ML20207H533
| ML20207H533 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/07/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207H530 | List: |
| References | |
| 50-395-99-03, 50-395-99-3, NUDOCS 9906170130 | |
| Download: ML20207H533 (27) | |
See also: IR 05000395/1999003
Text
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U. S. NUCLEAR REGULATORY COMMISSION
REGION ll
Docket No.:
50-395
License No.:
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. Report No.:
50-395/99-03
Licensee:
South Carolina Electric & Gas (SCE&G)
Facility:
Virgil C. Summer Nuclear Station
Location:
P. O. Box 88
Jenkinsville, SC 29065
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Dates:
March 28 - May 8,1999
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~ Inspectors:
M. Widmann, Senior Resident inspector
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M. King, Resident inspector
L. Garner, Project Engineer, Ril (Sections M1.2.b.5, M8.3 and E8.1)
J. Coley, Reactor inspector, Ril (Sections M1.6 and M7.1)
E. Testa, Radiation Specialist, Ril (Sections R1.2 and R8.1)
Approved by:
R. C. Haag, Chief, Reactor Projects Branch 5
Division of Reactor Projects
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9906170130 990607
ADOCK 05000395
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EXECUTIVE SUMMARY
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Virgil C. Summer Nuclear Station
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NRC inspection Report No. 50-395/99-03
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This integrated inspection included aspects of licensee operations, rnaintenance, engineering,
and plant support. The report covers a six-week period of resident inspection;in addition, it
includes the results of announced inspections by a regional reactor inspector, project engineer,
and radiation specialist.
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Operations
The inspectors conducted frequent reviews and observations of ongoing plant
operations associated with refueling outage RF-11. In general, the conduct of
operations was professional, and safety conscious. Conservative decision making by
members of management was noted throughout the outage (Section 01.1).
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The power reduction, plant cooldown and shutdown operations in preparation for
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refueling were conducted safely and were well controlled with good communications
established between personnel. Operations management appropriately stressed the
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importance of monitoring and understanding the relationship between reactor vessel
level indication and inventory balances to ensure proper reactor coolant system
inventory control during shutdown conditions (Section 01.2).
A negative observation was noted for control board operators not being aware of the
cause for several illuminated control room annunciators. An example was the " Source
Range Hi Flux at Shutdown Blocked" annunciator being illuminated during fuel reload
with the operator being unaware of why it was acceptable to block this alarm function
(Section 01.2).
The inspectors concluded that core offload, reload and core verification were performed
in accordance with established procedures. Fuel handling activities were well controlled
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(Section 01.3).
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A non-cited violation was identified for failure to adequately perform a Technical
Specification required visual inspection for loose debris in the reactor building.
Following completion of the licensee's reactor building closecut inspection, the
inspectors found loose debris, including a rubber shoe, a plastic bag, and a cloth booty,
in the reactor building. Subsequent evaluation determined that the debris would have
had a negligible impact on sump performance (Section O2.1).
Maintenance
The inspectors observed good maintenance practices during refueling outage RF-11.
Preventative maintenance and maintenance activities were appropriate and properly
implemented in accordance with instructions provided and established work documents.
The inspectors concluded that outage maintenance activities were well performed
(Section M1.1).
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Selected design modifications and maintenance work requests on the A emergency
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diesel generator, the service water system, the A reactor coolant pump seal, the station
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batteries, and a safety injection valve were successfully implemented and satisfactorily
tested. Documents generated to support plant changes were thorough and provided
sufficient detail to accomplish the design changes (Section M1.2),
The licensee's troubleshooting plan for failures of General Electric 7.2 kV Magne-Blast
breakers to close was effective. Through the use of high speed video cameras the
licensee was able to identify the root cause. Corrective actions necessary to prevent
recurrence were completed. Additionally, the licensee made an 10 CFR 21 notification
for reporting a defect with substantial safety hazards that involved a common mode
failure (Section M1.3).
The observed surveillance activities were successfully completed by knowledgeable
personnel. When problems were encountered appropriate corrective actions were
implemented and adequate retests were performed. Procedures provided sufficient
detail and guidance for the intended surveillance activities. The licensee established
good communication and coordination between departments prior to commencement of
surveillance tests (Section M1.4).
During preparations for the train A integrated safeguards test, the control room
operating crew failed to establish an initial test condition for volume control tank (VCT)
level. After the inspectors identified this discrepancy, operators properly established
VCT level prior to the start of the test (Section M1.4).
Main steam safety valve (MSSV) surveillance tests were conducted in accordance with
approved procedures and all acceptance criteria was met. A discrepancy for an out of
date TS administrative control was identified and appropriately corrected. The licensee
informed the inspectors of their intentions to submit an amendment for TS Table 3.7-1 to
correct non-conservative controls for multiple inoperative MSSVs (Section M1.5).
inservice examination and test activities were performed, documented and evaluated in
accordance with approved procedures by certified, skilled, and knowledgeable
examiners (Section M1.6).
A non-cited violation was identified for failure to adequately vent the residual heat
removal pump casings as required by Technical Specifications (Section M8.1).
A non-cited violation was identified for the failure to functionally test portions of breaker
contrel circuits as required by Technical Specifications (Section M8.2).
A non-cited violation was identified for failure to perform a load test on a refueling
manipulator crane load cell prior to use as required by Technical Specifications
(Section M8.4).
Enaineerina
Based on the results of a Westinghouse safety assessment and the licensee's
replacement of 28 fuel assembly top nozzles prior to core reload, the inspectors
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concluded that the licensee appropriately evaluated and resolved issues associated with
fuel assembly top nozzle hold down spring screw failures. The licensee's conclusions
were reasonable and there are no safety concerns that would preclude the current Cycle
12 fuel load from meeting the reload safety analysis (Section E1.1).
A non-cited violation was identified for failure to correctly translate design requirements
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into specifications, drawings or procedures. Ten reactor building components, which
were required to operate after an accident and which could be submerged during an
accident, were not designed or evaluated for submergence (Section E8.1).
Plant Sucoort
Radiological conditions in radioactive material storage area.s, health physics facilities,
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and waste storage buildings were appropriate, areas were properly posted and material
~ was properly labeled. Personnel dosimetry devices were appropriately worn. Radiation
worker doses were being maintained well below regulatory limits and the licensee was
maintaining personnel exposure as low as is reasonably achievable (Section R1.2).
A non-cited violation was identified concerning failure to properly control access to a
high radiation area in the spent fuel pool building. Movement of spent fuel assemblies
past a drained spent fuel cask loading pit resulted in the high radiation area (Section
R1.2).
