ML14181A841
| ML14181A841 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 09/16/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14181A839 | List: |
| References | |
| 50-261-96-10, NUDOCS 9609270210 | |
| Download: ML14181A841 (37) | |
See also: IR 05000261/1996010
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-261
License Nos:
Report No:
50-261/96-10
Licensee:
Carolina Power & Light (CP&L)
Facility:
H. B. Robinson Unit 2
Location:
2112 Old Camden Rd.
Hartsville, SC 29550
Dates:
July 7 - August 17, 1996
Inspectors:
J. Zeiler, Acting Senior Resident Inspector
P. Byron, Resident Inspector
R. Chou, Region II Inspector (M1.1)
L. King, Region II Inspector
J. Lenahan, Region II Inspector (E1.1, E1.2, E6,
E7.1)
G. Salyers, Region II Inspector (R1.2)
W. Tobin, Region II Inspector (S1.2)
Approved by:
M. Shymlock, Chief, Projects Branch 4
Division of Reactor Projects
Enclosure 2
9609270210 960916
PDR ADOCK 05000261
G
EXECUTIVE SUMMARY
H. B. Robinson Power Plant, Unit 2
NRC Inspection Report 50-261/96-10
This integrated inspection included aspects of licensee operations, maintenance,
engineering, and plant support. The report covers a six-week period of resident
inspection; in addition, it includes the results of an engineering inspection by two
Region II inspectors and inspection of aspects of the licensee's spent fuel shipment
program by three Region II inspectors.
Operations
The licensee's preparations for Hurricane Bertha were prompt and thorough.
Good monitoring and assessment of potentially changing weather conditions
were performed during the period (Section 01.2).
The inspectors determined that the downpower evolution to replace a leaking
upper bearing oil cooler to the A Heater Drain Pump was well planned and
coordinated (Section 01.3).
The inspectors concluded that with the exception of a minor Foreign Material
Exclusion Area (FMEA) poor practice, the FMEA controls associated with spent
fuel pool shuffling activities observed were adequate (Section 01.4).
Based on review of the draft 1996 Institute of Nuclear Power Operations
annual assessment of site activities, issues identified were consistent with NRC
perceptions of licensee performance (Section 08.1).
Maintenance
The procedures used for the operation, inspection, and maintenance of the
spent fuel shipping casks and cranes were adequate in providing the details for
conducting work activities. In general, the craft, technicians, and operators
were knowledgeable and skillful in performing their assigned activities;
however, a control problem was identified with ensuring that personnel
protection measures were established along the lift path (Section M1.1).
Maintenance activities associated with the B charging pump packing
replacement were properly performed (Section M1.2).
Routine surveillance test activities were well coordinated and adequately
performed (Section M1.3).
Engineering
The licensee's design change control process was judged to be adequate and
the modification packages reviewed were of good quality (Section E1.1 and
E1.2).
Progress on Robinson Engineering Support Section (RESS) backlog reduction
was noted (Section E6.1).
Self assessments performed by RESS were effective in identifying engineering
performance deficiencies and were useful in providing oversight to
management (Section E7.1).
A Violation was identified involving the licensee's failure to take adequate
corrective actions after it was identified in May 1995 that the design pressure
rating was exceeded for fourteen safety related solenoid operated valves. This
resulted in the design application deficiencies remaining unanalyzed until
May 1996 (Section E8.1).
Plant Support
The inspectors concluded effective procedures were implemented to safely
control irradiated fuel shipment activities. Minor spent fuel cask handling
procedure discrepancies identified were appropriately resolved by the licensee.
A Nuclear Assessment Section (NAS) assessment of the spent fuel shipment
program readiness was thorough and probing. NAS assessment and table top
exercise weaknesses were properly resolved and effective program
enhancements were implemented prior to conducting shipment activities. The
inspectors observed that adequate controls were implemented during the
August 12, 1996, spent fuel shipment (Section R1.2).
A Non-Cited Violation was identified for the failure to affix a radiation material
label on a spent fuel shipping cask container that was located in the Radiation
Control Area (RCA) (Section R1.3).
A Violation was identified for the failure to provide minimum illumination for
several areas of the protected area (Section S1.1).
Report Details
Summary of Plant Status
Unit 2 remained at power during the entire inspection period completing 408 days of
continuous operation. On July 9, a downpower to 82 percent was conducted to
replace the upper bearing oil cooler to the A Heater Drain Pump after a leak was
discovered. On August 7 and again on August 8, power was reduced to 60 percent
after the Heater Drain Pumps tripped unexpectedly. The unit was returned to 100
percent power on August 9 and remained at full power for the remainder of the report
period.
I. Operations
01
Conduct of Operations
01.1
General Comments (71707)
The inspectors conducted frequent control room tours to verify proper staffing,
operator attentiveness and communications, and adherence to approved
procedures. The inspectors attended daily operations turnover, management
review, and plan-of-the-day meetings to maintain awareness of overall plant
operations. Operator logs were reviewed to verify operational safety and
compliance with Technical Specifications (TSs). Instrumentation, computer
indications, and safety system lineups were periodically reviewed from the
Control Room to assess operability. Frequent plant tours were conducted to
observe equipment status and housekeeping. Condition Reports (CRs) were
routinely reviewed to assure that potential safety concerns and equipment
problems were reported and resolved.
In general, the conduct of operations was professional and safety-conscious.
Good plant equipment material conditions and housekeeping was noted
throughout the report period. Specific events and noteworthy observations are
detailed in the sections below.
01.2
Preparations for Hurricane Bertha
a.
Inspection Scope (71707, 71750)
Between July 10-12, the inspectors reviewed licensee preparations in response
for Hurricane Bertha. This included a review of Operations Management
Manual (OMM) procedure OMM-021, Operation During Adverse Weather
Conditions, Rev. 13, and verification that the actions prescribed by the
procedure were properly implemented.
b.
Observations and Findings
On July 10, at 11:30 a.m., the licensee began preparations for the possible
impact from Hurricane Bertha. At the time, the hurricane was still several days
away but was heading toward the site. The licensee initiated actions for a
hurricane warning in accordance with OMM-021. The inspectors reviewed the
procedure and verified that applicable actions were being completed. This
verification also included several walkdowns of the site to ensure that loose
items were properly stored or secured. Only minor items were identified and
discussed with licensee management.
Based on weather projections that hurricane force winds would not be
expected near the site, licensee management decided that a plant shutdown
was not necessary. On July 12, the hurricane passed within approximately
150 miles to the east, traveling from south to north. Maximum sustained winds
of approximately 25-30 mph were observed at the site. No significant damage
occurred onsite. Offsite power was maintained throughout the storm as well as
normal communications.
c.
Conclusions
The inspectors concluded that the licensee's readiness for the Hurricane's
arrival was prompt and thorough. The licensee continuously monitored the
progress and status of the hurricane and was sensitive to the potential for
changing weather conditions that could occur at the site.
01.3
Downpower to Repair Heater Drain Pump Oil Cooler Leak
a.
Inspection Scope (71707)
On July 9, the licensee conducted a downpower to 82 percent in order to
replace the oil cooler tq the A Heater Drain Pump (HDP). The inspectors
reviewed the licensee's activities associated with preparations and conduct of
the downpower evolution.
b.
Observations and Findings
On July 9, the licensee discovered that service water was leaking into the
upper bearing oil cooler to the A HDP. While no appreciable increase in
bearing temperature was observed at the time, the licensee decided to replace
the oil cooler. These activities involved reducing power to 82 percent, stopping
the A HDP, replacing the oil cooler, and returning to full power operation. Prior
to the downpower, operations management issued a Night Order which
described the scope of the activities and provided operator guidance on
3
maneuvering reactor power. Reactor engineering personnel determined that
xenon would not be a concern due to the limited power decrease.
Nonetheless, they provided Operations personnel with power maneuvering
rates to lessen the impact of the transient on core reactivity. The inspectors
verified that a pre-job briefing was held with operations, engineering, and
maintenance personnel to discuss details of the evolution, precautions, and
contingencies for potential problems. The downpower commenced at 10:00
p.m. on July 9. Following replacement of the oil cooler, the unit was returned
to full power at 3:40 a.m. on July 11 without incident.
c.
