71750)
i The inspectors observed the control of escorted visitors during routme tours and inspection activities.
b.
Observations and Findinas On February 22 during observations of fuel oil offload activities near the EDG building, the inspectors observed that an operator, assigned escort responsibilities, failed to maintain control of a visitor. Upon identification of the incident by the inspectors the h
operator regained control of the visitor and then relinquished escort responsibilities to a t
L l
-
.
l l
security officer. The licensee subsequently conducted their own investigation and
,
substantiated the inspectors observations.
L (~
A fuel oil truck driver was admitted into the protected area on February 22, as a visitor, to offload fuel oil into the underground storage tanks located near the EDG building.
The driver was initially escorted by a security officer, but during fuel oil offload, escort j
responsibility for the driver was transferred to the operator. During the fuel oil transfer i
.
evolution a problem occurred with a tank level indicator and the operator walked towards the EDG building to use a different level indicator leaving the visitor at the truck. The inspectors noted that the operator had walked approximately 50 feet away from the visitor, thereby, failing to maintain visual contact and control of the visitor. During discussions with the inspectors, the operator indicated that he forgot about the visitor as a result of being involved in simultaneous tasks (i.e., offload of the fuel oil).
i l
Security Plan Implementing Procedure SPP-214 " Security identification Badge
Issuance," Revision 12A, Attachment 3, Duties and Responsibilities of Visitor Escorts, I
required the escort to remain with the assigned visitor at all times while in the protected area and ensure the visitor remains within the escort's control at all times. This failure to properly control a visitor within the protected area is a Severity Level IV violation that is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy.
_
This NCV is identified as NCV 50-395/99002-01. This incident was captured in the licensee's corrective action program as Condition Evaluation Report CER 99-0157.
i As part of the licensee's review of the contributing factors to the lost of visitor control, l
the inspectors learned that the transfer of the visitor from the' security officer to the operator was not performed in accordance with the licensees' expectations. At the time of the incident the inspectors questioned the security officer and the operator concerning escort responsibilities. Both persona indicated the escort responsibilities had been turned over to the operator. However, the licensee concluded that the
informal process used by the involved individuals contributed to the operator forgetting
'
l about his escort responsibilities.
c.-
Conclusions
- A non-cited violation was identified concerning failure to properly control an escorted visitor in the protected area. A contributing factor was an informal turnover of escort
,
responsibilities prior to the occurrence.
)
V. Manaaement Meetinas
.
X1
' Exit Meeting Summary The regional maintenance and engineering inspectors presented inspection results to l
members of the licensee management on February 26 and March 12,1999, i
respectively.
l l
The resident inspectors presented the inspection results to members of licensee
- management at the conclusion of the inspection on April 1,1999. The licensee l
acknowledged the findings presented.
'
/
r
'
i=
V
,
f-
f The inspectors asked the licensee whether any materials examined during the
_;
inspection should be considered proprietary. No proprietary information was identified.
L X2 Meeting with Local Officials l
On March 3,23, and April 5,1999, the inspectors met with local representatives from the surrounding Fairfield, Newberry, and Richland counties. Representatives from the t
county Emergency Preparedness Divisions, Emergency Services, Department of Public
- Safety, and the South Carolina Emergency Preparedness Division were contacted.
PARTIAL LIST OF PERSONS CONTACTED Licensee F. Bacon, Manager,- Chemistry Services L. Blue, Manager, Health Physics and Radwaste
- M. Browne, Manager, Plant Support Engineering S. Byrne, General Manager, Nuclear Plant Operations R. Clary, Manager, Quality Systems M. Fowlkes, Manager, Operations S. Furstenberg, Manager, Maintenance Services L. Hipp, Manager, Nuclear Protection Services D. Lavigne, General Manager, Nuclear Support Services G. Moffatt, Manager, Design Engineering A. Rice, Manager, Nuclear Licensing and Operating Experience G. Taylor, Vice President, Nuclear Operations R. White, Nuclear Coordinator, South Carolina Public Service Authority B. Williams, General Manager, Engineering Services G. Williams, Associate Manager, Operations INSPECTION PROCEDURES USED IP 37550: Engineering i
IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems -
IP 61726:- Surveillance Observations lP 62700: Maintenance implementation
,
IP 62707: Maintenance Observations
)
IP 71707: Plant Operations lP 71750:- Plant Support Activities IP 92903: Followup - Engineering L.
.u,
,
'
.
ITEMS OPENED AND CLOSED Opened 50-395/99002-01 NCV Improperly escorted visitor outside the diesel generator building (Section S1.1)
Closed 50-395/97014 03 VIO Failure to correctly prepare safety evaluation screening questions (Section E8.1)
50-395/98003-01 VIO Failure to follow procedures (Section E8.2)
50-395/98-04-00,-01 LER Unanalyzed condition for equipment qualification during a secondary system line break outside containment - emergency feedwater system (Section E8.3)
50-395/98-09-00,-01 LER Unanalyzed condition for speed controller on turbine driven emergency feedwater pump (6ection E8.4)
50-395/98007-02 URI Incorrect pipe displacements shown on the pipe support drawings (Section E8.5)
50-395/99002-01 NCV Improperly escorted visitor outside the diesel generator building (Section S1.1)
i