ML20196H413
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| ML20196H413 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/05/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20196H401 | List: |
| References | |
| 50-338-98-05, 50-338-98-5, 50-339-98-05, 50-339-98-5, NUDOCS 9812090064 | |
| Download: ML20196H413 (34) | |
See also: IR 05000338/1998005
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U. S. NUCLEAR REGULATORY COMMISSION
REGION ll
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Docket Nos.: . 50-338,50-339
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License Nos.: NPF-4, NPF-7
Report Nos.: 50-338/98-05,50-339/98-05
Licensee: Virginia Electric and Power Company
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Facility: North Anna Power Station, Units 1 & 2
Location: 1022 Halley Drive
Mineral, Virginia 23117
Dates: July 13 - 17, July 27 - 31, and September 21 -25,1998
Team Leader: J. Lenahan, Senior Reactor inspector
Engineering Branch
Division of Reactor Safety
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Inspectors: P. Fillion, Reactor Inspector
R. Chou, Reactor inspector *
J. Panchison, Engineering Consultant !
S. Rohrer, NRC Graduate Fellow
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Approved By: Kerry D. Landis, Chief
Engineering Branch ;
Division of Reactor Safety
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Enclosure 2
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9812090064 981105
PDR. ADOCK 05000338
e PDR
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EXECUTIVE SUMMARY
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North Anna Power Station
NRC Inspection Report 50-338/98-05,50-339/98-05
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This inspection included a review of Qe licensee's calculations, analysis, performance test !
procedures and other engineering documents that were used to support design and i
performance of the auxiliary feedwater (AFW) system during normal and accident or abnormal l
conditions. The report covered a two-week period of onsite inspection.
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Overall the inspection found the system operation to be consistent with the design and licensing
basis.
Maintenance !
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The maintenance of the auxiliary feedwater system components has been good. The i
AFW system has performed reliably. Maintenance practices have been adequate. !
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e The material condition of AFW equipment and components examined was good as well ;
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as housekeeping in the general areas around equipment and components in the AFW
pump houses and main steam valve houses.
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e The material condition of piping and housekeeping in the AFW pipe tunnels was very !
poor. A violation was identified for failure to take effective corrective action to address >
corrosion problems in the Unit 2 AFW pipe tunnelidentified previously in September, i
1996. j
e A violation was identified for failure to construct the Unit 1 AFW pipe supports in
accordance with the design drawing requirements. l
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A violation of 10 CFR3 50.55(a) was identified for failure to include safety-related AFW
pipe supports installed in the AFW pipe tunnels in the ISI program.
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e The maintenance program for the 4 kV circuit breakers was considered a strength. This )
conclusion was based on good procedures, refurbishment of breakers at ten-year j
maximum intervals, and the excellent practice of periodic functional testing of all control l
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circuit devices.
e AFW pump surveillance test procedures permitted operation of the pump in excess of
the piping design pressure and relief valve setpoint during surveillance testing. j
Enclosure 2
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Enoineerina i
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e The design control procedures complied with the requirements of 10 CFR 50.59 and 10
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CFR 50, Appendix B, Criterion Ill. The instructions / checklist for preparation of 10 CFR {
50.59 safety evaluations were indepth and thorough. The design control
program / procedures are good.
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e The licensee's process for screening potential changes to determine if a 10 CFR 50.59 j
safety evaluation was required. The licensee's process does not include an independent
review of the initial safety screening.
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e Design and installation of AFW electrical equipment was in accordance with good l
industry practices, NRC requirements, and the licensing basis. The AFW system" meets .
the single failure criterion. l
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The electrical calculations reviewed were found to be accurate and consistent with l
licensing commitments. The electrical calculation quality was good. I
e The licensee was unable to provide documentation for one subpart of the diesel !
generator loading analysis identified during a previour. inspection. This issue was
identified as an IFl. ,
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e Instrument setpoint calculations used an approved methodology and considered f'
appropriate sources of instrumentation inaccuracies.
* The quality of AFW mechanical design calculations was poor. Numerous discrepancies [
were identified between various design calculations and design documents.
Calculation quality was identified as a weakness.