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Report Details
Summarv of Plant Status
The inspection period began with the unit at approximatety 91 percent power continuing with the
planned end-of-life coastdown in preparation for refueling outage (RF)-11. On April 1, power
was reduced to approximately 87 percent for main steam safety valve testing. On April 3, the
unit entered Mode 2 and the licensee opened the turbine generator breaker to commence
RF-11. At the end of the inspection period the refueling outage was nearing completion with
'the unit in Mode 3.
l. Operations
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Conduct of Operations
01.1 General Comments (71707)
The inspectors conducted frequent reviews and observations of ongoing plant
operations associated with refueling outage RF-11. In general, the conduct of
operations was professional, and safety conscious. Conservative decision making by
members of management was noted throughout the outage. Specific events and
noteworthy observations are detailed ~in the sections below.
01.2. Power Reduction and Plant Shutdown /Draindown
a.
Inspection Scope (71707)
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The inspectors observed portions of the end-of-life coastdown, power reduction for main
steam safety valve testing and plant shutdown and cooldown operations conducted to
place the reactor in a cold shutdown condition required to support RF-11 activities.
These operations were controlled by use of General Operating Procedures (GOPs).
b.
Observations and Findinas
The inspectors observed plant shutdown, cooldown, solid plant operations and reactor
coolant system (RCS) drain down. The following procedures were reviewed.
GOP-4, " Power Operation (Mode 1)," Revision 12
GOP-5," Reactor Shutdown from Startup to Hot Standby (Mode 2 to Mode 3),"
Revision 10
GOP-6, " Plant Shutdown from Hot Standby to Hot Shutdown (Mode 3 to
Mode 4)," Revision 8
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GOP-7," Plant Shutdown and Cooldown from Hot Shutdown to Cold Shutdown
- (Mode 4 to Mode 5)," Revision 7
GOP-10," Core Refueling (Mode 5 to Mode 6, Defuel and Refuel to Mode 6),"
Revision 10
Overall, the power reduction, plant cooldown and shutdown operations in preparation for
refueling were conducted safely and were well controlled with good communications
between personnel. Plant maneuvers from opening the generator breaker to cooling
down the plant to Mode 4 were performed without incident and in accordance with the
applicable GOPs. The inspectors verified proper shutdown margin, and that RCS and
pressurizer cooldown rates were maintained in mcordance with Technical Specifications
(TS) and procedurallimits. Unexpected cond? sns and alarms were properly addressed
with maintenance work requests or condition evaluation reports being generated when
necessary. Generally the inspectors noted good operator response to control room
annunciators. However, severalinstances were noted by inspectors where operators
were not cognizant of annunciator status. One specific example was the " Source Range
Hi Flux at Shutdown Blocked" annunciator. During fuel reload the board operator was
unable to state why, with core alterations in process, it was acceptable to block the
alarm function. The inspectors later learned that the alarm function was blocked as a
result of a change to Reactor Engineering Procedure (REP)-107.013," Core Reload,"
Revision 2, to eliminate nuisance alarms during core reload. The inspectors reviewed
the 10 CFR 50.59 evaluation associated with the procedure change and concluded that
the evaluation was appropriately performed. The inspectors highlighted to licensee
management that board operators were unaware of the cause of some important
Prior to RCS drain down the inspectors walked down the temporary RCS level sight
glass, used when level is below the vessel flange, and verified it was correctly installed
and vented. During the RCS draindown an unexpected offset developed between the
sight glass and the Reactor Vessel Level Indicating Systern (RVLIS). Although the
operators appropriately stopped the draindown and ensured that RVLIS was properly
vented, RVLIS continued to indicate approximately six percent above sight glass level.
Operations management discussed the condition with instrumentation and control
supervision and shift engineers ed decided to proceed with the draindown. The
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operators slowly lowered vessel le.el 'o approximately two feet below the reactor vessel
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head elevation while carefully monitor. 4 sight glass level, RVLIS indication and the
predicted reactor vessel level based on an RCS water inventory balance Although the
RVLIS indication offset continued following this level change, operations management
was satisfied that the draindown could be conducted safely. The operators completed
the vessel draindown to nine inches below the reactor vessel flange without incident.
The inspectors concluded that operator and operations management actions
appropriately stressed the importance of monitoring and understanding the relationship
between multiple level indicators and inventory balances to ensure proper RCS
inventory control during shutdown conditions. Condition Evaluation Report (CER) 99-
0432 was written to address the RVLIS offset from the predicted values.
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Concluqions
The pcwer reduction, plant cooldown and shutdown operations in preparation for
refueling were ccnducted safely and were well controlled with good communications
established between personnel. Operations management appropriately stressed the
importance of monitoring and understanding the relationship between reactor vessel
level indication and inventory balances to ensure proper RCS inventory control during
shutdown conditions.
' A negative observation was noted for control board operators not being aware of the
cause for several illuminated control room annunciators. An example was the " Source
Range Hi Flux at Shutdown Blocked" annunciator being illuminated during fuel reload
with the operator being unaware of why it was acceptable to block this alarm function.
01.3 Cor. 3ffload and Reload
a.
Inspection Scope (71707)
The inspectors observed portions of core offload and reload activities from the reactor
building manipulator crane, at the spent fuel poolin the fuel handling building and the
control room. The inspectors reviewed the adequacy of and adherence to established
core offload and reload procedures. These included REP-107.002," Core Offload,"
Revision 7A, and REP-107.013," Core Reload," Revision 2. In addition, the inspectors
observed the core verification activities.
b.
Observations and Findinas
During the offload of fuel the licensee encountered a fuel handling manipulator gripper
engagement problem as a result of fuel assembly top nozzle defects (reference Section
E1.1). Prior to recommencement of fuel offload the licensee implemented several
measures to ensure that each assembly was properly engaged prior to movement. With
those measures employed, successful completion of the fuel offload was accomplished
with no additional significant fuel handling problems. Following completion of
maintenance and activities related to defueled conditions, core reload was successfully
performed with no fuel handling issues.
The licensee offloaded and reloaded the core in accordance with their reshuffle plan.
Fuel handling activities were performed in a controlled manner and in accordance with
the specified procedures. Based on review of the instruction and guidance provided in
the fuel handling procedures the inspectors determined that the procedures
appropriately addressed fuel handling activities and as such were adequate.
c.
Conclusions
The inspectors concluded that core offload, reload and core verification were performed
in accordance with established procedures. Fuel handling activities were well controlled.
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02
Operational Status of Facilities and Equipment
02.1
Reactor Buildino Closcout
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a.
Inspection Scope (71707)
The inspectors performed a walk $wn of the reactor building (i.e., containment) to verify
no loose debris was evident.
b.
Observations and Findinos
On May 7, the inspectors conducted an inspection of containment to assess material
condition and general area housekeeping prior to reactor startup. At the time of this
inspection the licensee had completed a reactor building closecut visual inspection per
Quality Systems Procedure (OSP)-208," Inspection of Housekeeping and items in
Storage," Revision 9 and had entered Mode 4. In general, the material condition of the
containment was satisfactory. However, the inspectors identified debris and tools, such
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as a rubber shoe, a plastic bag, a cloth booty and other small items within accessible
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areas of containment. The debris identified by the inspectors was subsequently
removed from the reactor building. The inspectors estimated the total amount of debris
to be approximately two square feet. The inspectors were informed by engineering
personnel that the transportable debris could block approximately two percent of one
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train's sump suction screen and therefore this debris represented a negligible challenge
to sump performance. The inspectors agreed with this assessment.
