ML20216H443
| ML20216H443 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/16/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20216H421 | List: |
| References | |
| 50-324-98-04, 50-324-98-4, 50-325-98-04, 50-325-98-4, NUDOCS 9804210139 | |
| Download: ML20216H443 (26) | |
See also: IR 05000324/1998004
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION II
Docket Nos: 50-325, 50-324
Report No: 50-325/98-04. 50-324/98-04
Licensee: Carolina Power & Light (CP&L)
Facility: Brunswick Steam Electric Plant Units 1 & 2
Location: 8470 River Road. SE
Southport. NC 28461
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Dates: February 9 - 13. March 9 - 12. and
March 16 - 20. 1998
Inspectors: J. Lenahan, Reactor Inspector
G. MacDonald. Project Engineer
Approved by: K. Landis. Chief
Engineering Branch
Division of Reactor Safety
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9804210139 980416
PDR ADOCK 05000324
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EXECUTIVE SUMMARY
Brunswick Steam Electric Plant. Units 1 & 2
NRC Inspection Report 50-325/98-04. 50-324/98-04
This special inspection was conducted to followup on the licensee's corrective
i actions for electrical equipment environmental qualification and a review of I
instrument response time testing.
Results:
One violation and four apparent violations were identified:
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An apparent violation for an inadequate 10 CFR 50.59 evaluation which i
resulted in deleting requirements for performance of response time j
testing of instrumentation (Paragraph M1.1).
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An apparent violation for failure to perform RTT of reactor protection ;
system instrumentation required by Technical Specification 4.3.1.3 I
(Paragraph M1.2).
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An apparent violation for failure to perform RTT for actuation of the
primary and secondary containment isolation systems required 'oy TS
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4.3.2.3 (Paragraph M1.2).
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An apparent violation for failure to perform RTT for actuation of
the emergency core cooling system as required by TS 4.3.3.3
(Paragraph M1.2).
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A violation for use of an ESR. which was not design verified (ED ESR).
to change plant design documents (Paragraph E.8.3).
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REPORT DETAILS
II. Maintenance
M1 Conduct of Maintenance i
M1.1 Instrument Resoonse Time Testina i
a. Insoection Scoce
The inspector reviewed the 10 CFR 50.59 safety evaluations which
were performed to delete performance of instrument response time
testing required by Technical Specification 3/4.1.3. 3/4.2.3. and
3/4.3.3.
b. Observations and Findinas
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Generic Letter (GL) 93-05. Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing
During Power Operations, was issued by NRC on September 27. 1993.
This GL recommended deletion of isolation instrumentation response
time testing (RTT) requirements for functions where the required
response time corresponded to the diesel generator start time.
The GL specifically stated that a Technical Specification (TS) l
amendment was required to delete these response time testing
requirements from the TS.
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On December 29, 1993. NRC issued GL 93-08. Relocation of Technical
Specification Tables of Instrument Response Time Limits. The GL
encouraged licensees to submit a TS Amendment request to relocate
these tables to the Updated Final Safety Analysis Report (UFSAR)
and then control changes to the tables using the 10 CFR 50.59
process. The licensee requested an amendment to remove Tables
3.3.1.2. 3.3.2.3 and 3.3.3 from the Technical Specifications and j
to relocate the tables to the UFSAR. This was approved by NRC in
TS Amendments 171 and 202 which were issued on May 31. 1994. The ,
licensee incorporated the Tables into the UFSAR as change number
31 which was issued by the licensee on June 17. 1994.
On October 3.1994. the licensee issued UFSAR change log number 4
94FSAR056 which deleted the isolation response time testing
requirements for instrumentation with the required response time
corresponded to the diesel generator start time (less than 13
seconds). This change was made in accordance with the
recommendations contained in GL 93-05. However a TS Amendment was
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not requested by the licensee or approved by the NRC prior to ,
i elimination of the RTT requirements for these instruments. The
licensee revised the implementing plant procedures after the UFSAR
i change was issued. The UFSAR change was not submitted to NRC
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prior to elimination of RTT for the instrumentation covered by GL 93-05.
-Instrumentation affected by this change is listed in Table 1, below.
Table 1
GL 93-05 Instrumentation *
1. Primary Containment Isolation
1. Low, level 1
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b. Drywell Pressure - High
c. Main Steam Line l
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1. Radiation - High
2. Pressure - Low
3. Flow - High
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4. Flow - High (Unit 2 only)
d. Main Steam Line Temperature - High
e. Condenser Vacumm - Low
2. Secondary containm3nt Isolation
a. Reactor Building Exhaust Radiation - High
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b. Drywell Pressure - High
c. Reactor Vessel Water Level - Low. Level 2 l
3. Reactor Water Cleanuo System Isolation
e. Reactor Vessel Water Level - Low. Level 2
4. Core Standby Coolina Systems Isolation
a. High Pressure Coolant Injection System Isolation
1. HPCI Steam Line Flow - High
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.,3. HPCI Steam Supply Pressure - Low -
b. Reactor Core Isolation. Cooling System Isolation
1. RCIC Steam Line Flow - High
- Instrumentation with response times less than the diesel
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generator start time (13 seconds).