Conclusions
The inspectors determined that the downpower evolution was well planned and
coordinated. No discrepancies were identified.
01.4
Foreign Material Exclusion Discrepancies
a.
Inspection Scope (71707)
On July 18, the inspectors observed foreign material exclusion area (FEMA)
activities in the area of the spent fuel pool. The licensee was in the process of
shuffling spent fuel in the spent fuel racks in preparation for an upcoming
shipment to the Shearon Harris Nuclear plant. The inspectors reviewed the
FMEA procedure, and the material and personnel equipment logs.
b.
Observations and Findings
The inspectors observed some minor inconsistencies involving a Radiation
Control (RC) Technician in the foreign material exclusion area around the
spent fuel pool. The FMEA boundary was setup around three sides of the
spent fuel pool with a single access control point. A swinging gate was located
at the far end of the FMEA and was labeled as an emergency exit. The
inspectors observed an RC technician enter and exit the FMEA three times
using the emergency exit. The RC technician removed items from the FMEA
through the Access Control Point and reentered the FMEA with the items
through the emergency exit. These items were not logged out and into the
FMEA. While this demonstrated weak FMEA controls, the inspectors observed
no foreign material event.
The inspectors discussed their observations with the licensee. The licensee
counselled the RC technician, and indicated that the FMEA program would be
reassessed.
c.
Conclusions
The inspectors concluded that with the exception of a minor FMEA poor
practices, the FMEA controls associated with spent fuel pool shuffling activities
observed were adequate.
08
Miscellaneous Operations Issues
08.1
Institute of Nuclear Power Operations (INPO) Assessment
a.
Inspection Scope (71707)
The inspectors reviewed the second draft of the INPO annual assessment of
site activities conducted in April 1996.
b.
Observations and Findings
The inspectors found that issues identified were consistent with the NRC
perceptions of licensee performance. No safety significant issues that required
immediate attention were identified.
c.
Conclusions
No regional followup of the INPO identified issues is planned.
II. Maintenance
M1
Conduct of Maintenance
M1.1 Observation of Spent Fuel Cask Liftinq Operations
a.
Inspection Scope (60855)
The inspectors observed portions of spent fuel cask lifting and drying process
activities to verify that the activities were performed in accordance with the
applicable procedures and American National Standards Institute (ANSI)
Codes. The procedures and ANSI Codes used are listed below:
Procedure MMM-009, Operation, Testing, and Inspection of Cranes and
Material Handling Equipment, Rev. 20,
Training Procedure, Crane Operator, Rev. 15,
Procedure SFS-001, IF-300 Shipping Cask Operations, Rev. 13,
5
Procedure SFS-004, Spent Fuel Cask Crane Restricted Mode
Procedure, Rev.9, and,
ANSI B30.2, Overhead and Gantry Cranes.
b.
Observation and Findings
The inspectors observed portions of spent fuel cask operations in preparation
for cask shipment of the Shearon Harris Nuclear Plant. The IF-300 cask used
for this shipment can handle seven assemblies with a total weight for the fuel
and cask of approximately 68 tons. One of the two casks used for this
shipment had initially been lifted, installed at one end of the spent fuel pool,
and loaded with seven assemblies prior to the inspectors' arrival onsite. Since
the cask location was at the end of the spent fuel pool next to the yoke storage
and decontamination pits, the lifting and movement of the cask from the spent
fuel pool to the decontamination pit would not go over any adjacent spent fuel
assemblies.
The inspectors observed the following activities:
Daily morning briefing at 6:00 a.m. on August 6 and 7,
The preparation and movement of the cask closure head from the
decontamination pit to the top of the cask stored in the spent fuel pool,
The connection engagement for the primary yoke to trunnions at two
sides of the cask and the secondary yoke (redundant purpose) to the
bottom of cask,
The lifting from the pool, the movement of the cask from the spent fuel
pool over the yoke storage pit to the decontamination pit, and the
unloading in the decontamination pit,
The demineralized water rinse of wire cables, load block, yokes, etc.,
The tool and equipment lifting using the auxiliary load block,
The lifting and movement of load blocks and yokes back to the storage
pit,
The installation of nuts to tighten the closure head,
The drying process and leak inspection for the cask.
6
During the observation of cask, yokes, and equipment box movement, the
inspectors noticed one problem as discussed below.
The cask was moved over people inside the yoke storage pit as it traveled
from the spent fuel pool over the yoke storage pit to the decontamination pit.
During this move, the lower signalman walked immediately down from the
spent fuel floor to the decontamination pit floor after he directed the crane
operator to move the cask from the spent fuel floor, over the yoke storage pit,
to the decontamination pit. Because the crane operator could not see people
entering the yoke storage pit and the signalman was on the way down to the
decontamination pit, the cask inadvertently passed over people inside the yoke
storage pit. In addition, the upper signalman was also unable to see into the
yoke storage pit at this time. The signalmen did not observe the load path at
all time resulting in the cask movement over people. In this instance, the
licensee had not sufficiently preplanned this lift to ensure that either the entire
lift route was observable by a signalman or operator, or, for those areas not
observed, that controls were in place to restrict entry during the lift. The
problem was identified as a personnel safety hazard issue and did not affect
the actual performance of the cask movement.
The inspectors immediately informed the licensee about this issue. The
licensee issued CR 96-01836, "Lifting Heavy Loads", for evaluation.
The inspectors also reviewed data and completed information in the
performance copy of Procedure SFS-001 and records of crane operator
training, certification, recertification, and medical data for three crane operators
who performed the lift. The annual medical data reviewed included the check
on visual acuity, hearing, color vision, and fitness for crane operation.
c.
Conclusions
Overall, the cask lift was satisfactory. A control problem was identified with
ensuring that personnel safety measures were established along the lift path.
M1.2 B Charging Pump Packing Replacement
a.
Inspection Scope (62703)
On July 18, the inspectors reviewed and witnessed aspects of the packing
replacement of the B Charging Pump.
b.
Observations and Findings
In early July, the licensee identified that reactor coolant system unidentified
leakage had increased. The B Charging Pump was in operation at the time.
After swapping charging pumps, the licensee verified that unidentified leakage
was reduced back to its normal value, indicating that the increased unidentified
leakage was attributed to the B charging pump.
On July 18, the inspectors witnessed aspects of Work Request/Job Order 96
ACL1l for replacing the packing on the B charging pump. The instructions for
performing the packing replacement were contained in Corrective Maintenance
(CM) procedure CM-034, Charging Pump Stuffing Box Maintenance, Revision
9. One of the three pump plungers was also replaced as a conservative
measure after some minor scoring on the outside surface of the plunger was
identified. After completing the work, post-maintenance testing on the pump
was performed in accordance with Operations Surveillance Test (OST)
procedure OST-101-2, Chemical and Volume Control System (CVCS)
Component Test Charging Pump B. The pump was returned to service without
any major problems later that day. No discrepancies were identified.
c.
Conclusions
The inspectors concluded the B charging pump maintenance and testing was
performed in accordance with applicable procedures in a conscientious and
professional manner.
M1.3 Maintenance Surveillance Observations
a.
Inspection Scope (61726)
During the inspection period, the inspectors observed all or portions of various
maintenance surveillance activities performed by the licensee. These
surveillances were performed to meet the surveillance requirements of
applicable sections in TSs. The inspectors verified that approved procedures
were available and in use, test equipment in use was calibrated, test
prerequisites were met, shift pre-job briefings were performed, TS Limiting
Conditions for Operations (LCOs) were entered and adhered to, and testing
was accomplished by qualified personnel. Upon test completion, the
inspectors verified that test data was complete and met acceptance criteria,
and equipment restoration was properly completed. The inspectors observed
all or portions of the following surveillances:
OST-401
Emergency Diesels Slow Speed Start
EST-124
Response Time Testing of Reactor Coolant System RTDs
b.
Observations and Findinqs
The inspectors determined that the surveillances were performed in
accordance with the prescribed procedures. The inspectors reviewed the
results of the surveillance tests and verified that test acceptance criteria were
satisfied. Pre-job briefings were conducted by operations prior to testing which
resulted in good test coordination. The procedures provided detailed
precautions and instructions. The inspectors concluded that the tests were
properly performed.
c.