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* Two examples of fail'ure to update the UFSAR were identified. Discretion was exercised
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for this violation of 10 CFR 50.71(e) and was not cited (EA 98-500). l
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e The system design basis document (SDBD) was a comprehensive consolidation of I
design and licensing base criteria for the APN system. However some deficiencies in
design documentation references were identified.
e A weakne ss in the licensee's self-assessment was identified due to lack of consistency in
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the me%odology, reporting of results, documenting identified deficiencies, and tracking
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e A weakness was identified in the licensee's process for completion of operability
evaluations. The final operability evaluation which determined the effect of missing pipe
- supports on the operability of the AFW system was not completed until more than six l
weeks after the initial missing support was identified. l
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Report Details
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Introduction - -
' , The objective of this Safety System Engineering Inspection (SSEI) was to assess the adequacy '
of calculations, analysis, other engineering documents, and maintenance practices that were
used to support AFW system performance during normal and accident or abnormal conditions.
The inspection was performed by a team of inspectors that included a Team Leader, two Region
11 Inspectors, an NRC graduate fellow, and one engineering consultant. Prior to this inspection,
the licensee performed a review of the design, and licensing basis of the AFW system as part of
their Integrated Configuration Management Project. The results of this review were used to
revise and update the AFW System Design Basis Document (SDBD) which was issued a~s
Revision 2 on July 13,1998.
II. Maintenance
M2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Material Condition of Auxiliary Feedwater System
a. Inspection Scope (IP-93809)
The team reviewed maintenance records and conducted walkdown inspections to
determine the condition of the auxiliary feedwater system and the material condition of
equipment and components within this system.
b. Observations and Findinas
The team noted that material condition of equipment and components examined was
very good as well as housekeeping in the general areas arou.nd equipment and
components in the AFW pump buildings and main steam valve houses. However,
material condition and housekeeping in the AFW piping tunnels was very poor as
evidenced by standing water in the tunnels and the severe corrosion noted on some pipe
supports and baseplates. This problem is further discussed in paragraph M2.2.
The team reviewed maintenance and test records for AFW system components for the
last four years. The records included repairs, maintenance, replacement, and
inspections for the following AFW components: check valves, Terry Turbine throttle
valves, valves, actuators, Limitorque operators, pumps, and motors. The records
indicated that post maintenance tests were performed and that quality controlinspectors
witnessed maintenance work when this was required by the procedures. The test
records included the valve and flow instrument calibration data for AFW system. Some
of the tests were found out of the acceptable ranges in the "As found" data, and were
adjusted to be within the acceptable ranges.
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. c. Conclusions
The maintenance of the auxiliary feedwater system components has been good. The
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AFW system has performed reliably. Maintenance practices have been adequate. The
material condition of AFW equipment and components examined was very good as well
as housekeeping in the general areas around the equipment and components.
However, the material condition of piping and pipe supports and housekeeping in the
AFW tunnels was very poor.
M2.2 Inspection of AFW Pioina and Pipe Supports
a. Inspection ScoDe
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The team conducted walkdown inspections of the AFW piping systems to determine the
material condition of the piping and supports.
b. Observations and Findinas .
On May 26,1998, the licensee inspected the AFW piping in the Unit 1 AFW pipe tunnels
under PT 171.3, as part of a system pressure test. As a result of the May 26 inspection,
the licensee identified a concern regarding external corrosion on some of the AFW
piping. The licensee issued Deviation Report (DR) N-98-1677 to document and
disposition this issue, and inspected the piping in the Unit 2 AFW tunnel. In addition to
the piping corrosion problem, some material condition issues were identified in the pipe
tunnels conceming debris in the tunnels, " minor" corrosion of Unit 2 pipe supports and
baseplates, and some missing or damaged insulation on the Unit 2 steam supply line to
the AFW turbine driven pump. The licensee's corrective actions for the additional issues
were to issue minor maintenance cards.
The team reviewed the licensee's evaluation of the piping corrosion problem. The
licensee performed a nondestructive examination (NDE) of the piping using ultrasonic
testing (UT) to determine the existing pipe thickness. A design review and calculation
was also completed as a parallel effort to determine the minimum acceptable piping
thickness.