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NOTE: Numerical / letter identifications of instrumentation
L shown above corresponds with that shown in former TS
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and current UFSAR Table 7.3.1-3A
The BWR Owners Group submitted a report titled: BWR Owners Group
l Licensing Topical Report NED0-32291. " System Analyses for
! Elimination of Selected Response Time Testing Requirements."
i January 1994, in a letter to NRC dated January 14. 1994. The
report concluded that instrument calibration of selected
instruments would provide the data to detect degradation of RTT.
i and that RTT could be eliminated from surveillance requirements.
l These instruments were not included in GL 93-05. The NRC issued a
l- Safety Evaluation Report (SER) in a letter to the BWR Owners Group- ,
~ dated December 28, 1994. The SER approved the NEDO 32291 Report .
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and provided guidance to licensees on the information required to
be submitted with TS change requests to implement the NED0-32291
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recommendations. On February 14. 1995, the licensee initiated
UFSAR change log number 94FSAR100 to implement the recommendations
of NEDO-32291. The implementing plant procedures _were revised
after the UFSAR change was issued by the licensee and the RTT of
-instrumentation covered by NED0-32291 was deleted. The UFSAR
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change was not submitted to NRC prior to elimination of the RTT.
l Instrumentation affected by this change is listed in Table 2.
below. j
Table 2
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l Instrumentation Resoonse Time Deleted oer NEDO-32291
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<. UFSAR Table 7.2.1-3
! Reactor Protection System Instrumentation Resoonse Times
3. Reactor Vessel Steam Dome Pressure High s 0.55-
'4. Reactor Vessel Water Level - Low. Level 1 s 0.09
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l UFSAR Table 7.3.1-3A l
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Isolation System Instrumentation Resoonse Time l
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1. Primary Containment Isolation
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2. Low, Level 3 s 1.0
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3. Flow - High s 0.5
4. Flow - High s 0.5
UFSAR Table 7.3.3-5
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Emeroency Core Coolina System Resoonse Times
1. Core Spray System s 27 seconds ;
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2. LPCI Mode of RHR System s 40 seconds
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3. High Pressure Collant Injection System s 60 seconds
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On May 16, 1996. the licensee initiated UFSAR change log number i
l 96FSAR015. This change updated the UFSAR Table 7.3.1-3A by '
changing the RTT for the primary containment isolation instrument i
number 1.g. Main Stack Radiation - High function to N/A. The 10 l
CFR 50.59 evaluation for this change was based on UFSAR change log
Number 94FSAR100. The basis for the change was that licensee j
engineers concluded that this RTT for this instrument should have '
been deleted under 94FSAR100, since it was covered by NE00-32291.
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On March 21, 1997, the licensee notified NRC of a potential l
noncompliance with the Technical Specifications (TSs) for l
inappropriately deleting instrument response time testing
I surveillance requirements which were required by TS Sections i
4.3.1.3. 4.3.2.3. and 4.3.3.3. The licensee requested enforcement !
discretion to permit continued operation of the Brunswick plant
until the issue could be resolved. The NRC exercised discretion
not to enforce compliance with the applicable TS sections for a
period of 30 days based on an evaluation that granting the request l
involved minimal or no safety impact on the public health and
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The inspectors reviewed the actions surrounding the deletion of
the instrument response time surveillance testing. Review of the
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10 CFR 50.59 Safety Reviews for UFSAR changes log numbers
94FSAR056, 94FSAR100 and 96FSAR015 disclosed that the licensee's
reviewers incorrectly concluded that the proposed UFSAR changes j
did not involve a change to the Technical Specifications. The
licensee's reviewers incorrectly concluded that these UFSAR
changes only affected the Tables relocated from the TS to the
UFSAR per the recommendations of GL 93-08 approved by NRC in TS
Amendments 171 and 202. 10 CFR 50.59 permits licensees to make
changes in the facility or procedures without prior NRC approval
unless the proposed change involves an unreviewed safety question
or a change in the technical specifications. The elimination of
RTT did not involve an unreviewed safety question. These changes
were recommended and approved by NRC in GL 93-05 and the SER dated
December 28, 1994. However both documents clearly stated that the
RTT could not be deleted unless a TS change was requested and
approved by NRC. The requirements for RTT were specified in TS Sections 4.3.1.3. 4.3.2.3 and 4.3.3.3. Deletion of the RTT
constituted an unapproved change to the plant TSs. without prior
NRC approval. This was identified to the licensee as Apparent
Violation Item 50-325(324)/98-04-01. Inadequate 50.59 Evaluations
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which Resulted in Deletion of TS Response Time Testing
Requirements.
Conclusions
An apparent violation was identified for deleting the RTT
specified in the TS from the UFSAR and surveillance procedures due
to an inadequate 50.59 safety review. This issue did not involve
an unreviewed safety issue.
M1.2 Performance of Resoonse Time Testina
a. Insoection Scone (61700)
The inspectors reviewed response time testing performed by the licensee !
for conformance with Technical Specifications 4.3.1.3. 4.3.2.3. and I
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4.3.3.3. '
b. Observations and Findinas
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TS 4.3.1.3. 4.3.2.3, and 4.3.3.3 required demonstration that the
RTT was within its limit at least once per 18 months. TS 4.02
permits extension of the test interval by plus or minus 25
percent. Therefore, the RTT was required to be performed at a
maximum interval of 687 days (18 months plus 25%). The inspectors
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reviewed the results RTT performed on instrumentation listed in
Tables 1 and 2. above. The results of last RTT performed which
complied with TS are shown in Attachments 1 for Unit 1 and
Attachment 2 for Unit 2. The inspectors determined that as of
March 21. 1997, the licensee failed to perform RTT within the TS
required interval for the instrumentation listed in Table 2 above.