Conclusions
The inspectors determined that the observed surveillances were well
coordinated and controlled in accordance with applicable surveillance test
procedures.
M8
Miscellaneous Maintenance Issues (92902)
M8.1
(Closed) Licensee Event Report (LER) 93-015-00, Pressurizer Pressure
Transmitters Out of Calibration: This issue involved the repeated inadequate
calibration of the three pressurizer pressure transmitters due to personnel
error. Personnel performing the calibrations were part of the licensee's outage
traveling crew and were not adequately trained and experienced on the use of
the calibration equipment, specifically, a dead weight testing apparatus
resulting in the wrong weights being used for these transmitters. The
licensee's corrective actions for this problem included retraining traveling crew
personnel on the use of a dead weight tester and revising training
qualifications to include periodic retraining. The inspectors reviewed applicable
training records and qualifications and verified that the committed training
activities were completed. Since this incident, the inspectors noted that this
work is now only performed by site Instrumentation and Control (l&C)
personnel. The inspectors verified that training on the use of dead weight
testing instrumentation was being provided for site l&C personnel. This item is
closed.
M8.2 (Closed) LER 94-003-00, TS Required Shutdown Due to Emerqency Diesel
Generator Inoperability: This event involved the inoperability of the B
Emergency Diesel Generator (EDG) on February 18, 1994, when a locking pin
for the modulating air damper came loose and was propelled through the
engine's air system damaging the scavenging air blower and turbocharger.
The root cause of the damper pin failure was unknown at the time that the
LER was written.
The licensee provided a supplement to this LER (261/94-03-02) on
November 30, 1994, indicating that the root cause of the failure to be
inadequate corrective action for a similar failure of the air intake system
associated with the B EDG on February 12, 1994. The corrective actions for
this event included enhancements in the investigation procedures for plant
events. This issue was also the subject of Violation 50-261/94-08-02. The
licensee's corrective actions for this violation were previously reviewed and
found to be acceptable. The violation was closed in NRC Inspection Report
50-261/95-29. Supplement 2 of this LER (261/94-03-02) was reviewed and
closed in NRC Inspection Report 50-261/96-01. Therefore, based on these
previous reviews, this LER is closed.
M8.3 (Closed) LER 94-011-00, Technical Specification 3.0: Emergency Diesel
Generator Inoperability: This LER identified a condition where the plant was
operating at full power with one EDG out of service for maintenance and the
redundant EDG out of service for approximately three hours per day to meet
the operability testing requirements of the TS. During these testing evolutions
offsite power was available to the unit and operators were located in the room
of the EDG being tested with the ability to manually place the EDG in service
should offsite power be lost.
To resolve this issue, the licensee submitted a TS change request to the NRC
that eliminated, in most cases, the requirement to test the redundant EDG
when the other EDG is inoperable. The inspectors reviewed the TS and
verified that this change had been incorporated into the TS by Amendment
158. Based on this review, this LER is closed.
M8.4 (Closed) LER 94-019-01: TS Violation Due to Exceedinq Pressurizer
Cooldown Rate: This issue involved a condition prohibited by the plant TS.
Specifically, that the TS 3.1.2.3. limit for pressurizer heatup and cooldown had
been exceeded. Notice Of Violation (NOV) 50-261/94-23-02 was issued for
this event on November 28, 1994. The inspectors reviewed the licensee's
response to the violation dated December 27, 1994, and corrective actions for
LER 94-019-01.
The corrective action for Violation 94-23-02, specified in the licensee's
response to the NOV, had been reviewed by the NRC and found to be
adequate and properly implemented. The Violation was closed out in NRC
Report 50-261/95-30.
The inspectors verified that the corrective action specified in the response to
the NOV was the same as that specified in LER 94-019-01. Consequently,
LER 94-019-01 is closed out based on the previous review of the
implementation of corrective action.
10
Ill. Engineering
El
Conduct of Engineering
E1.1
Design Change Processes
a.
Inspection Scope (37550)
The inspectors reviewed the licensee's procedures which control the design
change program.
b.
Observations and Findings
The inspectors reviewed the revisions of the procedures listed below to verify
that design control measures were consistent with 10 CFR 50, Appendix B,
Criterion Ill and 10 CFR 50.59. The following procedures were reviewed:
PLP-032, 10 CFR 50.59 Reviews of Changes, Tests, and Experiments, Rev.7,
dated February 20, 1996; PLP-054, Configuration Control, Rev. 6, dated
July 17, 1996; PLP-064, Engineering Service Requests, Rev.5, dated July 19,
1996; MOD-022, Administrative Procedure for Engineering Service Request
Major Modifications, Rev.4, dated March 16, 1996; MOD-018, Temporary
Modifications, Rev. 16, dated January 23, 1996; EGR-NGGC-003, Design
Review Requirements, Rev. 0, dated June 3, 1996; EGR-NGGC-0005,
Engineering Service Requests, Rev. 1, dated July 29, 1996; and EGR-NGGC
0304, Maintenance of Design Documents, Rev.0, dated November 11, 1995.
The inspectors concluded that the procedures adequately addressed: design
input, training, drawing changes, post-modification testing, design verification,
control of field changes, 10 CFR 50.59 safety evaluations, and As Low As
Reasonably Achievable (ALARA) reviews. The inspectors concluded that
adequate controls were in place to ensure effective implementation of design
changes. However, the inspectors noted that when the new EGR-NGGC
procedures were issued to improve design control activities, previously issued
procedures which they were meant to replace were not deleted and/or
canceled. For example, EGR-NGGC-005 was issued to replace procedures
PLP-064 and MOD-022. The inspectors noted that PLP-064 and MOD-022
were still being maintained current. EGR-NGGC-0005, Engineering Service
Requests, streamlined the process for performing engineering work.
EGR-NGGC-0005 is 88 pages while procedure PLP-064 is 162 pages. The
162 page procedure is very cumbersome to use. The inspectors discussed
with licensee personnel involved in the change to the new procedure the
benefits involved in the change. The responsible engineer now has formalized
responsibility to track through modifications until completion. This includes
reviewing the work request that installs the modification and reviewing the
testing of the modification. The new procedure also contains a matrix that
shows which documents are required by the modification and which are
optional. Procedure EGR-NGGC-005 also contains checklists to simply the
design process. The EGR-NGGC series of procedures are corporate level
procedures being issued to standardize engineering work activities on all three
Carolina Power and Light (CP&L) nuclear plants.
c.
Conclusions
The inspectors concluded that the licensee's design change control procedures
complied with the requirements of 10 CFR 50.59, and 10 CFR 50, Appendix B,
Criterion Ill. However, the inspectors noted that duplicate procedures exist
which could possibly result in confusion in the future and could result in
potential design errors. This was discussed with the licensee.
E1.2
Review of Design Changes and Modification Packages
a.
Inspection Scope (37550)
The inspectors reviewed the design change and modification packages to: 1)
determine the adequacy of the safety evaluation screening and the 10 CFR
50.59 safety evaluations; 2) verify that the modifications were reviewed and
approved in accordance with Technical Specifications and administrative
controls; 3) verify that applicable design bases were included; 4) verify that
Updated Final Safety Analysis Report requirements were met; 5) verify that
both installation testing and post modification testing requirements were
specified so that adequate testing would be accomplished. The inspectors
selected major modifications, minor modifications and a temporary modification
to review. The only difference between major and minor modifications is cost
and engineering involvement.
b.
Observations and Findings
The inspectors reviewed the following design change and modification
packages:
ESR-9400731
Penetration Protection System (PPS) Design Change
Containment Vessel Penetration Repair
ESR-9500327
Dampening Adjustments to Steam Flow Transmitters
ESR-9500633
Main Steam Isolation Valve (MSIV) Manual Control Valve
Delete
12
ESR-9500738
Service Water Header Leak Repair
ESR-9500764
Replacement of End of Qualified Life Environmental
Qualification (EQ) Cables
ESR-9500783
Modify HVH-1,2,3,4 to Leave Butterfly Valves Open
ESR-9500782
Resolve Generic Implementing Procedure (GIP)
Issues for RFO 17
ESR-9500870
Power Operated Relief Valve (PORV) Block Valve Stem
Replacement
The above design change and modification packages were scheduled to be
implemented during the next refueling outage, Refueling Outage 17
(RFO 17). The inspectors found that the modification packages had been
reviewed and approved in accordance with the licensee's design control
procedures and that the format and content of the modification packages was
consistent with the design control procedure. The quality of the modification
packages was good overall with only a few minor discrepancies being noted in
the ESR 9500764 package. These discrepancies included errors in the bill of
materials and incomplete instructions pertaining to cable pulling. Cable pulling
was addressed by Note 6k on Drawing number HBR2-0B060, Electrical
Installation Practices. The note stated that care should be taken to ensure that
cables are not over tensioned during cable pulling; however, there were no
specific requirements for control of pulling tension or side-wall pressure.