The team reviewed calculation number CE-1412, Rev. 0; Engineering Transmittal (ET)
CME-98-0035 Rev. O, and Ultrasonic Thickness Record NDER 98-206, Rev. 4. NDER
98-206 which document the results of UT examinations of the piping showed the
min'imum wall thickness for all pipe examined was 0.234". ET CME-98-0035 calculated
a required minimum pipe wall thickness of 0.231" to resist internal pipe pressure (1480
psi) for pipe hoop stresses. Calculation CE-1412 calculated a minimum pipe wall
thickness of 0.1" to resist the total mechanical loads including pressure, dead wei6ht,
thermal, and dynamic for the longitudinal stresses. Therefore, the measured pipe
minimum wall thickness after the consideration of corrosion was greater than the
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niinimum pipe wall thickness required for all stresses and was acceptable. The team
reviewed the results of the UT exams which showed the minimum piing thickness
exceeded the minimum thicknesses calculated in Calculation number CE-1412 and ET
CME-98-0035.
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During the preparation of the calculation, licensee design engineers requested that
additional walkdown inspections be performed. Site engineering personnel performed
the walkdowns on June 16,1998, and identified a missing pipe hanger (gang hanger)
which was designed to support seven individual pipe knes, including the three AFW
supply lines and the steam supply to the turbine driven AFW pump. This problem was
documented on DR N 98-1881. On !uly 7,1998, an additional gang hanger, which was
designed as a pipe anchor for the three AFW supply lines was found to be missing. This
problem was documented on DR -98-2106. The licensee's operability evaluations for
these issues is discussed in paragraph E3.2.
The inspectors walked down portions of the auxiliary feedwater piping, including the
main steam supply piping to the turbine driven pump, to determine the condition of the
piping and pipe supports. Piping and supports examined included piping installed ~in the
AFW tunnels and the AFW pump buildings. Portions of approximately 60 seismic
supports were examined by the team to determine if they were installed in accordance
with drawing requirements. Numerous discrepancies between the as-built pipe supports
.and the requirements shown on the design drawings were identified by the team during
the walkdowns. Some of these discrepancies are listed in the Table in Appendix 1.
Approximately one-third of the welds inspected on these supports were determined to be
undersized. Some welds were found to be only one-half the size specified on the design
drawings. The failure to construct the pipe supports in accordance with the requirements
shown on the design drawings was identified to the licensee as a viciation of 10 CFR 50,
Appendix B, Criterion V, Failure to Construct Pipe Supports in Accordance with Design
Requirements (Violation Item 50-338,339/98-05-01). The licensee issued DR numbers
N-98-2200 and N-98-2204 to document and disposition the above discrepancies.
Discussions with licensee engineers disclosed that the pipe support corrosion problem in
the AFW tunnels had been identified by licensee operations personnel on
September 29,19963DR number N-96-2135 was issued on September 29,1996, to
document and disposition the problems. The corrective actions included installation of
sump pumps in the tunnels to remove any standing water, and cleaning corrosion from
the supports and recoating of the supports. The DR was closed on October 10,1996,
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by issuing work requests and a design change to complete the corrective actions.
Examination of the pipe supports in the Unit 2 AFW pipe tunnel disclosed that the
coatings had deteriorated on the baseplates and lower sections of the supports in
proximity of the baseplates. This was apparently due to standing water in the tunnels
which had been three to six inches deep. The coatings could be scraped off using a
fingernail or thin piece of metal using minimal effort. The coatings would peel off in
sheets and revealinat the steel members had been severely pitted and corroded. The
team estimated in some extreme cases, that approximately 25 percent of the basemetal
for the support members and baseplates was lost due to corrosion. The corrosion
issues were also documented in DR -98-2204, which was initiated on July 16,1998.
Planned corrective actions in the DR pertaining to the corrosion issues included
replacement of severely corroded components and cleaning and painting of components
which did not exhibit excessive corrosion. The licensee will also initiate a program to
ensure that the sump pumps installed in the tunnels remain operable and perform
periodic inspections in the tunnels to ensure that the supports are not continuing to
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degrade. A root cause analysis will also be performed to determine the cause of failure
to implement the corrective actions for DR -96-2135.