Enforcement discretion was granted for a period of thirty days
which permitted plant operation pending the issuance of a TS
amendment which approved deletion of RTT. Further review by the
inspectors disclosed that the licensee failed to request
enforcement discretion for the instrumentation listed in Table 1.
above. However approval to delete the RTT for this
instrumentation was also included in TS Amendments 184 and 215
issued on April 18, 1997.
Further review by the licensee disclosed that the RTT for the main
stack radiation monitor deleted by UFSAR change log Number
96FSAR015 did not meet the requirements of NED0-32291. The
licensee initiated Condition Report (CR) 97-02806 on August 19.
1997, to document and disposition this problem. The
instrumentation was declared inoperable. The licensee then
performed the required RTT on August 22, 1997, for Unit 2 and on l
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August 23. 1997, for Unit 1. The inspectors reviewed the results
of the RTT which were documented in WR/JO 97-AFBT1 for Unit 1 and
WR/JO 97-AQUB for Unit 2.
The following apparent violations were identified by the '
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inspectors for failure to perform RTT as required by the
applicable Technical Specifications (Amendment 175 - Unit 1 and
Amendment 206 - Unit 2):
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Apparent violation item 50-325 (324)/98-04-02. Failure to
Perform RTT for Reactor Protection System Instrumentation as
Required by TS 4.3.1.3.
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Apparent violation item 50-325 (324)/98-04-03. Failure to
Perform RTT for Instrumentation for Actuation of the Primary
and Secondary Containment Isolation Systems as Required by
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Apparent violation item 50-325 (324)/98-04-04 Failure to
Perform RTT for Instrumentation for Actuation of ECCS as
Required by TS 4.3.3.3.
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c. Canclusions
The inspectors concluded that the licensee failed to perform response
time testing in accordance with Technical Specifications 4.3.1.3.
4.3.2.3, and 4.3.3.3. The failure to perform RTT was identified as
Apparent Violations (50-325. 324/9804-02. -03, and -04).
M1.3 Imolementation of BWROG Licensina Tooical Reoort. NED0-3229
a. Insoection Scooe (62700. 61700)
The inspectors reviewed the licensee's implementation of NED0-32291 for
conformance with the NRC's December 28, 1994 Safety Evaluation Report
(SER).
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b. Observations and Findinas
The December 28, 1994 SER that accepted NEDO-32291 required that
licensees confirm the implementation of seven conditions as part of any
license amendment application to eliminate response time testing. The
inspectors reviewed the licensee's conformance with the seven
conditions. Review of licensee records and discussions with licensee
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engineers disclosed that the licensee documented compliance with the
seven conditions in response to a finding in a Nuclear Assessment
Section (NAS) review and in ESR 9700205. The NAS assessment of the !
Technical Specifications and Operating License performed from June 3 to
14. 1996. identified an issue regarding documentation of NRC required
actions for deletion of RTT. The results of the self assessment are
documented in Assessment Report number B-OL-96-01 dated July 3. 1996. l
The issue, documented in CR96-01804, concerned the fact that RTT was i
deleted from some procedures without documentation that required actions )
were being performed. ESR 9700205. Response Time Test Elimination. '
Revision 0 dated March 21, 1997. documented the review performed by
licensee engineers of the seven conditions in the December 28. 1994 SER.
The seven conditions and the licensee's conformance with those
conditions follow:
(1) Condition
Prior to installation of a new transmitter / switch or following
refurbishment of a transmitter / switch (e.g., sensor cell or
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variable damping components). hydraulic response time testing
shall be performed to determine an initial sensor-specific
j response time value.
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The licensee determined, based on an engineering review, that
revised procedures and an enhanced trending program will provide
sufficient means of identifying suspect transmitters.
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(2) Condition
For transmitters and switches that use capillary tubes, capillary
tube testing shall be performed after initial installation and
after any maintenance or modification activity that could damage
the capillary tubes.
Response
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The inspectors found that instruments installed at Brunswick do
not use capillary tubes.
(3) Condition
Calibration shall be performed with equipment designed to provide
a step function or fast ramp in the process variable.
, Response
The licensee's test equipment was designed to provide a step
function or fast ramp in the process variable. Test procedures
were revised to provide a fast ramp change of r.ransmitter input
while monitoring the transmitter output.
(4) Condition
Provisions shall be made to ensure that operators and technicians,
through an appropriate training program, are aware of the
consequences of instrument response time degradation, and
applicable procedures have been reviewed and revised, as
necessary, to assure that technicians monitor for. response time
degradation during the performance of calibrations and functional
tests.
Response
l The inspectors reviewed Lesson Plan IC7C078. Rosemount Transmitter
(Loss of Fill Oil), dated May 30. 1995. The consequences of
instrument response time degradation (sluggish response) were
discussed in this training. The inspectors reviewed the
procedures listed below and verified that I&C personnel verify
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monitor response time degradation. This is performed by verifying
instantaneous output response to input. Procedures reviewed were:
2MST-HPCI21R. 2MST-RHR22R. and 2MST-RPS23R. This requirement was
also covered during quarterly training.
(5) Condition
Surveillance testing procedures shall be reviewed and revised if
necessary to ensure calibrations and functional tests are
performed in a manner that allows simultaneous monitoring of both
the input and output response of instruments under test.