Licensee engineers stated that additional instructions would be issued to
address these requirements. None of the noted discrepancies would have
prevented successful implementation of the modification or resulted in an
inadequate modification package. The scope of each modification was found
to be consistent with the problem resolution outlined in the Engineering
Support Request. The 10 CFR 50.59 Safety Evaluations were found to be
adequate. The installation and test instructions were considered adequate to
implement the modification and verify that it performed in accordance with
design. The inspectors also verified that the UFSAR and other documents e.g.
drawings and procedures had been identified in the modification packages for
revision.
The modifications reviewed were prepared using procedure PLP-064.
Changes to the modifications can be processed using either procedure
PLP-064 or EGR-NGGC-005 guidance. The inspectors found that the
basic information was contained in the packages but that it varied in
content due to the flexibility allowed in procedure PLP-064. Examples
included the following: Form 4 which tracked action items was an option
13
and in some cases was included with the procedure and in other cases
it was used in close out to identify open items. Some of the procedures
included the ALARA review and others did not but checked the design
verification checklist as ALARA completed. The inspectors determined
that an ALARA review had been done on these modifications but was
not included in the package. The inspectors obtained copies of the
ALARA review from radiation protection and learned that it had been
accomplished using AP-040, ALARA Planning/ Dose Planning, Rev. 4,
dated June 26, 1996. The inspectors did not find the design verification
checklist in ESR-9500870; however, further review of this issue
disclosed that the licensee used an alternative method to document the
design review. The alternative method was conducted in accordance
with the procedure.
c.
Conclusions
In general, the modification packages were judged to be of good quality
and would not degrade plant performance, safety, or reliability. The
modification packages contained sufficient specifications, drawings and
procedures to be properly installed and tested. The licensee's 10 CFR
50.59 evaluations were completed in accordance with NRC
requirements.
E6
Engineering Organization and Administration
E6.1
Engineerinq Backlog
a.
Inspection Scope (37550)
The inspectors reviewed the backlog of open items in the Robinson
Engineering Support Section (RESS).
b.
Observations and Findings
The backlog of items in the RESS include engineering service requests
(ESRs) which include modifications, temporary modifications, drawing
changes, other engineering documents with outstanding changes, and
other engineering items, including open condition reports and
engineering commitments. The licensee's performance report for the
week of July 31, 1996, showed approximately 600 open engineering
work items. The licensee has recently completed a self-assessment,
discussed in paragraph E7, below, regarding management of the
engineering backlog. Actions were being planned to address the
problems identified during the self-assessment and to continue reduction
of te eginerig
wrk ackog.14
of the engineering work backlog. The long term goal was to reduce the
total number of open items in RESS to less than 200.
c.
Conclusions
The inspectors concluded that the licensee has made progress in
identification of the backlog of engineering work in RESS. Progress was
being made in reduction of the backlog.
E7
Quality Assurance in Engineering Activities
E7.1
Quality Assurance Assessment and Oversight
a.
Inspection Scope (37550 and 37551)
The inspectors reviewed self-assessments performed within the RESS.
b.
Observations and Findings
Self-assessments are part of the overall CP&L quality assurance
program at Robinson. The self-assessments were performed in
accordance with procedure PLP-057, Self-Assessment, Revision 4,
dated November 3, 1995. The results of these assessments were
categorized as strengths, or findings. The self-assessments reviewed
by the inspector were the results from recently completed assessment
numbers RESS96-015, RESS Organization & Administration; and RESS96-026, Environmental Qualification (EQ) Program at Robinson Nuclear
Plant (RNP). Several findings were identified in Assessment 96-026.
Six Condition Reports were written to document discrepancies identified
in the EQ program; however, none of the problems resulted in
identification of any inoperable equipment. The conclusion of the
assessment was that the Robinson EQ program meets overall EQ
requirements.
The inspectors discussed the results of Assessment 96-015 with the site
engineering manager. Several issues were identified regarding management
of the engineering backlog. These included overdue action items, work not
assigned to individuals or assigned to individuals no longer onsite,
discrepancies in the ESR data base, older modifications which require
closeout, and failure to include some items in the open engineering work which
affect the weekly/monthly engineering performance indicators. The final report
for assessment 96-015 had not been completed as of the inspection date;
however, CR number 96-01823 was issued and other CRs were being
prepared to document and disposition findings.
15
c.
Conclusions
The inspectors concluded that the self-assessments performed by RESS
were effective in identifying engineering performance deficiencies and
were useful in providing oversight to management. Managers in RESS
have been proactive in following up on issues identified at other sites to
identify and correct deficiencies in engineering work at RNP.
E7.2
Special UFSAR Review
A recent discovery of a licensee operating their facility in a manner contrary to
the Updated Final Safety Analysis Report (UFSAR) description highlighted the
need for a special focused review that compares plant practices, procedures
and/or parameters to the UFSAR descriptions. While performing the inspection
discussed in this report, the inspectors reviewed selected portions of the
UFSAR that related to the areas inspected. The inspectors verified that for the
select portions of the UFSAR reviewed, the UFSAR wording was consistent
with the observed plant practices, procedures and/or parameters.
E8
Miscellaneous Engineering Issues (37551 and 92903)
E8.1
(Closed) Unresolved Item (URI) 50-261/96-08-01, Review Licensee
Investigation and Resolution of Solenoid Valve Discrepancies:
Background
This issue involved the licensee's evaluations and corrective actions to
address design problems identified with solenoid valves (SOVs). The design
problems were identified after the ASCO 3-way SOV, which controls one of the
two containment isolation valves in the Steam Generator A Blowdown sample
line, was found to be leaking past its vent port while the SOV was deenergized
and closed. Subsequent investigations revealed that the regulated supply air
pressure (85 pounds per square inch gauge (psig)) exceeded the SOV's
maximum design rating (60 pounds per square inch differential (psid)). This
design rating is called the maximum operating pressure differential (MOPD)
and corresponds to the rating of the SOV's internal spring force acting to keep
the supply air from pressurizing the SOV inlet port. Supplying higher air
pressure than the SOV valve is designed for can result in air leaking past its
inlet or vent port seats. While leakage past the vent port does not create a
significant problem, leakage past the inlet port could prevent or interfere with
the closure of the associated air operated valve that the SOV controls. The
SOVs for the other five containment isolation valves in the Steam Generator
16
Blowdown sample lines were also found to be under-rated. All six SOVs in this
application were subsequently replaced.
Licensee Investigations
Further investigation by the licensee determined that this MOPD application
problem was much broader in scope. SOVs in both safety related and non
safety related applications were affected. In an effort to thoroughly investigate
and resolve this problem, engineering initiated an evaluation of the MOPD
versus supplied air pressure to all SOVs in the plant with priorities placed on
safety related applications. This included evaluation of approximately 850
SOVs. The inspectors verified that as the evaluation progressed, CRs were
initiated for MOPD application discrepancies identified. Generally, there were
three main areas of MOPD concerns identified by the licensee. These three
areas were as follows:
1)
MOPD Below Air Regulator Setting:
This area included SOVs where their MOPD was below the setting of
the regulator that was installed upstream to limit air pressure to the
SOV. In these cases, the SOV would be pressurized above its design
rating and leakage could potentially occur. A total of 12 safety related
SOVs were identified in this area. This number included the SOVs
associated with the six Steam Generator Blowdown Sample
Containment Isolation Valves discussed above. The other six SOVs
were for the feedwater flow control and bypass isolation valves. The
licensee performed testing of similar model ASCO SOVs. These valves
were determined to be acceptable for interim use until they could be
replaced during the upcoming refueling outage in September 1996.