FaiNre to take effective corrective action to address the corrosion problems in the Unit 2
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AFW tunnel was identified to the licensee as a violation of the corrective action program,
Failure to Correct Cause and Effects of Corrosion of Unit 2 AFW Pipe Tunnel Safety-
Related Pipe Supports, (Violation item 50-339/98-05-02). Implementation of the
licensee's corrective actions, documented on DR - 98-2200, will be reviewed by NRC in
a futuis inspection.
The team questioned licensee engineers as to why the corrosion issues had not been
identified during routine in-service inspections performed under the ASME Section XI
inservice inspections required by Technical Specification 4.05 and 10 CFR 50.55a(g).
After review of their current inservice inspection procedures and records, licensee ,
engineers determined that the supports for the AFW supply lines located in the Unit 2 .
AFW tunnels had been inadvertently deleted from the inservice inspection program. The i
licensee lssued DR -98-2211 to document and disposition this problem. In their initial
assessment of corrective actions, licensee engineers concluded that only 39 Class 3 l
supports located in the Unit 2 AFW pipe tunnel had been omitted from the program, and I
that failure to include the supports in the ISI program did not violate the Technical
Specifications. Subsequent to the inspection, the licensee identified some Unit 1 safety- :
related supports which also had not been included in the ISI program. The licensee j
issued DR -98-2375 on August 4,1998 to document and disposition this problem. The ;
licensee submitted Licensee Event Report 50-338/98-003-00 to the NRC in a letter dated l
August 21,1998 to report this problem. The cause of failure to include the supports in j
the ISI program was determined to be failure to include the drawings which documented :
the support locations in the ISI program data base. The licensee provided a j
supplemental response to the LER (Revision 1) on September 17,1998, which clarified ,
the additional corrective actions which had been initiated to identify any additional !
supports omitted from the ISI program. These corrective actions included a review of the !
ISI Isometric (WMKS) drawings and class boundary (CBM) drawings to ensure no other !
drawings had been omitted from the ISI program database and that no additional :
supports had been omitted.
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Failure to include 10 Unit 1 and 39 Unit 2 safety related AFW pipe supports installed in
the AFW tunnels in the Section XI in-service Inspection program was identified to the ;
licensee as a violation of 10 CFR 50.55a(g), Failure to include AFW Pipe Tunnel 4
Supports in The ISI Program (Violation item 50-338, 339/98-05-03). NRC will review the
licensee's corrective actions described in the LER in a future inspection to verify that the :
licensee's ISI program complies with the requirements of 10 CFR 50.55(a). !
c. Conclusion
The material condition of piping and housekeeping in the AFW pipe tunnels was very [
poor. This was identified as a weakness. A violation was identified for failure to take i
effective corrective action to address corrosion problems in the Unit 2 AFW pipe tunnel
identified previously in 1996. A violation was identified for failure to construct safety- i
related AFW pipe supports in accordance with design drawing requirements. A violation
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of 10 CFR 50.55(a) was identified for failure to include safety related pipe supports
installed in the AFW pipe tunnels in the ISI program.
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M3 Maintenance Procedures and Documentation
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M3.1 Maintenance of Electrical Equipment in the AFW System I
a. Inspection Scone -
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The team reviewed the preventive maintenance program for the 4 kV circuit breakers l
and protective relays associated with the AFW pump motors.
b. Observations and Findinas
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The team reviewed the licensee's procedures which control maintenance activities on 4
kV breakers and protective relays associated with the AFW pump motors. The team
compared the 4 kV breaker maintenance procedures to good industry practice as ;
determined from various published sources and knowledge of procedures employed by '
other licensees. The team had the following comments on Procedure 0-EPM-0302-01,
Revision 17:
* On page 21 of the procedure, in the section on examination of the main contacts,
the procedure erroneously stated that " Silver oxide h a good conductor." The
licensee stated that the procedure would be revised .o correct the statement to l
" Silver oxide is not a good conductor". Conceivably, the statement as originally
written could have caused a craftsmen to not clean off main contacts coated with
silver oxide. The team concluded that there was no operability concern here,
because the procedure specifies a contact resistance test which would detect j
abnormally high contact resistance. ;
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* The procedure did not specify high potential testing. The license responded that :
routine high potential testing was net their policy, however they agreed to check !
the manufacturer's instruction book and contact the' manufacturer's I
representative to see if high potential testing is specifically recommended.