Response
In review of the three procedures listed above. the inspectors
verified that testing was performed in manner that provides for i
simultaneous monitoring of instrument input and output.
(6) Condition !
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For any request involving the elimination of response time testing
for Rosemount pressure transmitters, the licensee shall be in
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compliance with the guidelines of Supplement 1 to NRC Bulletin 90- ,
01. " Loss of Fill-Oil in Transmitters Manufactured by Rosemount." !
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Response
The inspectors determined that the licensee was in compliance with
the guidelines of Supplement 1 to NRC Bulletin 90-01. These j
requirements were reviewed in detail as part of training performed
under lesson plan IC7078.
(7) Condition
For those instruments where the manufacturer recommends periodic
response time testing as well as calibration to ensure correct
functioning, the licensee shall ensure that elimination of
response time testing is nevertheless acceptable for the
particular application involved.
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Response
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The licensee revieweo vendor information for NED0-32291 components
and determined that there were no specific recommendations to
perform periodic response time testing.
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The inspectors also reviewed the types of instruments, components
and affected systems for which the response time testing was
eliminated using NED0-32291 and verified that the licensee
l complied with Tables 1 and 2 of the SER. The following
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instruments were reviewed through walkdown inspections and/or
review of equipment data base records:
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Instrument Numbers B21-LTS-N024A-1-2, and B -l-2. and B21-
l LTS-N025A-1-2 and B-1-2 for reactor vessel water level. Low,
i Level 3. These instruments were Rosemount transmitters.
Associated relays were Agastat GP/EGP family.
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Instrument Numbers B21-PDT-N006A. B. C. D and B21-PDT-008A.
B. C. D for main steam line high flow isolation. These
instruments were Rosemount transmitters. Associated relays
were Agastat GP/EGP family.
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Instrument Numbers B21-PT-N023A. B. C. D for reactor
protection system high reactor pressure. These instruments
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were Rosemount transmitters. Associated relays were Agastat
GP/EGP family.
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Instrument Numbers B21-LT-N017A-1. B-1. C-1. 0-1 for reactor
protection system reactor vessel water level. Low. Level 1.
These instruments were Rosemount transmitters. Associated
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relays were Agastat GP/EGP family.
i c. Conclusions
The inspectors concluded that the instrumentation for which RTT was
deleted per NED0-32291 complied with the conditions listed in the
December 28, 1994. SER.
III. Enaineerina
E.1 Conduct of Engineering
E1.1 Environmental Qualification (92903)
a. Insoection Scone
The inspectors reviewed the licensee's corrective actions for the
Environmental Qualification (EO) program. in response to findings
identified during Self-Assessment Numbers 95-0041 and 96-0271 and
the violations identified in NRC Inspection Report 50-325. 324/96-
14. The licensee's corrective actions to resolve deficiencies in
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the E0 program include revision and updating of Qualification Data
Packages (ODPs). The process for incorporating E0 program .
maintenance and replacement requirements into site procedures was
reviewed.
b. Observations and Findinas
The inspectors reviewed CP&L Procedure EGR-NGGC-0156. Environmental
Qualification of Electric Equipment Important to Safety. Revision 5. ,
dated February 27. 1998. which specifies the requirements for
preparation of the ODPs. Attachment 2 to the procedure contains the
specific requirements for preparation of the Brunswick ODP files,
including qualifications for E0 personnel responsible for preparation of
the ODPs. ODP format and content, qualification analysis. E0 equipment
data (summarized on system component evaluation worksheets).
qualification parameters, equipment test data, and maintenance
requirements. l
The inspectors reviewed the following ODPs:
ODP - 14. Westinghouse Electrical Penetrations. Class B. C. E. and F.
Revision 3. dated January 30, 1998.
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ODP - 16. Okonite Tape Splices. Revision 3. dated March 6. 1998.
00P - 61. Honeywell PTK and PTS Series Control Switches. Revision 2.
dated February 23, 1998.
ODP - 99. R. G. Laurence Series 500 and 600 Solenoid Valves. Revision
0 dated December 23. 1997.
The inspectors verified the ODPs addressed the following: qualification
level (NUREG 0588 Cat. I): tag numbers of equipment covered in the ODP:
test report applicability; similarity of test specimens to installed
equipment: E0 parameters, temperature, pressure, relative humidity,
radiation, chemical spray, submergence: qualified life: E0 maintenance
requirements; test anomalies; and operating experience items. The
inspectors verified that similarity analysis was included in the ODPs. l
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For ODP-14. the inspectors reviewed Patel Test Report PEl-TR-83- !
14-4 which documents the design basis accident testing performed
by Westinghouse. The inspectors also reviewed penetration design
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drawings, responses to NRC Information Notices and Bulletins, and
referenced ESRs. ESR 9700674. titled Evaluate E0 Status of
Westinghouse Electrical Penetrations documents a review of the l
Westinghouse electrical penetrations to determine which ones are
required to be environmentally qualified. The criteria for
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l determination of which penetrations were required to be E0 l
qualified were those which perform an electrical function '
important to safety. The inspectors noted that the 4160 volt
electrical penetrations, which supply power to the recirculation
pumps. were not included in the E0 program. The 4160 volt
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electrical penetrations do not perform any electrical safety
related functions. -The inspectors determined from review of test
reports and the licensee's response to IE Bulletin 77-06.