2)
MOPD Below Instrument Air System Normal Operating Pressure:
These problems included SOVs with MOPDs that were below 100 psig,
the normal operating pressure of the Instrument Air (IA) system. Credit
was not taken for the pressure regulators to limit pressure since they
were procured as non-safety related components. Assuming the
regulator fails would result in the SOV being pressurized to the normal
IA system pressure. This would allow SOV overpressurization if it were
rated below 100 psig resulting in potential leakage. A total of 28 safety
related valves were identified in this area. The licensee planned to
replace these SOVs during the upcoming refueling outage in September
1996.
17
3)
MOPD Below Instrument Air System Maximum Design Pressure:
While the normal operating pressure of the IA system is 100 psig, its
maximum design pressure is 125 psig. Therefore, the licensee
assumed SOVs with an MOPD less than 125 psig were also susceptible
to overpressurization. Again, credit was not taken for the pressure
regulators. A total of 19 SOVs were identified in this area. The
licensee planned to replace these SOVs during the upcoming refueling
outage in September 1996.
At the end of this inspection period, the licensee was in the process of
completing their evaluation of safety-related SOVs. Similar evaluations were to
be completed for non-safety related SOVs that could have an adverse impact
on the plant.
The inspectors concluded that the licensee was conducting an exhaustive
investigation to completely resolve the SOV MOPD concerns.
Root Cause
The inspectors reviewed the licensee's actions in response to NRC Information
Notice 88-24, Failures of Air-Operated Valves Affecting Safety Related
Systems, dated May 13, 1988. This Notice alerted licensees of potential SOV
overpressurization failures caused by exceeding the MOPD rating. The
inspectors learned that the licensee had failed to evaluate the concerns
addressed by the Notice in 1988 due to an engineering organization oversight.
In February 1995, during an NRC commitment to perform a sample review of
their operating experience program, the licensee became aware that
Information Notice 88-24 had not been adequately evaluated. At that time, CR
95-00549 was initiated to reevaluate the concerns addressed by the Notice.
Action Item #1 of the CR, requested a review of MOPD versus supplied air
pressure for all safety related SOVs. As a result of this review, 14 safety
related SOVs were identified where the supplied air pressure exceeded the
MOPD rating. The evaluator failed to recognize the potential significance of
this finding and did not initiate a separate Condition Report or Operability
Determination for the deficiencies identified. As a result, the impact of the
MOPD discrepancies was not evaluated. The inspectors also noted that the
evaluator was unable to determine the MOPD rating for eight other SOVs due
to their model numbers being unknown at the time. No further review or
apparent attempt was made to determine the model numbers and MOPD
ratings. A status of "unknown" was documented for these valves with respect
to whether their MOPD was exceeded. Action Item #1 was closed after review
by the evaluator's supervisor on May 19, 1995, without having initiated a CR,
operbilty
eterinaion or
18
operability determination, or identifying the missing SOV information. Action
Item #2 of CR 95-00549 stated that the SOVs identified with their MOPD
exceeded would be replaced during the September 1996 refueling outage.
However, the inspectors noted that Work Requests had not been prepared to
ensure that this work would be scheduled. While this action item was still
open in the licensee's CR database, with a due date of Refueling Outage 17
(September 1996), it was unclear whether the action item would have been
identified to have completed the work during the outage.
Conclusion
10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires in part, that
measures be established to assure that conditions adverse to quality, such as
failures, malfunctions, defective material and equipment, are promptly identified
and corrected.
The inspectors concluded this issue was a violation of 10 CFR 50, Appendix B,
Criterion XVI, in that the licensee failed to take adequate corrective actions
after it was identified that the supplied air pressure exceeded the MOPD for 14
safety related SOVs. As a result, the adverse conditions remained unanalyzed
until May 1996. This item is identified as Violation (VIO) 50-261/96-10-01:
Inadequate Corrective Actions for SOV Design Discrepancies.
E8.2
(Closed) VIO 261/94-24-01, Inadequate Testing of Alternate AC Power
Source: The licensee responded to this violation in a letter dated November
11, 1994. This violation involved inadequate procedures for testing to
demonstrate the one hour capability of the Station Blackout alternate AC power
source. The licensee's corrective actions involved review of the test data and
procedure changes to improve management controls over test activities. The
licensee concluded that the testing performed demonstrated the station
blackout capability. In an acknowledgement letter to the licensee dated
January 30, 1995, NRC concurred with the licensee that the test activity was
adequate. The inspectors verified that all of the corrective actions had been
completed. The inspectors verified that all test document records were
assembled into a consolidated and readily available package. The licensee's
November 11, 1994, letter also contained a commitment to replace submerged
cables associated with NCV 261/94-24-02 which could not be qualified by
testing. The completion of upgrading or replacing submerged cables as
outlined in Modification M-1165 was completed by end of RFO16. The
inspectors reviewed the modification package documentation and verified that
the unqualified cables were replaced. This item is closed.
19
E8.3 (Open) Inspector Followup Item (IFI) 261/94-08-02, Incorporate 24 Hr Load
Testing into TS Surveillance Requirements: This IFI involved the licensee's
commitment to revise the TSs to require testing during each refueling outage
with a proper power factor to demonstrate the ability of the emergency diesel
generators to carry accident loads. This commitment was part of the
licensee's corrective actions for NRC Violation 50-261/93-07-01. The licensee
submitted the TS change in a letter to NRC dated January 30, 1996. A
request for additional information was sent by the NRC to the licensee on April
12, 1996, which the licensee responded to in a letter dated May 20, 1996.
This IFI will remain open pending review of implementation of the new TS
requirements after the revised TS is issued.
E8.4 (Closed) Escalated Enforcement Item (EEI) 50-261/94-16-04, Inadequate
Control Room Ventilation Testing Program: The Control Room Ventilation
System (CRVS) design was incomplete in that it did not consider all modes of
Heating, Ventilation, and Air Conditioning (HVAC) system lineups and the
effect of these lineups on Control Room habitability. The UFSAR, Sections 6.4
and 9.4.2, requires that the CRVS be capable of maintaining the control room
at a positive differential pressure with respect to adjacent areas and the
outdoors when the CRVS is operated in the emergency pressurization mode.
Neither design reviews nor surveillance testing identified that the CRVS was
unable to meet this requirement.
On May 7, 1994, the licensee identified during special ventilation balancing
testing that air pressure in Room E1/E2 which is adjacent to the control room
exceeded control room pressure. The licensee determined that, under certain
accident conditions, the Auxiliary Building Supply Fan (HVS-1) would continue
to supply air to Room E1/E2 while credit could not be taken for the ventilation
exhaust fan (HVE-7).
This event is described in more detail in Inspection
Report 50-261/94-16 and the licensee's September 29, 1994, response to the
violation.
The licensee revised plant operating procedures and emergency operating
procedures to place restrictions on HVS-1.
Engineering Surveillance Test,
EST-023, Control Room Emergency Ventilation System (once per 18 months),
was revised by Revision 10 to test the CRVS in the "worst case" mode and to
compare Control Room pressure to adjacent areas to ensure positive Control
Room pressure.
The inspector reviewed EST-023, Revisions 10 and 12 and noted that Section
8.6.2 requires that both HVE-7 and HVS-1 be secured prior to taking pressure
measurements in the Control Room and adjacent areas. Abnormal Operating
Procedure, AOP-005, Radiation Monitoring System, Revision 15 was also
reviewed and the response to the Control Room Radiation Monitor alarm is to
20
open the circuit breaker for HVS-1. The inspectors verified that the licensee
revised its procedures and tests the CRVS in the "worst case" mode. The
inspectors consider that the licensee has completed its corrective actions and
this item is closed.
E8.5
(Closed) LER 50-261/94-008-01, Condition Outside Design Basis Due to
Control Room HVAC Inoperability: This event was described and reviewed in
the previous Section E8.4. The item is closed.
IV. Plant Support
R1
Radiological Protection and Chemistry Controls (71750)
R1.1
Tours of the Radiological Control Area (RCA)
The inspectors periodically toured the RCA during the inspection period.
Radiological control practices were observed and discussed with radiological
control personnel including RCA entry and exit, survey postings, locked high
radiation areas, and radiological area material conditions. With one exception
discussed in Section R1.3, the inspectors concluded that radiation control
practices were proper.