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TS Surveillance Requirement 4.5.3.1," Emergency Core Cooling Systems," which is
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applicable for Mode 4, requires that the licensee perform a visual inspection per TS 4.5.2.c, to verify that no loose debris is present in the reactor building. Section 8.2.1.B
of OSP-208, which implements the requirements of surveillance requirement 4.5.2.c,
requires, in part, that the licensee identify and correct reactor building housekeeping
discrepancies prior to reactor building closeout. The failure to adequately remove loose
debris from the reactor building is a violation of TS 4.5.3.1. This Severity Level IV
violation is being treated as a Non Cited Violation (NCV), consistent with Appendix C of
the NRC Enforcement Policy. This violation is in the licensee's corrective action
program as CER 99-0748 and is identified as NCV 50-395/99003-01.
During their walkdown the inspectors also identified several minor material deficiencies
such as loose insulation and evidence of RCS leakage (boric acid crystals) at an incore
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seal table connection. The licensee was informed of these minor items.
c.
Conclusions
A non-cited violation was identified for failure to adequately perform a Technical
Specification required visual inspection for loose debris in the reactor building.
Following completion of the licensee's reactor building closecut inspection, the
inspectors found loose debris, including a rubber shoe, a plastic bag, and a cloth booty,
in the reactor building. Subsequent evaluation determined that the debris would have
had a negligible impact on sump performance.
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Miscellaneous Operations issues (92901)
08.1 The NRC recently revised NUREG-1600, Revision 1, " General Statement of Policy and
Procedures for NRC Enforcement Actions," (Enforcement Policy) by the addition of
Appendix C. Appendix C, Interim Enforcement Policy for Power Reactor Severity Level
IV Violations, effective March 11,1999, revises the NRC's enforcement approech for
Severity Level IV violations. Appendix C permits closure of most Severity Leval IV
violations, based on the violation being entered into the licensee's corrective action
program, as well as other considerations as described in the Appendix. The NRC has
conducted a review of the following Severity Level IV violation, verified the f allowing
corrective action document was initiated for the violation, and considers it e ppropriate to
close this violation consistent with Appendix C of the Enforcement Policy:
Violation Number
Corrective Action Proaram
File Number
50-395/98009-01
CER 98-1007
11. Maintenance
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Conduct of Maintenance
M1.1 Observation of Work Activities
a.
Inspection Scope (62707)
' The inspectors observed, reviewed or discussed all or portions of the maintenance
activities and associated site procedures listed below.
EMP-405.001
'7.2 kV Breaker Maintenance," Revision 15A, Alignment Checks
on XSW1DA12 Pressurizer Backup Group 1 Heater Breaker
GMP-100.007
" Maintenance Support For Refueling," Revision 12
lCP-310.008
" Source Range (N31) High Flux at Shutdown Bistable Calibration,"
Revision 4
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ICP-365.009
" Delta T-Tavg Control Rod Speed Power Mismatch Steam Dump
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TY-408," Revision 7
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MMP-500.006
" Reactor Vessel Internals and installation," Revision 9
MWR 9802760
Replace "O" Rings on intake Manifold and Cylinders B EDG
MWR 9803491
Repack of PCV-00445A (Pressurizer Power Operated Relief
Valve)
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MWR 9803728
Perform Preventative Maintenance in accordance with
MMP-180.033,"EDG Miscellaneous Maintenance," Revision 9
MWR 9803818
Hydroblast and Cleaning of A CCW Heat Exchanger
MWR 9807713
Installation of 90 Degree Elbow for Service Water Pipe at
XVG9627A-CC
MWR 9813468
XBA-1B NCN 98-0272 Battery Replacement for A Battery Cells
MWR 9813599
Repack XVX09356A-SS (Pressurizer Gas Sample Header
Isolation Valve)
MWR 9815047
Remove and Replace Piping per NCN 960510 Disposition
Number 3 Service Water for EDG Air Start Compressor
MWR 9817586
C Chiller Pipe Replacement (IPX04495B-HR-SW)
MWR 9900099
Replace EGB unit for Woodard Governor per NCN 971474
PMTS 9802949
Reactor Building Closeout inspection per QSP-208
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PMTS 9803919
Process l&C Rack Calibration for IPT00444 and IPT00445
(Pressurizer Pressure Control Transmitters)
PMTS 9812166
Clean and Inspect DG B Lube Oil Cooler
PMTS 9812167
Clean and Inspect Jacket Water Heat Exchanger
PMTS 9812168
Clean and Inspect Intercooler and injector Cooling Water Heat
Exchanger
b.
Observations and Findinas
The inspectoo observed that work was performed in accordance with written procedural
instructions w!+h the Work packages present and actively referenced. Procedures
provided sufficient detail and guidance 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.
c.
Conclusions
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The inspectors observed good maintenance practices during RF-11. Preventative
maintenance and maintenance activities were appropriate and properly implemented in
accordance with instructions provided and established work documents. The inspectors
concluded that outage maintenance activities were well performed.
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M1.2 Review of Sianificant Desian Modifications / Maintenance Work Reauests (MWRs)
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a.
inspection Scope (62707)
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The inspectors reviewed and observed portions of the following MWRs, modifications
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and design changes implemented during RF-11.
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b.
Observations and Findinas
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1.
Emergency Diesel Generator (EDG) A Heat Exchanger Replacement
(Engineering Change Request (ECR) 50056)
The inspectors observed portions of the heat exchanger replacement activities
associated with EDG A. The replacements were being performed as a result of
significant pitting identified on the intercooler, lube oil and jacket water heat
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exchangers. Issues identified during the removal and installation were
appropriately resolved by maintenance and engineering personnel. Licensee
and contractor personnel were sufficiently skilled to perform the assigned tasks.
Foreign material exclusion controls were properly established at the start of the
work and were maintained throughout.
The inspectors' review of design package ECR 50056 concluded that the design
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contained the appropriate documentation (i.e., safety evaluation, revised
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drawings, and updated equipment records) to support implementation of the field
work. The licensee completed the modification and performed the required
post-maintenance testing successfully.
2.
Service Water Elbow Replacement at XVG-09627A (Non-Conformance Notice
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(NCN) 98-0360)
Prior to RF-11 the licensee had requested a Class 3 pipe repair relief request to
allow use of code case N-416-1 for Class 1,2, or 3 piping and components. The
code case allowed the use of a temporary patch over the pin hole leak at the
elbow weld and allowed deferring permanent repairs until the unit was in a
shutdown mode where the system could be drained. The NRC granted use of
code case N-416-1. During RF-11 the licensee removed and replaced the
elbow. The system was returned to service and post maintenance functional
testing was completed successfully.