Potential Problems with Containment Penetration Assemblies, that
, environmental qualification of the electrical penetrations was not
affected by the 15 psi nitrogen pressure normally maintained in
the drywell penetrations. The reason the pressure was maintained
in the penetrations was to monitor potential leakage. Any
detected leakage was evaluated under the requirements of 10 CFR
50. Appendix J.
ODP-16 and the following associated DRs were reviewed:
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DR 3.1-'2. Revision 5. dated August 5. 1997."0konite Report
No. NORN-3. Revision 4 Nuclear Environmental Qualification
Report for Okoguard Insulated Cables and T-95 & No. 35
Splicing Tapes. October 24. 1988".
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DR 3.1-2A. Revision 1. dated December 10. 1997."Patel
Engineers Assessment Report No. PEI-TR-82-4-41. Revision 0.
Final Assessment Report on Okonite Tape Splice Insulation
for Power and Control Cables and Rockbestos Pyrotrol III and
Firewall III Cables Used in the James A. Fitzpatrick Nuclear
Power Plant. December 5, 1994".
The' inspectors concluded that.the ODP documentation was complete.
that the ODP was prepared in accordance with the requirements of
procedure EGR-NGGC-156. that appropriate ESRs.were incorporated
into the 0DP, and that generic industry communications were
addressed in the ODP. The inspectors identified several minor
inconsistencies'with the ODP which were addressed by the licensee.
These items included some inconsistencies between the SCEW sheets
and equipment master list an incorrect reference, and some test
results which were not discussed in the CDP. The inconsistencies
did not affect the accuracy of the ODP. The inspectors reviewed
Procedure OSPP-CBL003. Disconnection and Reconnection of Taped
Splices for Electrical Equipment. Revision 19. dated September 5,
1997, and verified that the procedure complied with the
installation requirements of the referenced qualification test
reports.
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ODP-61 and associated Document Reference (DR) DR 59.1. titled Nuclear
Environment'al Qualification Report of Terminal Blocks. Limit Switches.
Control Switches. Indicating Lights, and Solenoid Valves for BSEP Units
1 and.2 was reviewed. The inspectors concluded that the ODP
documentation was complete and addressed all qualification issues. The
inspectors determined that ODP was prepared in accordance with the
requirements of-Procedure EGR-NGGC-156. that appropriate ESRs were -
incorporated into the ODP. and that generic industry communications
! related to the switches were addressed in the ODP. The inspectors
l identified several minor inconsistencies' with the ODP which were
, addressed by the licensee. These items included some discrepancies in
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the SCEW sheets and equipment data base system (EDBS), and a test
~ anomaly which was not addressed in the ODP. The minor discrepancies did
not affect the accuracy of the completed ODP.
j Review of ODP-99 showed that the valves used on the post accident'
sampling system met the requirements of 10 CFR 50.49. The
inspectors identified some minor editorial comments in the
completed ODP. but these did not affect the accuracy or
. conclusions of the ODP.
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The inspectors also reviewed the licensee's process for
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incorporating the maintenance and qualified life data into the
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plant procedures upon completion and approval of ODPs by the E0
group. The licensee used a checklist to control the ODP' approval
process and post ODP approval items. The checklist contained l
requirements that the following items be completed by an engineer i
and checked by a second individual to verify that the actions were !
, accomplished:
. EDBS changes
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E0 maintenance activities match Preventive Maintenance (PM)
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( . EQ Maintenance requirements are captured in appropriate
E procedures and/or installation specifications. l
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. Appropriate procurement specifications reflect E0 l
procurement requirements.
c. Conclusions
The ODPs were complete and addressed all environmental qualification
related issues and related industry communications. Minor
inconsistencies were identified with the ODPs which were addressed by
the licensee. The licensee has a controlled process to ensure that EQ
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maintenance., qualified life, and procurement requirements are
incorporated into existing site procedures.
El.2 Review of Environmental Oualification Condition Reoorts (92903)
a. Insoection Scooe
The inspecters reviewed condition reports (CRs) initiated to
document and disposition discrepancies involving environmental I
qualification issues.
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b. Observations and Findinas
The inspectors reviewed the corrective actions to disposition the
condition reports (CRs) listed below. These CRs were initiated by
the licensee to document and disposition nonconforming items which
were identified during the ongoing E0 reconstitution project. The
nonconforming items were identified as a result of E0 equipment
walkdowns, review and updating of E0 equipment qualification data
packages (00Ps) omissions from the original program, or changes
to the operating environment. The CRs reviewed were as follows:
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CR 97-02844 was initiated to address a discrepancies in the
maximum secondary containment temperatures. This was the result ,
of NRC questions raised during review of ODP - 67 during the I
inspection documented in NRC Inspection Report numbers 50-325.
324/97-09. The licensee revised their operating instructions to
require that operations personnel monitor temperatures in the
reactor buildings on a daily basis and report any occurrences when
the ambient temperatures exceeds 100 F to angineering (EO) for
evaluation.