R1.2
Irradiated Fuel Shipment
a.
Inspection Scope (40500, 71750, and 86750)
The inspectors reviewed the licensee's requirements and procedures related to
irradiated fuel shipment. The review included a NAS assessment of shipment
readiness and table top exercises conducted with outside agencies.
In addition, the inspectors observed activities associated with the receipt of
empty spent fuel shipping casks, cask loading, decontamination, and the
August 12, 1996 shipment.
b.
Observations and Findings
NAS Assessment of Spent Fuel Shipment Readiness
In May 1996, NAS conducted an assessment of the Robinson spent fuel
shipping program in order to determine the readiness of the program to
conduct effective shipping activities. The results of this assessment were
documented in NAS Report R-SF-96-01, dated May 31, 1996. The inspectors
reviewed the report and determined that the assessment was thorough and
21
probing. The assessment identified one strength, four issues, and one
weakness. The major problems identified involved the following:
Some safety features for the spent fuel handling system which were
described in various licensing documents, were not procedurally
controlled or tested,
The training and qualification of spent fuel team members was not
effectively administered,
The Spent Fuel Shipping Manual, Certificate of Compliance, Safety
Analysis Report, and various technical manuals were not effectively
controlled,
The inspectors verified that CRs were initiated to address the problems
identified and that necessary actions were initiated to correct these problems
prior to initiation of actual spent fuel shipment activities. The inspectors noted
good management attention and sensitivity in correcting these problems prior
to the fuel shipment.
Receipt of Empty Spent Fuel Shippinq Casks
On July 9, the inspectors attended a pre-job briefing held prior to bringing two
empty spent fuel shipping casks on railcars inside the protected area. The
meeting was attended by personnel from maintenance, operations, radiation
protection, and corporate fuel shipping area that had actions or responsibilities
in moving the cask inside the protected area. The inspectors noted that good
discussions were held on the details and logistics for moving the casks. A
management representative was assigned to coordinate the activity in
accordance with PLP-37, for infrequent evolutions. As a result of good
coordination and planning, the casks were brought in without any major
incident. The inspectors reviewed the shipping receipt package, including the
radiological surveys of the railcars, to verify that the railcars were properly
received. No discrepancies were identified.
Table Top Exercise with Outside Agencies
On July 23, the licensee held a "table top" exercise with their staff and a
representative from the South Carolina Emergency Preparedness Divison
(EPD) Director's office. The exercise was to validate the procedures
necessary to address an accident involving the spent fuel shipment. This was
the first spent fuel shipment in several years. The exercise revealed that the
coordination between the licensee and the state organizations needed to be
improved.
On uly31,anoher"tale op"
22
On July 31, another "table top" exercise was held and included representatives
from all the involved South Carolina state and local agencies. The State
Police were concerned about the timely transmission of radiological
information. The differences between the state's and CP&L's emergency plan
was the most significant issue that surfaced during the exercise. The State
EPD had written their plan based on the licensee's plan. The licensee revised
their plan in the interim and had not advised the state of their action. The
issue was resolved by both organizations working together to resolve the
differences which consisted of reporting protocol. The licensee documented
the identified issues in CR 96-01797.
The inspectors concluded that the table top exercises with the outside
agencies revealed communication and coordination weaknesses in sufficient
time to have been resolved prior to the shipment.
Compliance with the Cask Certificate of Compliance
The inspectors reviewed whether the licensee met the conditions specified in
the Model No. 300 Spent Fuel Shipping Cask Certificate of Compliance (COC).
Based on a review of the list of authorized users, the inspectors verified that
Carolina Power & Light was a registered user of the IF-300 spent fuel shipping
cask. The inspectors reviewed Revision 31 of the COC and selected eighteen
of the specifications in the COC for verification of compliance.
The inspectors reviewed the licensee's Irradiated Fuel Data Sheets (IFDS)
dated July 22, 1996. The IFDSs provided information on the fuel's physical
characteristic, fissionable isotopic composition, and history, and a direct
comparison between specifications in the COC fuel requirements and the fuel
being shipped. The inspectors made an independent comparison between the
information in the IFDSs to the fuel's specifications in the COC.
Selected pages from the license's completed procedure, Corrective
Maintenance Procedure CM-M0303, Cask and Equipment Skid Annual
Inspection (IF-300 Series), Revision 6, were reviewed to verify that the casks
were being maintained. The completed procedure indicated that the
maintenance was performed on IF-303 and IF-304 during June and July 1996,
and that the following selected maintenance specifications from the COC were
performed:
96-ACG, Hydrostatic pressure test and annual leakage test
Installation of a new head gasket,
Installation of a new rupture disk,
96-ACG, test of cask precon valves and circle seal valves, and,
23
0 Testing of the relief and reseat pressure for the two neutron shield relief
valves plus their leak test results.
The inspectors reviewed completed procedure, Spent Fuel Shipping Procedure
SFS-001, IF-300 Shipping Cask Operations Revision 13, dated July 30, 1996,
and verified the procedure contained steps for draining and purging the cask.
Purging of the cask was a specification in the COC.
The inspectors determined that all of the COC specifications selected by the
inspectors were completed as required and the licensee was meeting the
conditions specified.
Procedures Controlling the Handling of Spent Fuel Shipments
The inspectors reviewed the licensee's "Spent Nuclear Fuel Shipping Program
Manual" (Plan) Revision 10, dated July 22, 1996, which discussed the Concept
of Operations, Organization and Responsibilities, and Training.
The inspectors reviewed the licensee's Spent Fuel Shipping Procedure
SFS-001, IF-300 Shipping Cask Operations Revision 13, dated July 30, 1996,
which discussed the spent fuel shipment process for receiving and inspection
of the spent fuel container railcars, relocating the cask to the decontamination
building, loading the fuel into the cask, transferring the cask to the
decontamination building, filling the cask with inert gas, and loading the cask
back onto the rail car. The inspectors selected Section 8.20 through 8.23 for a
detailed review. In the review, the inspectors noted that:
In SFS-001, a necessary procedural step to open cask drain valve CD-1
after performing Step 8.2.1.14 was missing. A closer review of the
missing step in SFS-001, revealed that the step was in place in Revision
12, of SFS-001.
It was concluded that while revising SFS-001 Revision
12 after performing a table top review of the procedure, a word
processing error resulted in the step being deleted in Revision 13 of
SFS-001. The inspectors verified that the operators had actually
performed the step and that a procedure step deviation was
documented.
As written, SFS-001 did not appear to accomplish three purges with
inert gas as required in the COC. SFS-001 step 8.21.15 stated that
"When helium exhausts from the drain hose, close the cask fill/drain
valve CD-i". The procedure proceeded to clearly require two distinct
purges. The licensee stated that although not proceduralized or
documented, that during this step, they allowed helium to flow through
the drain hose for approximately 10 to 15 minutes. In order to remove
24
any uncertainties concerning the adequacy of the initial purge, the
licensee performed an additional purge on each of the casks. The
licensee agreed that the procedure was not clear and that it would be
revised to clearly indicate three distinct purges.
The inspectors noted that an independent assessment of the licensee spent
fuel operating procedures was performed by VECTRA Technologies
Incorporated. The inspectors reviewed a letter dated May 23, 1996, from
VECTRA that stated VECTRA had reviewed the licensee's operating
procedures referenced as conditions of approval in the COC and determined
that the criticality control provisions were acceptable.
The inspectors concluded from the review that:
The licensee's Plan was organized and satisfactorily defined roles and
responsibilities in the fuel shipment,
Procedures were in place to maintain the cask, and
Procedures were in place to receive the cask, load spent fuel into the
cask, and ready the cask for shipment.
Radiological Surveys for Shipment
The inspectors reviewed licensee's radiological surveys to verify that the
licensee adequately decontaminated the spent fuel shipping cask to meet the
radiological requirement for transportation specified in 49 CFR 173.441.
The inspectors observed the licensee morning Health Physics briefings of
Radiation Work Permit (RWP) 96-0185 which was used to perform work on the
spent fuel cask. The briefings were detailed and informative. The briefings
updated personnel on the status of the cask decontamination efforts,
radiological conditions, clothing requirements for the area, and where they
were in the procedure. The inspectors accompanied the licensee into the work
areas and observed the licensee Health Physics practices around the cask
decontamination area and the railcar.