The inspectors reviewed the disposition to NCN 98-0369, which was used as a
basis to replace the elbow, and determined it to be appropriate. The inspectors
observed portions of that work and reviewed the work package after installation
was complete. No discrepancies or issues were noted during this review.
3.
Reactor Coolant Pump (RCP) A Seal Replacement (MWR 9809129)
As part of planned maintenance on RCPs, the licensee replaced the A RCP seal
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package during RF-11. This work was conducted in accordance with the
guidance prescribed in the MWR package. Examination of the removed seal
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package did not identify unusual or excessive wear. Installation of the new seal
package was completed successfully. A minor problem identified during post
maintenance functional testing was addressed by the licensee and
Westinghouse. Specifically, the number 2 seal was not seated properly and
allowed flow to bypass the seal and fill the standpipe. As a result, inadequate
sealleakoff concerns needed to be addressed prior to placing the pump in
service. The licensee was able to isolate the standpipe, create sufficient back
pressure and rotate the pump enough to properly seat the number 2 seal.
Adequate sealleak off was subsequently verified by the licensee along with
successful RCP operation following maintenance.
The inspectors concluded that the work was well controlled and actions taken to
resolve the seal seating issue were appropriate.
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Safety Related Station Battery Replacement (NCN 98-0272 and NCN 98-0755)
To resolve problems with battery post seal leakage the licensee replaced both
the train A and B batteries during RF-11. This issue was previously documented
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in NRC Integrated Inspection Report No. 50-395/99-02, Section E1.1.
The inspectors observed portions of the battery replacements and reviewed the
associated NCN dispositions. The licensee appropriately completed charging
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and discharging of the batteries in accordance with surveillance requirements
prior to returning the system to service.
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The inspectors concluded that the work was well controlled. The inspectors also
concluded that the replacement of the batteries was a conservative decision and
ensured the reliability of the batteries to provide their safety function.
5.
Pressure Relief Modification on XVG08839 (ECR 50088 and ECR 50088A)
In a December 29,1997, supplemental response to Generic Letter 95-07,
" Pressure Locking and Thermal Binding of Safety-Related Power Operated Gate
Valves," the licensee committed to modify valve XVG08889-SI by the end of RF-
11 to eliminate a potential pressure locking condition. XVG08889-Si is the
containment isolation valve for the safety injection flow path to the reactor
coolant system hot legs. The need to address this issue was discussed in
section E1.3.b.6 of NRC Integrated Inspection Report No. 50-395/97-01. During
RF-11, XVG08889-Si was modified under ECR 50088A, " Installation of
Pressure / Bonnet Relief System for XVG08889-SI," MWRs 9903940 and
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9903975 and other applicable engineering documentation. The inspectors
reviewed the modification documentation including the associated safety
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evaluation. The pressure relief line installed between the bonnet area and the
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valve body is an acceptable solution to address the potential pressure locking
condition and met the licensee's supplemental response commitment. The
safety evaluation was technically valid.
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c.
Conclusions
The inspectors concluded that selected design modifications and maintenance work
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requests on the A emergency diesel generator, the service water system, the A RCP
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seal, the station batteries, and a safety injection valve were successfully implemented
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and satisfactorily tested. Documents generated to support plant changes were thorough
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and provided sufficient detail to accomplish the design changes.
M1.3 7.2 kV Breaker Troubleshootina
(Closed) Licensee Event Report (LER) 50-395/99006-00: substantial safety hazard with
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GE 7.2 kV Magne-Blast circuit breakers
a.
inspection Scope (62707. 92700)
The inspectors observed portions of the troubleshooting activities and investigations to
address the failure of General Electric 7.2 kV Magne-Blast breakers to close.
b.
Observations and Findinos
On February 18, the licensee initiated NCN 99-0110 due to failure of circuit breaker
XSW1DB14 (C Charging Pump breaker) to close during a surveillance test. The
breaker would attempt to close, but as the mechanism started to move, it immediately
returned back to the open position. The breaker was shipped to the vendor for root
cause failure determination.
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During a receipt inspection the licensee discovered a second breaker that exhibited the
same failure mechanism as the charging pump breaker. The inspectors observed the
licensee's troubleshooting efforts which used a high speed video camera to analyre the
rapid actions involved with the breaker's failure to close. Analysis of the video tapes
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resulted in identification of the cause which was a cotter pin impacting on the latch
check switch mounting bracket. This problem was considered a defect in repair during
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b;eaker refurbishment by the original equipment manufacturer and is a common mode
failure mechanism for 7.2kV safety-related breakers. Subsequently on May 5, the
licensee issued a 10CFR21 notification for reporting defects with a substantial safety
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hazard.
The latch check switch was an optional feature when the breakers were purchased. As
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a result of the root cause determination the licensee's corrective actions removed this
feature and the associated latch mounting bracket and actuating paddle. The
inspectors verified removal of these components on the safety-related breakers.
Nonsafety-related breakers were also modified. These corrective actions addressed the
operability concerns that previously existed with the breakers. LER 50-395/99006-00
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was issued on May 17 documenting this condition. The inspectors reviewed the LER
and determined this condition was properly resolved and documented. This LER is
closed.
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c.
Conclusions
The licensee's troubleshooting plan for failures of General Electric 7.2 kV Magne-Blast
breakers to close was effective. Through the use of high speed video cameras the
licensee was able to identify the root cause. Corrective actions necessary to prevent
recurrence were completed. Additionally, the licensee made an 10 CFR 21 notification
for reporting a defect with substantial safety hazards that involved a common mode
failure.
M1.4 Surveillance Observation
a.
Inspection Scoce (61726)
The inspectors observed or reviewed all or portions of the following surveillance
activities and associated surveillance test procedures (STP) listed below.
STP 110.001
" Pre-Core Alterations Verifications," Revision 6A
STP-125.004A
" Diesel Generator Load Rejection Test," Revision 0A
STP-125.008
" Diesel Generator A Refueling Operability Test," Revision 4A
STP 125.010
" integrated Safeguards Test Train A," Revision 7H
STP-125.011
" Integrated Safeguards Test Train B," Revision 7J
STP 125.016
" Engineered Safety Features Loading Sequencer B Output 5
Test," Revision 1
STP-125.017
" Diesel Generator A Loss of Offsite Power Test," Revision 2D
STP-125.018
" Diesel Generator B Loss of Offsite Power Test," Revision 2G
STP-130.004
" Valve Operability Testing (Mode 4)," Revision 8
STP-130.005H
" Safety injection Valve Operability Testing (Mode 5)," Revision 5
STP-131.002
" Refueling Machine Hoist Test for Rod Unlatching," Revision 0
STP-172.001
"AMSAC System Output Relay Test," Revision 1
STP-205.017
" Accumulator Check Valve Flow Test," Revision 3A
STP-230.006A
"ECCS / Charging Pump Operability Testing (Refueling),"
Revision 3A
STP-230.006C
"ECCS / Residual Heat Removal Valve Operability Testing
(Refueling)," Revision 4
'
.