CR 97-02902 was initiated by maintenance to document a potential I
problem with the licensee's supply inventory system which did not
clearly designate between 0 listed parts and E0 parts. The
licensee reviewed 1010 work requests and determined that no
unqualified parts had been installed in E0 applications. ~he
problem documented in the CR involved the difficultly of the
process that maintenance planners were required to use to
determine if the parts they specify are required to be EQ
qualified. The corrective actions were to generate an engineering
bulletin on the subject of CR 97-02902 for use by maintenance
planners and procurement engineers to clirify E0 requirements when
designating replacement parts. ,
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CR 98-00172 was initiated to document the existence of local " hot
spots" in some plant areas which may affect qualification of some
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E0 equipment. The licensee prepared ESR 9800032 to address these
areas and updated DR-227. Environmental Qualification Service
Conditions. Revision 3. to address evaluation of the local high
temperatures. Equipment qualified under nine 0DPs were affected
by hot spots. The licensee determined that three ODPs may require
revision to consider the affects of the hot spots, while the
remaining six were acceptable. The licensee will perform
temperature monitoring in these areas to determine if-additional
revisions to the 0DPS are necessary,
c. Conclusions
The inspectors concluded that the licensee was making progress in
resolving and closing CRs identified by the E0 group.
E2 Engineering Support Of Facilities and Equipment
E2.1 Followuo on Service Water System Reoairs (37550)
a. Insoection Scooe
The inspector reviewed the licensee's followup actions to monitor
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two through wall leaks in the Unit 1 vital service water header.
b. Findinos and Observations
The licensee identifi- two leaks in a section of piping on the
Division 1 vital service water header. These problems were
documented on CR Numbers 97-02013 and -02108. Each leak was
estimated to be approximately one drop per minute. Based on the
results of ultrasonic testing (UT) licensee engineerr determined
that the leaks were of the " pinhole" type. The inspectors
reviewed ESR 9700326. Revisions 0 and 1 which evaluated the piping
wall thickness using the guidance provided in Generic Letter 90-
05. The conclusion of the ESR was that the piping was operable
until the next refueling outage. The licensee submitted a relief
request to NRC regarding this issue in a letter dated July 3.
1997. The licensee committed to perform weekly assessments of the
leakage from the pinhole and to assess the integrity of the flawed
area once every three months using nondestructive examination
methods. The inspectors reviewed the results of UT evaluations
performed on June 15. 1997 August 19, 1997. November 3. 1997. and
January 30, 1998. The UT results showed the piping remained
serviceable. The inspectors also reviewed the results of weekly
inspections of the temporary repair which are performed by
operations personnel in accordance with Operating Instruction 1-
01-03.4.2. Auxiliary Operator Temporary Check Sheet. Unit 1
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Reactor Building. The inspectors concluded that the licensee was
performing ongoing evaluations of the pipe flaws in accordance
with their commitments to NRC and good engineering practices.
c. Conclusions
The licensee's actions to followup on the pinhole leaks in the
Unit 1 vital service water header were performed in accordance
wi th -engi neeri ng - recommendati ons .
E.8 . Miscellaneous Engineering Issues
E.8.1 (Closed) Inspector Followup Item 50-325(324)/96-14-06. Accuracy of ERFIS
and SPDS Data.
I
Review of various documents resulted in identification of an issue
concerning the accuracy of reactor vessel level instrumentation
using the emergency response facilities information system (ERFIS)
and the accuracy of containment isolation valve position
l
indications in the safety parameters display system (SPDS). A
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. problem had been identified in 1994 with the algorithm u;ed to
- compute the water level for small break LOCA scenarios. A
software change, number SCR 94-0114. was generated to resolve the
y- problem. However, since the water level data displayed on ERFIS
was an average value. operators were trained not to depend on the
l ERFIS system when responding to accident scenarios. The
l inspectors noted the following caution in E0P-01-UG Attachment 6:
Reactor water level instruments may be used to determine reactor
water level only when the conditions for use as listed in Table 1
are satisfied for that instrument. Table 1 listed the conditions
for various instruments. The inspectors noted, during observation
of training that operators did not to rely on ERFIS data, The
operators were trained to rely on water level data indicated on
control board instruments using the limitations contained in the
E0Ps. The same was true regarding valve positions. The data
shown on SPDS was' used for information only. The operators
referred to the control board indications when determining valve
position.
E.8.2 (Closed) Unresolved Item 50-325(324)/97-05-06. Deletion of RTT
Requirements
Resolution of this URI is discusstJ in paragraphs M1.1 and M1.2,
above. Four apparent violations were identified.
E.8.3 (Closed) Unresolved Item 50-325(324)/97-09-02. Accuracy of Measured
Temperature Data for EQ Evaluations
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l Calculation..BNP-E0-5.003. Revision 0. Reactor Building Ambient
l Temperature For Qualified / Expected Life Calculations was reviewed. This
l calculation was performed to justify use of actual environmental
l temperature data in lieu of design maximam temperature data for
l determining qualified life. The calculation presented actual Brunswick
l plant temperature data from the ERFIS computer for various locations in
l Unit 1 and Unit'2 Reactor Buildings for 1992. 1993. 1996 and 1997. The
l data showed that Unit 1 had higher temperatures than Unit 2 and that
l 1993 was the highest temperature year from comparison of daily average ,
l temperature data for the past 20 years (20 year mean) for statewide and l
l Southport. North Carolina. This provided justification that if a i
temperature model could be shown to provide the same degradation as the l
l 1993 data. that the model would be conservative and representative of l
l past plant equipment aging due to actual plant environments. I
The calculation demonstrated that the actual data coulo be enveloped by
an ambient profile of 85 degrees Fahrenheit (F) for six months and 95
l
,
degrees F for six months. The analysis showed that the equivalent '
! degradation of the 85/95 degree F profile was worse than the degradation ,
due to the actual environmental data for the combined 1992 and 1993 l
years which had been previously shown to be higher than past
temperatures. The evaluation was performed using a spreadsheet which
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had been validated by use of independent verification. Therefore, the
inspectors concluded that the temperature values used for evaluation of
the motor control centers were acceptable.