The inspectors observed the licensee decontaminate cask IF-304 using high
pressure spray, cleaning solvents, and scouring pads, in the cask
decontamination area. When the radiological surveys indicated that the
surface contamination was below the licensee's limits of 1000 disintegrations
per minute (dpm)/100 centimeter (cm) square, the licensee used procedure
SFS-001, IF-300 Shipping Cask Operations Revision 13 to transfer cask IF-304
from the decontamination area to the railcar.
25
The inspectors observed the licensee load the cask onto the railcar, conduct
radiation surveys around the railcar, perform gamma and neutron surveys
around cask IF-304, and perform contamination surveys (swipes) of the cask
and count the swipes in the lab. Once the surveys were completed, the
inspectors reviewed the licensee's survey sheets. This gamma/neutron survey
also satisfied one of the specifications of the COC discussed above.
After IF-304 cask was loaded onto the railcar, the inspectors conducted an
independent survey of the cask to determine if the radiation and contamination
levels were below the limits in 49 CFR 173.441. The inspectors determined
that the contact readings on the surface of the cask ranged from 2
millirem/hour (mRem/hr) to an isolated area that read 36 mRem/hr and, at 2
meters, radiation levels were less than or equal to 3.5 mRem/hr. The
inspectors also performed an independent surface contamination survey by
performing swipes of ten areas of the cask and observing the licensee count
the swipes. Most of the swipes averaged approximately 100 dpm/100 cm
square. All of the swipes taken by the inspector were less than the licensee's
limit of 1000 dpm/100 cm square.
The inspectors concluded that radiation and contamination levels were below
the transportation limits for shipments contained in 49 CFR 173.441 of 10
mRem/hr at 2 meters, 200 mRem/hr on contact, and less than of 2200
dpm/100 cm square loose surface contamination.
c.
Conclusions
The inspectors concluded effective procedures were implemented to safely
control irradiated fuel shipment activities. Minor spent fuel cask handling
procedure discrepancies identified were appropriately resolved by the licensee.
A NAS assessment of the spent fuel shipment program readiness was
thorough and probing. NAS assessment and table top exercise weaknesses
were properly resolved and effective program enhancements were
implemented prior to conducting shipment activities. The inspectors observed
that adequate controls were implemented during the August 12, 1996, spent
fuel shipment.
R1.3
Inadequate Labeling of Spent Fuel Cask Container
a.
Inspection Scope (71750)
While performing routine inspection activities in the RCA, the inspectors
determined that a loaded spent fuel shipping cask container did not have a
radioactive material label attached. The licensee initiated CR 96-01867 to
address this discrepancy. .
b.
Observations and Findings
On August 12, the inspectors observed the radiological controls for storing two
spent fuel shipping cask containers, loaded on separate railcars, inside the
RCA. The casks had recently been loaded with spent fuel and were awaiting
shipment from the site. The inspectors noted that radiological rope and
posting had been setup around both containers that housed each cask,
however, radioactive material labels were not attached to one of the cask
containers. The inspectors recalled during previous observations over the past
week, that the container had been properly labeled. After notifying RC
personnel of the potential problem, the cask container was surveyed and the
appropriate labels were affixed. The licensee initiated CR 96-01867 to address
this discrepancy.
10 CFR 20.1904, Labeling Containers, requires that containers of licensed
material be labeled with the words "CAUTION, RADIOACTIVE MATERIAL" or
"DANGER, RADIOACTIVE MATERIAL," and provide information regarding the
radiation levels and date the measurement was made. This information is
necessary to alert personnel working in the vicinity of the containers to take
precautions to avoid or minimize exposures. 10 CFR 20.1905 provides certain
exemptions from labeling containers. One of these exemptions include the
case where containers are in transport and the railcars carrying them are
placarded in accordance with the Department of Transportation regulations in
49 CFR 172. The inspectors determined that when the missing label was
identified, the licensee had not yet properly placarded the railcars in
accordance with 49 CFR 172, therefore, the labeling requirements of 10 CFR
20 were still applicable.
The inspectors reviewed the licensee's procedures for controlling the
radiological labeling requirements for containers with radioactive material in
excess of the limits established by Appendix C to 10 CFR 20. This included a
review of the following procedures:
Health Physics Procedure HPP-007, Handling and Storage of
Contaminated and Radioactive Material, Revision 18, and,
HPP-255, Shipping and Receiving the IF-300 Cask, Revision 10.
The inspectors determined that the procedures provided adequate guidance
and expectations for conforming to the requirements of 10 CFR 20.1904,
20.1905, and 49 CFR 172, for labeling containers. Based on discussions with
the licensee, they believed that the label had been removed by RC personnel
on August 11, when the cask was removed from its container and
decontaminated. In accordance with 10 CFR 20, a label is not required to be
27
affixed to the container when the cask is not loaded. Apparently, when the
cask was returned that same day, a label was not re-affixed to the container.
As corrective action, the licensee planned to revise the cask receipt checklist
contained in HPP-255 to include requirements and signoffs that a radiation
label be affixed to the spent fuel shipping container upon receipt and removed
only once the cask is accepted for shipment by the shipping carrier. The
inspectors concluded that this procedure enhancement would provide more
positive labeling controls and should prevent recurrence of this problem.
c.
Conclusions
The inspectors concluded this issue to be a violation of 10 CFR 20.1904 for
failure to label a container of licensed radioactive material in excess of
quantities listed in Appendix C to 10 CFR 20. This failure constitutes a
violation of minor significance and is being treated as a Non-Cited Violation,
consistent with Section IV of the NRC Enforcement Policy. This item will be
identified as NCV 50-261/96-10-02: Failure to Label Spent Fuel Cask
Container in Accordance with 10 CFR 20.
R2
Status of Radiation Protection Controls and Equipment
R2.1
Failure of Radiological Information Management System (RIMS)
a.
Inspection Scope (71750)
The inspectors reviewed the results of the licensee's investigation of loss of
RIMS. Two CRs and exposure estimates of affected individuals were
reviewed.
b.
Observations and Findings
The licensee has an electronic dosimetry system (EDS) and the EDS work
stations for all three sites are connected to a centralized computer. The dose
for each individual leaving the radiological control area is down loaded into the
RIMS database.
On June 8, 1996, RIMS entered a scheduled 33 hour3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> outage to make software
changes. The licensee took steps not to affect the Access Control software.
The system worked properly when an individual logged in. When the individual
logged out, the local work station indicated that a normal transaction took
place. However, the data was not loaded into the database. The licensee
discovered the problem when it was observed that there was excessive
downtime with PC Access Control. Two CRs were written. CR 96-01481 was
)written
by the site and CR 96-1612 was written by Corporate Radiological
Services.
The site E&RC organization obtained security records to determine who had
entered the RCA and the duration of their stay. The licensee determined that
39 individuals entered the RCA. Thermoluminescent dosimeters (TLDs) of
those individuals were pulled and read to obtain a conservative estimate of the
exposure dose received.
Exposure estimates were made for made for those
individuals without TLDs. The licensee was conservative in their estimates.
All individuals were interviewed and all reviewed and signed their estimated
dose. Thirty-eight mRem was the maximum dose assigned. The inspectors
observed and reviewed the licensee's investigation of the CR.
c.
Conclusions
The inspectors determined that the licensee's investigations and corrective
actions for CR 96-1612 were adequate. The investigation revealed that PC
Access Control software system worked as designed except that the Access
Control Recovery Screen appeared to accept data but did not. The inspectors
considered this incident an isolated occurrence requiring no additional
corrective action.
S1
Conduct of Security and Safeguards Activities (71750 and 81310)
S1.1
Inadequate Lighting of Spent Fuel Cask Rail Cars in the Protected Area
a.
Inspection Scope (71750)
The inspectors observed during a tour of the protected area that the lighting
under two rail cars inside the protected area was not adequate. The
inspectors discussed the discrepancy with security personnel and reviewed the
licensee's Industrial Security Plan and security procedures with regard to
protected area minimum lighting requirements.
b.