11
STP-345 034
"First Stage Turbine Pressure Instrument IPT- 446 Calibration,"
Revision 9
STP 345.038
" Reactor Trip System instrumentation Response Time Test,"
Revision 6A
u
l
STP 360.002
" Fuel Handling Bridge Area Radiation Monitor, (RM-G8)
Operational Test," Revision 8
j
)
STP-409.001
" Emergency Diesel Generator Refueling Inspection (for B EDG),"
Revision 5
STP-501.005
"DC Battery Charger Service Test," Revision 9
b.
Observations and Findinas
The inspectors observed performance of and reviewed in detail the integrated
)
safeguards test procedures due to their complexity and safety significance. These TS
required surveillances test the reactor protection, solid s: ate protection system logic, and
safeguards actuation systems to ensure proper operation of engineered safeguards
i
equipment. The inspectors reviewed the surveillance results and verified each test (six
l
per train) met the established acceptance criteria. The inspectors also reviewed the test
l
exception logs and verified that each item was being appropriately addressed. No
'
significant issues were identified in the test exception log.
Overall, the test activities observed were well-controlled. However, prior to performance
of the train A integrated safeguards test, the inspectors noted that the control room
operating crew did not properly establish the required plant initial conditions in
accordance with STP 125.010. Specifically, the board operators did not recognize tnat
volume control tank (VCT) level was approximately 70 percent prior to commencement
of the test, whereas, the STP required the volume to be initially established between 20-
25 percent and be maintained low enough to prevent over filling. If the test had
commenced with the VCT level at approximately 70 percent this would have resulted in
the lifting of the VCT relief valve and a resultant loss of inventory to the recycle hold-up
tank. This condition would have unnecessarily caused the VCT relief valve to lift and
potentially become a distraction during performance of the test. However, this condition
would not have impacted the results of the test and, therefore, was determined to be of
minor safety significance. Although a violation of the STP's initial conditions did not
occur because the discrepant initial condition was identified and corrected prior to the
start of the test, the inspectors were concerned that the board operators were prepared
to start the test without having recognized the discrepancy.
c.
Conclusions
The observed surveillance activities were successfully completed by knowledgeable
personnel. When problems were encountered appropriate corrective actions were
implemented and adequate retests were performed. Procedures provided sufficient
detail and guidance for the intended surveillance activities. The licensee established
F
e
12
good communication and coordination between departments prior to commencement of
surveillance tests.
During preparations for the train A integrated safeguards test, the control room
operating crew failed to establish an initial test condition for VCT level. After the
inspectors identified this discrepancy, operators properly established VCT level prior to
the start of the test.
M1.5 Main Steam Line Code Safety Valve Testino
a.
Insoection Scope (61726)
The inspectors observed main steam safety valve (MSSV) testing and reviewed test
data for surveillance procedure STP-401.002, " Main Steam Line Code Safety Valves
ASME Section XI Test," Revision 9A.
.
b.
Observations and Findinos
On April 1, the inspectors observed the MSSV tests and reviewed the data collected for
STP-401.002. All acceptance criteria was met and no concerns were identified with the
performance of the testing. The inspectors reviewed associated TS 3.7.1.1," Safety
'
Valves," and verified compliance with the maximum allowable power for on-line MSSV
testing of 87 percent power. The inspectors reviewed an associated Technical
Specification Relocation (TSR) form TSR-1000 which had been issued on November 17,
1995. This TSR provided the licensee's interpretation to TS 3.7.1.1 and amplifying
information relative to TS Table 3.7-1 which was found to be not conservative per
Westinghouse Vendor Notice (VEN 940006). TSR-1000 provided more limiting
administrative controls for multiple inoperable MSSVs than TS Table 3.7-1 and was to
be utilized until a subsequent TS change could be developed. Based on the inspectors'
followup questions, the licensee acknowledged that the TSR did not reflect values
consistent with the power uprate performed in 1996. The licensee generated CER 99-
0498 to capture this discrepancy and issued TSR-1000, Revision 2, which provided
values for Table 3.7-1 that reflected the uprate power. No adverse impact resulted from
use of incorrect values due to the licensee's testing methodology in which only one
MSSV is tested at a time. The TSRs did not change TS Table 3.7-1 limits for a single
inoperable MSSV. The inspectors concluded tnere are no operability concerns for
current or past operation as a result of this issue. The licensee has informed the
inspectors of their intentions to submit an amendment for TS Table 3.7-1.
c.
Conclusions
The observed MSSV surveillance tests were conducted in accordance with approved
procedures and all acceptance criteria was met. A discrepancy for an out of date TS
administrative control was identified and appropriately corrected. The licensee informed
the inspectors of their intentions to submit an amendment for TS Table 3.7-1 to correct
non-conservative controls for multiple inoperative MSSVs.
I
I
.
13
M1.6
Inservice inspection (ISI) - Observation of Work Activities
a.
Inspection Scoce (73753.}
The inspectors observed two ultrasonic nondestructive examinations (NDE) performed
on piping components to detect erosion / corrosion. The inspectors also observed two
snubber visual examinations, two snubber bench tests and two snubber spring tests.
,
The following component examinations and tests were observed:
HD-09-E01, Erosion / corrosion examination
HD-09-E03, Erosion / corrosion examination
a
,
MK-RCH-0058, Snubber bench test
j
MK-RCH-0077, Snubber bench test
MK-RCH 0355, Spring test and visual examination
!
MK-RCH-0373, Spring test and visual examination
The inspectors also held discussions with examiners, supervisors and cognizant
'
engineers, and reviewed documentation which included Summer's second 10-Year
Interval inservice Examination Program (ISE-3), Relief Requests RR-02 and RR-PT-1 A,
RF-11 ISI inspection plan, NDE and snubber test procedures, examiner certifications,
steam generator eddy current examination plan, erosion / corrosion outage plan and
microbiologically influenced corrosion (MIC) growth monitoring program.
b.
Observations and Findinas
The applicable code for the second ten-year ISI interval examinations at Summer was
the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel Code,Section XI, Division 1. The licensee was in the end-of-cycle
(EOC) 11 refueling outage, which was the last outage of the second 40 month
examination period. The inspectors observed certified ISI ultrasonic examiners calibrate
and conduct erosion / corrosion examinations and an examiner perform snubber visual
examinations. These examinations were observed to evaluate the effectiveness of NDE
inspection procedures; examiner skills; knowledge; and thoroughness in their
performance of the NDEs; and interpretation, evaluation, and acceptance of the test
results. The inspectors concluded that the NDE examination activities and snubber
tests were performed in accordance with the applicable procedures by skilled,
knowledgeable, and certified examiners.
c.
Conclusions
Inservice examination and snubber test activities were performed, documented and
evaluated in accordance with approved procedures by certified, skilled, and
knowledgeable examiners.
-
.
14
M7
Quality Assurance in Maintenance Activities
.
M7.1 Effectiveness of Licensee Controls in Inservice inspection (ISI) Activities
a.