[ The inspectors also reviewed DR 227. Environmental Qualification ;
,
Service Condition. Revision 4. dated February 2. 1998. This
i document discusses environmental service conditions for
l environmental qualification of equipment in the reactor buildings
and drywells. In addition to temperature, precsure effects,
radiation and moisture are evaluated. The inspectors reviewed
various ESRs and the reactor building environmental report which
contain temperature data and verified that additional information
was not available which could possibly invalidate the conclusions
i of Calculation BNP-E0-5.003. However, during review of DR-227.
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the inspectors identified the following problem. Reference 146 of
l DR-227 was ESR 9700314. Drywell Temperature Monitorng of Past
Outages. Revision 0. dated June 3, 1997. The purpose of this ESR
was to tabulate a listing of all plant outages and to calculate an
average outage temperature for the drywells, based on the reduced
drywell temperatures which occur during outages. This ESR. which
was incorporated into Section 4.1.1.1 of DR-227, was used to
reduce the temperatures used in the Arrhenius technique for E0
equipment life. The inspectors identified an error in the ESR
which failed to consider the dates Unit 2 was not operating for
extended periods in 1992 and 1993. Although the error was in the
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conservative direction (slightly higher than actual drywell ,
temperatures resulted from evaluation), the inspectors noted that j
no design verification had been performed on the ESR since it was I
classified as an Engineering Disposition (ED) ESR. The definition 1
of an ED ESR per CP&L Procedure EGR-NGGC-0005. Engineering j
Service Rquests in part, was an ESR, which should not produce
design output documents. nor should it change existing engineering
documents. The use of an ED ESR as a design output document to
change plant design parameters (drywell temperatures) was
identified to the licensee as violation item 50-325 (324)/98-04-
05. Use of ED ESR to Revise Plant Design Data.
E.8.4 (Closed) Unresolved Item 50-325(324)/97-09-04 Effect of ,
Depressurized Electrical Penetrations on E0 Requirements )
Review of ODP-14. discussed in paragraph , above, disclosed that
the Westinghouse penetrations were not required to be pressurized
to maintain environmental qualification. The penetrations were
pressurized to identify potential leakage which could affect
compliance of the containment with 10 CFR 50. Appendix J.
E.8.5 (Closed) LER 1-97-003. Response Time Test Elimination
.
This LER covered the inappropriate elimination of response time
testing. Resolution of this LER is discussed in parage.,-hs M1.1
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and M1.2. above.
V. MANAGEMENT MEETINGS
,
The inspectors presented the inspection results to members of licensee
management at the conclusion of the inspection on March 12. 1998, and
l during a telecon on March 26. 1998. The licensee acknowledged the
i findings presented. Dissenting comments were not received from the
! licensee. The licensee did not identify ar. materials used during the
- inspection as proprietary information.
PARTIAL LIST OF PERSONS CONTACTED
Licensee
G. Attarian. Manager. E0 Task Force
R. Delong. Superintendent. Electrical /l&C. Brunswick Engineering Support
Section (BESS)
W. Dorman Manager. Licensing and Regulatory Affairs
J. Gawron, Manager. Nuclear Assessment Section
K. Jury. Manager. Regulatory Affairs
R. Krich. Chief Engineer. Nuclear Engineering (NED)
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J. Lyash Plant Manager
J. McPadden',' Project Engineer. BESS
G. Miller. Manager. BESS
'R. Mullis, Manager. Operations l
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l S. Tabor Senior Specialist. Regulatory Compliance
Other licensee employees contacted included engineers Nuclear
Assessment personnel and administrative personnel.
INSPECTION PROCEDURES USED l
IP 37550: Engineering
IP 92903: Followup - Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED l
'
Doened
50-325(324)/98-04-01 EEI Inadequate 10 CFR 50.59 Evaluations which
Resulted in Deletion of TS Response Time
Testing Requirements (Paragraph M1.1.b)
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50-325(324)/98-04-02 EEI Failure to Perform RTl for Reactor j
l Protection System Instrumentation as i
Required by TS 4.3.1.3 (Paragraph M1.2.b) !
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50-325(324)/98-04-03 EEI Failure to Perform RTT for Actuation
of the Primary and Secondary .
Containment Isolation Systems as !
l Required by TS 4.3.2.3 (Paragraph
M1.2.b)
l
l 50-325(324)/98-04-04 EEI Failure to Perform RTT for Actuation
of ECCS as Required by TS 4.3.3.3
l (Paragraph M1.2.b)
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50-325(324)/98-04-05 VIO Use of ED ESR to Change Plant Design
Documents (Paragraph E'.8.3)
Closed
50-325(324)/96-14-06 IFI Accuracy of ERFiS and SPDS Data (Paragraph :
E.8.1)
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l 50-325(324)/97-05-06 URI Deletion of RTT Requirementh (Paragraph
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E.8.2)
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50-325(324)/97-09-02 URI Accuracy of Measured Temperature Data for
EO Evaluations (Paragraph E.8.3)
50-325(324)/97-09-04 URI Effect of Depressurized Electrical ,
,
Penetrations on E0 Requirements !