Observations and Findings
On July 25, the inspectors observed that extra lighting installed to illuminate
the space under two rail cars that were temporarily located inside the protected
area was inadequate. A string of incandescent light bulbs had been placed on
the outside of one of the cars and a single Halogen lamp was placed on the
ground near the other car. The inspectors observed that one of the
incandescent bulbs and the Halogen lamp had failed. The inspectors notified
the licensee of the lighting discrepancies and questioned whether minimum
lighting illumination under the cars was met under the conditions observed.
29
The licensee later measured the lighting levels under the rail cars and found it
to less than the required 0.2 foot-candles. However, the licensee believed that
the area was adequately backlighted.
The licensee performed an inspection within the protected area during the
evening of July 25 and identified four additional areas which had inadequate
lighting. Two paint sheds, a trailer, and an air compressor were identified as
requiring additional lights or bulb replacement. Corrective action was
completed the next day. CR 96-01731 was issued to document the lighting
discrepancies.
The inspectors later became aware that on July 12, a NAS individual had
identified to security personnel a lighting level concern for another trailer
located inside the protected area. The on-shift security staff failed to followup
on the concern indicating a lack of sensitivity to the lighting requirements.
10 CFR 73.46(c)(4) and 73.55(c)(5) requires that all exterior areas within the
protected area be illuminated to at least 0.2 foot candles measured horizontally
at ground level. In addition, Section 3.1.3 of the licensee's Industrial Security
Plan, Revision 32, dated April 26, 1996, states, in part, "the exterior protected
area will be lighted to a level sufficient for monitoring, surveillance, and
observation requirements, but not less than 0.2 foot-candles measured
horizontally at ground level. Compensatory measures for degraded illumination
(less than 0.2 foot-candles) in exterior portions of the protected area will be in
the form of increased visual surveillance." The inspectors reviewed the
Security surveillance sheets for July 1996. No additional surveillances were
logged for the rail cars indicating that the discrepant conditions had not been
identified.
c.
Conclusions
The installation of security lighting under the rail cars was inadequate, and
routine patrols of the area failed to identify this condition and correct the
deficiencies or implement compensatory measures. The failure to meet the
illumination level of at least 0.2 foot-candles or implement compensatory
measures for the degraded illumination conditions was identified as Violation
50-261/96-10-03: Failure To Follow Security Plan for Minimum Lighting
Requirements.
30
S1.2 Security Controls of Spent Fuel Shipments
a.
Inspection Scope (81310)
The inspectors reviewed the licensee's compliance with 10 CFR 73.37(f) with
regard to advance notification of irradiated fuel shipment, protection of
Safeguards shipment information, and security controls established for
irradiated fuel shipments from the site.
b.
Observations and Findings
By letter dated July 30, 1996, to the NRC the licensee complied with the prior
notification requirements of 10 CFR Part 73.37(f), by providing 10 days
advance notice of a shipment of irradiated fuel. This letter was also furnished
to the designated representatives of the Governors of South and North
Carolina, thus meeting the requirement to provide the states with 7 days
advance notice. The requirements of 10 CFR Part 73.21 were met in that the
licensee stamped as "Safeguards Information" those portions of the letter
which revealed dates, times and routes of the actual shipment.
Throughout this inspection, the licensee's efforts to protect Safeguards
Information from unauthorized disclosure was evident at all levels of
involvement.
Also noted was the compensatory measure utilized at the Robinson perimeter
when the site vehicle barrier was removed to allow the opening of the railroad
gate. An officer was continuously posted who was armed with a contingency
high-power rifle.
The inspectors found that the 3 Escorts were assigned to this shipment were
knowledgeable of their duties and responsibilities. They were also very
familiar with Emergency Procedures and the content of the Emergency Kit
located in the caboose of the train. The Senior Escort, a trained health
physicist from the Harris Nuclear Plant, explained to the inspectors the function
of the three radiation detectors found in this Kit as well as other contingency
equipment located therein. The multi-means of communication from the
locomotive and the caboose were also demonstrated to the inspectors, a
review of logs revealed that prior to the arrival of the CSX locomotive engine
the licensee had verified all telephone numbers, radio frequencies and cellular
capabilities for state, county and local law enforcement agencies along the
route of this shipment.
The inspectors learned that these Escorts were aware of the guidance found in
the following licensee procedures:
31
HPP-256, Advance Notification For Shipments
SEC-2120, Protection of Safeguards Information
HPP-255, Shipping of IF 300 Cask
NGG-006, Spent Fuel Manual
RSP-1.1, Duties of Shipment Escorts
SEP-2.1, Shipment Emergency Duties
On August 12, at 8:28 p.m., the train left the Robinson site and was
periodically monitored by the inspectors throughout the night until it arrived at
the Harris site at 3:45 a.m. the next day. Upon arriving at the Harris facility the
inspectors reviewed the licensee's record of communication checks and
determined that the required checks were accomplished as required every 90
minutes.
c.
Conclusions
The inspectors concluded that the licensee's program for shipping irradiated
fuel was found to be in compliance with 10 CFR Part 73.37. No discrepancies
were identified with the security controls for the spent fuel shipment conducted
on August 12.
V. Management Meetings
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management
at the conclusion of the inspection on August 26, 1996. An interim exit was
conducted on August 7, 9, and 12, 1996. The licensee acknowledged the findings
presented.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary. No proprietary information was
identified.
32
PARTIAL LIST OF PERSONS CONTACTED
Licensee
J. Clements, Manager, Site Support Services
D. Crook, Senior Specialist, Licensing/Regulatory Compliance
C. Hinnant, Vice President, Robinson Nuclear Plant
J. Keenan, Director, Site Operations
R. Krich, Manager, Regulatory Affairs
B. Meyer, Manager, Operations
G. Miller, Manager, Robinson Engineering Support Services
R. Moore, Manager, Outage Management
J. Moyer, Manager, Maintenance
D. Stoddard, Manager, Operating Experience Assessment
R. Warden, Acting Manager, Nuclear Assessment Section
T. Wilkerson, Manager, Environmental Control
D. Young, General Manager, Robinson Plant
NRC
P. Byron, Resident Inspector, Brunswick
J. Zeiler, Acting Senior Resident Inspector
33
INSPECTION PROCEDURES USED
IP 37550:
Engineering
IP 37551:
Onsite Engineering
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and
Preventing Problems
IP 60855:
Operation of an Independent Spent Fuel Storage Installation (ISFSI)
IP 61726:
Surveillance Observations
IP 62703:
Maintenance Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 81310:
Physical Protection of Shipments of Irradiated Fuel
IP 86750:
Solid Radioactive Waste Management and Transportation of Radioactive
Materials
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Type Item Number
Status
Description and Reference
50-261/96-10-01
Open
Inadequate Corrective Actions for SOV
Discrepancies (Section E8.1)
NCV 50-261/96-10-02
Open
Failure to Label Spent Fuel Cask Container in
Accordance with 10 CFR 20 (Section R1.3)
50-261/96-10-03
Open
Failure To Follow Security Plan for Minimum
Lighting Requirements (Section S1.1)
Closed
Type Item Number
Status
Description and Reference
LER
50-261/93-015-00
Closed
Pressurizer Pressure Transmitters Out of
Calibration (Section M8.1)
LER
50-261/94-003-00
Closed
TS Required Shutdown Due to Emergency
Diesel Generator Inoperability (Section M8.2)
LER
50-261/94-011-00
Closed
Technical Specification 3.0: Emergency
Diesel Generator Inoperability (Section M8.3)
34
LER
50-261/94-019-01
Closed
TS Violation Due to Exceeding Pressurizer
Cooldown Rate (Section M8.4)
50-261/96-08-01
Closed
Review Licensee Investigation and Resolution
of Solenoid Valve Discrepancies (Section
E8.1)
50-261/94-24-01
Closed
Inadequate Testing of Alternate AC Power
Source (Section E8.2)
50-261/94-16-04
Closed
Inadequate Control Room Ventilation Testing
Program (Section E8.4)
LER
50-261/94-008-01
Closed
Condition Outside Design Basis Due to
Control Room HVAC Inoperability (Section
E8.5)
Discussed
Type Item Number
Status
Description and Reference
IFI
50-261/94-08-02
Open
Incorporate 24 Hr Load Testing into TS
Surveillance Requirements (Section E8.3)