Insoection Scope (73753)
During review of ISI documentation, the inspectors examined two examples where
controls setup by the licensee identified, resolved and prevented violations of ASME
Code requirements in the area of inservice inspection.
b.
Observations and Findinos
-
{
The inspectors reviewed two CERs that identified problems that were corrected before
the problems violated the ISI program,
c.
Conclusions
Problems within the ISI program were identified and resolved by the licensee before the
problems negatively impacted the ISI program.
M8
Miscellaneous Maintenance issues (92700,92902)
M8.1 (Closed) Escalated Enforcement item (EEI) 50-395/98007-01: missed Technical
Specification (TS) Surveillance Requirement (SR) to vent the Residual Heat Removal
(RHR) pump casings. This failure to vent RHR pump casings in accordance with TS SR - 4.5.2.b.2, " Emergency Core Cooling System," had been documented in CER 98-0754
and LER 50-395/98008 00," Missed Surveillance Test For ECCS Subsystems Tavg
2350 degrees F." Based on the piping and pump configuration, the licensee
determined that the performance of the surveillance as delineated in the procedure did
not adequately vent a non running pump. LER 50-395/98008-00 will remain open
pending inspector review of licensee actions to address long term venting requirements
for ECCS pumps and associated piping. As an interim measure, the licensee revised
STP-105.006, " Safety injection / Residual Heat Removal Monthly Flowpath Verification
Test," to address the required venting. The licensee subsequently performed the
venting procedure successfully. The failure to adequately vent the RHR pump casings
is a violation TS 4.5.2.b.2. This Severity Level IV violation is being treated as an Non-
Cited Violation (NCV), consistent with Appendix C of the NRC Enforcement Policy. This
violation is in the licensee's corrective action program as CER 98-0754 and is identified
as NCV 50-395/99003-02.
M8.2 (Closed) LER 50-395/99001-00: missed surveillance test for electrical equipment
protective devices. As a result of a review conducted into circulating water 7.2 kV circuit
breaker testing issues, the licensee discovered that they had missed surveillance tests
for electrical equipment protection devices for reactor coolant pump 7.2 kV breakers.
' Specifically, test current was not passed through a metal jumper which is needed for the
instantaneous trip function of the breaker. Due to the failure to test a portion of the
breaker control circuit, surveillance testing was not completed in accordance with TS 4.8.4.3.a.1(b)," Circuit Protection Devices." This condition has existed since initial plant
4
15
operation began. From the time of discovery the licensee was able to test the missed
portion of the RCP circuit successfully within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance of TS 4.0.3. The
failure to functionally test portions of breaker control circuits is a violation of TS 4.8.4.3.
T his Severity Level IV violation is being treated as an NCV, consistent with Appendix C
of the NRC Enforcement Policy. This violation is in the licensee's corrective action
)
program as CER 99-0196 and is identified as NOV 50-395/99003-03.
M8.3 { Closed) Violation 50-395/98004-01: failure to promptly identify and correct a condition
adverse to quality.
l
(Closed) LER 50-395/98003-00.-01: inadequate solid state protection system testing.
The LERs involved failure to completely test the P-14 and P 11 circuits of the solid state
protection system. The violation was issued for the failure to promptly test portions of
3
the feedwater isolation circuits when incomplete testing was identified. The inspectors
'
reviewed the licensee's July 14,1998, response to the Notice of Violation. Through
discussions with the licensee and review of revised procedures and completed test
documentation, the inspectors verified that the actions described in their response and
the LERs had been completed as discussed.
M8.4 (Closed) LER 50-395/99003-00 : missed surveillance on manipulator crane load cell.
This LER documents missed TS SR 4.9.6.2, * Manipulator Crane," which requires a load
test of the manipulator crane auxiliary hoist and associated load indicator prior to
movement of drive rods within the reactor pressure vessel. The inspectors reviewed the
cause of the event, analysis of the event and corrective actions. Based on this review,
the inspectors concluded the missed surveillance did not adversely affect activities
important to plant safety. The untested load cell, which was used to uniatch the control
rod drive shaft in location P-6, was subsequently tested satisfactorily. Corrective actions
described in the LER included a check of the load cell for accuracy and briefing the fuel
handling crew on use of the hoist equipment and the requirements to test load cells.
The licensee also plans to revise procedures associated with core alterations to include
cautions on load cell operation. The failure to perform a load test on the auxiliary hoist
i
load cell prior unlatching a rod drive is a violation of TS 4.9.6.2. This Severity Level IV
violation is being treated as an NCV, consistent with Appendix C of 'r,e NRC
Enforcement Policy. This violation is in the licensee's corrective a-6cn program as CER
99-0482 and is identified as NCV 50-395/99003-04.
Ill. Enaineerina
.
E1
Conduct of Engineering
E1,1
Enoineerina Evaluation of Fuel Assembly Too Nozzle Defect
a.
Inspection Scope (37551)
The inspectors reviewed the licensee's engineering evaluations for the apparent failure
of hold down spring screws on fuel assembly top nozzles in Region 12/M-series fuel.
7
'
.
16
b.
Observations and Findinas
On April 13, during core offload the licensee experienced difficulty latching a fuel
assembly in the fuel handling building upender with the spent fuel handling tool gripper.
In addition, while attempting to remove a fuel assembly from the reactor vessel, the
licensee encountered a similar gripper problem with the manipulator crane.
Investigation through close visual examination using high resolution cameras revealed a
gap between the top nozzle and the hold down spring clamp which indicated a potential
failure of the hold down spring screws. The gap was caused by upward movement of
the spring clamp (due to screw failure) which caused interference with the manipulator
crane gripper and the spent fuel handling tool gripper. This interference prevented the
gripper from fully seating on the top nozzle such that the gripper fingers would not fully
engage the nozzle. Engineering evaluation of this issue was conducted by the licensee
under CER/NCN 99-0542.
The licensee performed a camera /visualinspection of all the fuel assemblies. The twice
burn (Region 12/M-series) fuel assemblies appeared to be the only assemblies affected
by the failed cap screw phenomenon. There were 68 M-series assemblies in Cycle 11
and 28 of these assemblies were scheduled to be reloaded into the core for Cycle 12.
After consulting with the fuel suppiier (Westinghouse), the licensee decided to replace
all 28 top nozzles for the fuel assemblies in question. This replacement work allowed
the core design to remain unchanged. The new top nozzles had three minor differences
from the existing top nozzles, but were evaluated as meeting the original design
requirements. The inspectors reviewed Final Safety Analysis Report (FSAR) Section
4.2.1.2.2.2, " Top Nozzle" and Section 4.4.2.7.2, " Hydraulic Loads," which addressed the
top nozzle design features and the design basis for the fuel assembly hold down
springs. The licensee has submitted an LER 50-395/99004-00 in accordance with 10
CFR 50.73(a)(2)(ii) as a condition outside the design basis of the plant because the
Westinghouse assessment concluded fuel assembly upward movement was possible.