(Paragraph E.8.4) j
LER 1-97-003 Response Time Test Elimination
(Paragraph E.8.5)
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ATTACHMENT .
UNIT 1 RESPONSE TIME TEST DATES
REACTOR PROTECTION SYSTEM (TS 4.3.1.3)
l
' TRIP FUNCTIONAL UNIT MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
Reactor Vessel
Steam Dome Pressure - High 3/24/94 2/10/96
'
Reactor Vessel
Water Level - Low. Level 1 1/31/94 12/19/95
I
ISOLATION SYSTEM INSTRUMENTATION (TS 4.3.2.3)
TRIP FUNCTION MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
Isolation
Reactor Vessel' Water
Level Low. Level 1 & 3 1/31/94 12/19/95
l
Drywell Pressure digh '7/28/94 6/15/96
Pressure - Low 4/10/94 2/27/96
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Flow - High 4/14/94 3/2/96
Main Steam Line Tunnel 1/30/94 12/18/95 l
Temperature - High
Condenser Vacuum - Low 4/11/94 2/28/96 >
Secondary Containment System
Isolation
Reactor Building Vent
Exhaust Plenum Radiation - High 2/1/94 12/20/95 ;
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Drywell Pressure - High 7/28/94 6/15/96 !
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. Attachment 1
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Reactor Vessel
WaterLevel-Low, Level 2 4/28/95 3/16/96
Reactor Water Cleanup System
Isolation
Reactor Vessel
.Water Level Low. Level 2 4/28/95 6/9/96
Reactor Core Isolation Cooling
System Isolation
1
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HPCI Steam Line Flow - High 1/30/94 12/18/95 l
HPCI Steam Supply Pressure - Low 1/31/94 12/19/95
RCIC Steam Line Flow - High 9/22/93 8/10/95
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EMERGENCY CORE COOLING SYSTEM (T3 4.3.3.3) l
TRIP FUNCTION MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
..
Core Spray System 11/9/95 9/27/97
Low Pressure Coolant Injection 1/31/94 12/19/95
Mode of RHR System
High Pressure Core Spray System 2/10/94 12/29/95
(a) Dates in this column represent the most recent date when response time was
satisfactorily demonstrated for at least one channel in a trip system associated with
the trip functional unit or trip function.
(b) Dates in this column represent the required test date based on; the last
satisfactory test date in column (a): Technical Specification 4.3.1.3, 4.3.2.3. or
4.3.3.3 from License Amendment 175 which, respectively, required a demonstration of
response time at least every 18 months: and Technical Specification 4.0.2. which
allowed the surveillance interval to extended by a maximum of 25 percent. Dates in
column (b) are 687 days after thcse in column (a).
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ATTACHMENT 2 l
UNIT 2 RESPONSE TIME TEST DATES
1
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REACTOR PROTECTION SYSTEM (TS 4.3.1.3)
TRIP FUNCTIONAL UNIT MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
Reactor Vessel
Steam Dome Pressure - High 10/3/94 8/20/96
Peactor Vessel
Water Level - Low, level 1. 6/24/94 5/12/96
l l
ISOLATION SYSTEM INSTRUMENTATION (TS 4.3.2.3) i
I
TRIP FUNCTION MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
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Isolation
,
Reactor Vessel Water
l. Level Low. Level 1 & 3 6/24/94 5/12/96
Drywell Pressure - High 4/17/93 3/5/95
Pressure - Low 3/9/93 1/25/95
, - Flow - High 6/7/94 4/25/96
Main Steam Line Tunnel 5/10/94 3/28/96
s Temperature - High
'
Condenser Vacuum - Low 3/9/93 1/25/95
Secondary Containment System
Isolation
Reactor Building Vent
Exhaust Plenum Radiation - High 4/15/93 3/3/95
., Attachment 2
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DrywellPressure-H5gh 4/17/93 3/5/95
i Reactor Vessel '
,
Water Level - Low Level 2 1/31/96 12/19/97
l
Reactor Water Cleanup System I
,
Isolation
!
Reactor Vessel
Water Level Low. Level 2 1/31/96 12/19/97
Reactor Core Isolation Cooling
System Isolation
l
i HPCI Steam Line Flow - High 8/5/93 6/23/95
l HPCI Steam Supply Pressure - Low 8/31/93 7/19/95
l
RCIC Steam Line Flow - High 3/10/93 1/26/95
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l. EMERGENCY CORE COOLING SYSTEM (TS 4.3.3.3)
l
l; TRIP FUNCTION- MOST RECENT SATISFACTORY TEST DATE (a) TEST REQUIRED BY (b)
i Core Spray System 1/5/95 11/23/96
Low Pressure Coolant Injection 6/13/94 E/1/96
l Mode of RHR System
i
i High Pressure Core Spray System 5/11/93 3/29/95
l
(a) Dates in this column represent the most recent date when response time was
satisfactorily demonstrated for at least one channel in a trio system associated with
the trip functional unit or trip function.
(b) Dates in this column represent the required test date based on: the last
satist'actory test date in column (a): Technical Egecification 4.3.1.3. 4.3.2.3. or
4.3.3.3 from License Amendment 206 which. respectively. required a demonstration of
response time at least every 18 months; and Technical Specification 4.0.2. which
allowed the surveillance interval to extended by a maximum of 25 percent. Dates in
columa (b) are 687 days after those in column (a).
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