Failure of the hold down spring screws would prevent the springs from applying the
designed downward force on a fuel assembly. If this condition was to occur it would
create a condition not described in the FSAR (Section 4.4.2.7.2) which states the fuel
assembly hold down springs are designed to keep the fuel assemblies in contact with
the lower core piste under all Condition I and 11 events with the exception of the turbine
overspeed transient associated with a loss of externalload. The inspeciors concluded
the licensee has submitted the required report for a condition not described in the FSAR
and outside the design basis of the plant.
On April 30, the inspectors attended a Plant Safety Review Committee (PSRC)
presentation by engineering personnel on the Westinghouse safety assessment for this
issue. The inspectors reviewed Westinghouse safety assessment," Fuel Assembly Top
Nozzle Holddown Spring Screws." This safety assessment supported the continued
j
safe plant operation and specifically evaluated the 28 replacement top nozzles and the
potential for N-series assembly top nozzle failures. The N-series fuel was considered by
Westinghouse to be susceptible to the same phenomenon based on the fact that it
would be the only twice burn fuelin the core design. The Westinghouse assessment
i
discussed potential areas of concern such as generation of loose parts, reduction in
l
spring hold down force, accident effects, thermal-hydraulic and nuclear design
assessment, fuel rod design effects and affects on the Individual Plant Examination
!
.
,
i
17
(IPE). Additionally, the attachment to the safety assessment addressed special
concerns and precautions when handling fuel with failed top nozzle hold down spring
screws. Potential hold down spring screw failure root causes under investigation by
Westinghouse include screw material heat lot, improper torque, machine and equipment
changes and potential contamination. The PSRC, following review of the Westinghouse
safety assessment, concluded operation can proceed safely without any additional
restrictions during Cycle 12 and released any Mode 5 restrictions associated with
CER/NCN 99-0542.
On April 16, the NRC issued a Headquarters Daily Report," Potential Fuel Assembly Top
Nozzle Defect," on this issue. The licensee issued an INPO Operating Experience
report (OE9872)," inability to Properly Engage Fuel Assemblies During Core Offload,"
on April 27 and issued LER 50-395/99004-00," Failure of Top Nozzle Hold Down
i
Springs," on May 6. Westinghouse has entered this issue into their 10 CFR 21
evaluation process as a Potential Issue (PI-99-009) and is proceeding with the
investigation. Expected completion of the root cause analysis is August 29,1999,
c.
Conclusions
Based on the results of a Westinghouse safety assessment and the licensee's
replacement of 28 fuel assembly top nozzles prior to core reload, the inspectors
concluded that the licensee appropriately evaluated and resolved issues associated with
fuel assembly top nozzle hold down spring screw failures. The licensee's conclusions
were reasonable and there are no safety concerns that would preclude the current Cycle
12 fuel load from meeting the reload safety analysis.
E8
Miscellaneous Engineering issues (92700)
E8.1
(Closed) LER 50-395/99005-00: engineered safety features (ESF) components
potentially outside the design basis of the plant. During their FSAR verification program,
the licensee identified that under certain conditions Emergency Operating Procedure
2.2,' Transfer to Cold Leg Recirculation," Revision 11, did not require completion of the
changeover from the refueling water storage tank (RWST) to the reactor building sump
until the RWST was drained down to 6% level. As a result the flood levelinside the
reactor building could be approximately four inches higher than previously evaluated in
June 1997 (elevation 418.883' versus elevation 418.448') and additional important or
safety-related components could be submerged. The LER lists ten components which
needed to be addressed by the licensee. The inspectors reviewed the licensee's
revised flood level calculations, independently assessed what equipment would be
affected by the increased flood level, evaluated the acceptability of immediate actions
taken during RF-11 to address this issue, and reviewed the justifications to operate until
RF-12. The inspectors determined that the licensee's calculations and corrective
actions taken during RF-11 were adequate to support restart of the unit. In RF-12 the
licensee plans to take actions to move or modify equipment to allow flooding up to
elevation 420'. Calculations indicated that if the entire P.WST tank volume, from the
overfill line to the tank bottom, were added to the reactor building, the flood level would
not reach elevation 420'.
r
.
l
l
18
l
The design bases for the unit requires that reactor building components, which were to
operate after an accident and which could be submerged during an accident, be
designed for submergence. The June 1997 calculation to determine the proper flood
level of the reactor building based on the procedurally allowed amount of water which
can be transferred from the RWST was not correct. The ten affected reactor building
components were not designed nor evaluated for submergence. Because either
l
alternate or redundant components and operator actions would have been available to
l
accomplish the function of each affected component, the safety significance of this issue
l
was minimal. The failure to correctly translate the design bases into specifications,
l
drawings or procedures is a violation of 10 CFR 50 Appendix B Criterion 111. This
i
Severity Level IV violation is being treated as an NCV consistent with Appendix C of the
NRC Enforcement Policy. This violation is in the licensee's corrective action program as
NCN 990483. The NCV is identified as NCV 50-395/99003-05.
l
IV. Plant Support
l
R1
Radiological Protection and Chemistry (RP&C) Controls
R1.1
General Comments (71750)
!
The inspectors observed radiological controls during conduct of routine inspections and
observation of operation and maintenance activities during RF 11 and found them to be
l
acceptable.
l
l
R1.2 Tour of Radioloaical Protected Areas
l
l
a.
Insoection Scope (83750)
l
l
The inspectors reviewed personnel monitoring radiological postings, high radiation area
i
controls, posted radiation dose rates, radiologically controlled areas (RCAs),
contamination controls and container labeling. In addition, the inspectors discussed As
I
l
Low As is Reasonably Achievable (ALARA) work planning, attended pre-job worker
briefings, and observed job performances. The inspectors reviewed licensee records of
'
l
personnel radiation exposure and discussed ALARA program details, implementation
and goals. The inspectors toured health physics facilities, the auxiliary building, and
outside radioactive waste storage areas. Requirements for these areas were specified
j
l
in 10 CFR 20 and Technical Specifications.
{
.
b.
Observations and Findinas
During tours of the facility the inspectors observed that RCAs including radioactive
material storage areas (RMSAs), High Radiation Areas, and Locked High Radiation
Areas were appropriately posted and radioactive material was appropriately stored and
'
labeled.
Reviewed records showed that the licensee was tracking and trending personnel
contamination events (PCEs). The licensee had tracked approximately 130 PCEs for
the 1999 calender year to date which included skin and clothing contaminations.
L-
.
19
From review of steam generator and reactor building fuel gripper replacement radiation
work permits (RWPs), and observations of pre-job briefings and work activities in
progress via closed circuit television, the inspectors verified appropriate dose controls
and acceptable work practices.
The inspectors reviewed the corrective action program. In particular the inspectors
reviewed CER 99-0520 which described the movement of spent fuel within the spent
fuel pool (SFP) past a weir gate which separates the cask loading pit from the SFP.
The cask loading pit had been previously drained to perform maintenance. Health
physics failed to identify that this area would have significantly increased transient dose
rates with the cask loading pit drained. The licensee had previously transferred 22 fuel
assemblies past this location prict a the discovery of the transient elevated dose rates.
Three people observing fuel movement registered maximum dose rates of 215 mrem /hr
while standing at the hand rail between the drained cask loading pit and the
decontamination pit. After the cask loading pit was filled with water an underwater
reading of 20.5 R/hr was measured on contact with the weir gate as a spent fuel
assembly past the gate. The licensee discovered the elevated dose rates when
individuals logged out of the RCA with their electronic dosimeters. The licensee
immediately investigated the elevated dose rates, stopped the fuel transfers, removed
maintenance scaffolding and reflooded the pit prior to resuming fuel movement. Two of
the individuals each received a total dose of 1 mrem and the other individual received 2
mrem. A contributing factor to this event was limited manpower resources in that the
licensee elected to not remove the scaffolding after completion of the maintenance.
With scaffolding in the cask loading pit, the normal practice of having the pit full of water
when moving spent fuel assemblies could not be accomplished. The requirements to
control access to high radiations areas are specified in 10 CFR 20.1601. This failure to
properly control access to a high radiation area is a violation. This Severity IV violation
is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy.
This NCV is identified as NCV 50-395/99003-06. This issue is captured in the licensee's
corrective action program as CER 99-0520.
The outage exposure goal was set at 90 person-rem. At the time of the inspection
(April 23,1999), the uncorrected outage person-rem was about 72.319 person-rem.
This was slightly above the preaicted outage dose for that time.
The inspectors reviewed operational and administrative controls for entering the RCA
and performing work. These controls included the requirement that the applicable RWP
be reviewed and understood by workers prior to entering the RCA. The inspectors
reviewed selected RWPs for adequacy of the radiation protection requirements based
on work scope, location, and conditions. For the RWPs reviewed, the inspectors noted
that appropriate protective clothing and dosimetry were required. During tours of the
plant, the inspectors observed the adherence of plant workers to the RWP
requirements, personal dosimetry being worn in the appropriate location, workers
properly using friskers at RCA exit locations and properly exiting the protected area
through exit portal monitors.
1
l
}
-
.
20
,
c.
Conclusions
Radiological conditions in radioactive material storage areas, health physics facilities,
and waste storage buildings were appropriate, areas were properly posted and material
was properly labeled. Personnel dosimetry devices were appropriately worn. Radiation
,
worker doses were being maintained well below regulatory limits and the licensee was
'
maintaining personnel exposure as low as is reasonably achievable.
A non-cited violation was identified concerning failure to properly control access to a
high radiation area in the spent fuel pool building. Movement of spent fuel assemblies
past a drained spent fuel cask loading pit resulted in the high radiation area.
j
i
R8
Miscellaneous RP&C lssues (92904)
R8.1
(Closed) Violation 50-395/98001-04: failure to establish adequate programmatic
controls for temporary shielding. The inspectors reviewed the licensee's response to
the Notice of Violation dated April 17,1998. The inspectors selectively reviewed the
corrective actions. Procedure HPP-819," Temporary Shielding Evaluation, installation,
and Removal," Revision 10, was revised to include additional requirements. The
inspectors selectively reviewed the training attendance records and found them to be
acceptable for engineering evaluation of temporary shielding installations.
{
V. Manaaement Meetinas
X1
Exit Meeting Summary
i
\\
The regional inspectors presented their inspection results to members of licensee
management on April 9 and 23,1999.
The inspectors presented the inspection results to members of licensee management at
the conclusion of the six-week inspection on May 12,1999. Additionally,
teleconferences with licensee management were conducted on June 2 and June 7. The
licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary. No proprietary information was identified.
I
-.
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.
21
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. Waselus, Manager, Strategic Planning and Development
R. White, Nuclear Coordinator, South Carolina Public Service Authority
B. Williams, General Manager, Engineering Services
G. Williams, Associate Manager, Operations
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 61726:
Surveillance Observations
IP 62707:
Maintenance Observations
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 73753:
Inservice inspection, Observation of ISI Work Activities
IP 83750:
Occupational Radiation Exposure
IP 92700:
Licensee Event Reports
IP 92901:
Followup - Plant Operations
IP 92902:
Followup - Maintenance
IP 92904:
Followup - Plant Support
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-395/99003-01
failure to remove loose debris from the reactor
j
building (Section O2.1)
'
50-395/99003-02
missed technical specification surveillance
requirement to vent the residual heat removal
pump casings (Section M8.1)
f.
~
.
L
i
)
22
50-395/99003-03
missed surveillance test for electrical equipment
protective devices (Section M8.2)
50-395/99003-04
missed surveillance on manipulator crane load cell
(Section M8.4)
'
50-395/99003-05
failure to comply with 10 CFR 50 Appendix B
Criterion ill, in that, design basis involved with
l
submerged components was not correctly
)
translated into procedures (Section E8.1)
50-395/99003-06
failure to properly control access to a high radiation
area (Section R1.2)
Closed
1
l
50-395/99003-01
failure to remove loose debris from the reactor
building (Section O2.1)
50-395/98009-01
failure to follow procedures for documenting LCO
,
entries, two examples (Section 08.1)
50-395/99006-00
LER
substantial safety hazard with GE 7.2 kV magne-
blast circuit breakers (Section M1.3)
i
50-395/98007-01
eel
missed technical specification surveillance
requirement to vent the residual heat removal
pump casings (Section M8.1)
50-395/99003-02
missed technical specification surveillance
j
requirement to vent the residual heat removal
pump casings (Section M8.1)
l
50-395/99001-00
LER
missed surveillance test for electrical equipment
'
protective devices (Section M8.2)
1
50-395/99003-03
missed surveillance test for electrical equipment
protective devices (Section M8.2)
50-395/98004-01
failure to promptly identify and correct a condition
adverse to quality (Section M8.3)
50-395/98003-00,-01
LER
inadequate solid state protection system testing
(Section M8.3)
50-395/99003-00
LER
missed surveillance on manipulator crane load cell
(Section M8.4)
,
,
.
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23
50-395/99003-04
missed surveillance on manipulator crane load cell
(Section M8.4)
50-395/99005-00
LER
engineered safety features (ESF) components
potentially outside the design basis of the plant
(Section E8.1)
50-395/99003-05
failure to comply with_10 C'FR 50 Appendix B
Criterion ill, in that, design basis involved with
submerged components was not correctly
translated into procedures (Section EB.1)
50-395/99003-06
failure to properly control access to a high radiation
area (Section R1.2)
50-395/98001-04
failure to establish adequate programmatic controls
for temporary shielding (Section R8.1)
Discussed
50-395/93008-00
LER
missed surveillance test for ECCS subsystems-
Tavg 2 350 degrees F (Section M8.1)
50-395/99004 00
LER
failure of top nozzle hold down springs
(Section E1.1)
I
1
1
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