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* The procedure did not specify functional testing at minimum rated control voltage.
The system engineer stated that he recently recommended to incorporate
reduced control voltage testing into the procedure. This recommendation was !
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made based on new information provided by NRC and by the circuit breaker
owners group.
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* The procedure did not specify inspection of the main toggle link pin. The licensee )
agreed to check with the manufacturer, j
* The procedure did not contain sufficient detail for inspection of the primary
disconnects. The licensee stated they would evaluate the need for more detailin
the procedure.
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The team found that the procedures contained nearly all maintenance items considered
good industry practices. The team reviewed records of the last two times that preventive !
maintenance was performed on the 4 kV circuit breakers for the AFW pump motors. !
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- From examination of these records, the team found that the licensee has been
- performing preventive maintenance on the 4 kV circuit breakers at nominal 4.5-year
intervals (five years maximum). The breaker preventive maintenance program specified
j in Maintenance Department Administrative Procedure MDAP-0017, Revision 0, dated 1
July 1,1990, through the current revision, alternated two different procedures: One -
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consisted of a basic set of inspections, tests and functional checks; and the other
consisted of the basic set plus a complete tear down and refurbishment. The intent was .
] to refurbish each circuit breaker at nine-year nominalintervals. Examination of the ! a '
breaker maintenance records showed that, with one exception, the defined program had
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been implemented. The actual maintenance interval for one of the breakers (Unit'1,3B
AFW PP) was 7 years rather than the program specified 5 years. The licensee initiated
j Deviation Report No. N-98-2277 documenting the deviation from the normal program. j
Licensee engineers reviewed the records for 14 other breakers, and did not identify any ;
additional maintenance interval deviations. The team did not review these other records.
Examination of the maintenance records showed that the licensee was performing a '
functional test on the 4 kV breaker control circuits at 18-month intervals. This test was a
3' detailed functional test of all control circuit devices, and essentially was a periodic repeat !
of the pre-operational functional test.
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Examination of the maintenance records showed that the protective relays for the AFW i
j pump motors were calibrated within the last few years. The records showed that two test j
points were used on each inverse time relay. The relay calibration data sheets were
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- The team reviewed tt)e manufacturer's instruction book for the motor driven AFW pumps (
and determined that there were not any special precautions or limitations affecting
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c. Conclusions
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The maintenance program for the 4 kV circuit breakers was considered a strengtn. This ;
conclusion was based on good maintenance procedures, refurbishment of breakers at !
ten-year maximum intervals, and the excellent practice of periodic functional testing of all !
control circuit devices. ;
M3.2 AFW Pump Surveillance Testina ;
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a. Inspection Scope (IP-93809)
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The inspectors reviewed surveillance test procedures for the AFW pumps to determine if
acceptance criteria were consistent with the design and licensing bases.
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b. Observations and Findinas
Surveillance test procedure 1-PT-71.10 was reviewed. This procedure covered the
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quarterly pump surveillance test performed on AFW turbine driven pump 1-FW-P2. The
surveillance test procedure specified an acceptance value for the upper discharge
pressure limit which exceeded the piping design pressure and the discharge piping relief
valve s'etting. The acceptable range for pump differential pressure noted in procedure
Attachment 1 was 1274.0 - 1551.6 psig. The design pressure of the pump discharge
piping was 1480 psi. The discharge piping relief valve was set at 1480 psi. The revised
pipe stress analysis for this piping used 1500 psi. The test pressure acceptance range
noted in the procedure is in accordance with OM-6, ASME Operational and Maintenance
Code, Part 6, which permits test pressure acceptance range to be ten percent greater
than the nominal design pressure. However, the team noted that the calculation basis
that determined this acceptance limit did not consider the relief valve setpoint. The
licensee indicated that the delta P calculation basis in the procedure would be revised.
This was not a safety concern since the relief valve setting would have prevented over
pressurization of the piping.
c. Conclusions
AFW pump surveillance test procedure permitted operation of the pump in excess of the
system relief valve setpoint during surveillance testing.
111. ENGINEERING
E1. Conduct of Engineering
E1.1 Design Change Cont,rol and 50.59 Processes
a. Inspection Scope
The team reviewed the licensee's procedures which control t'he design change process.
b. Observations and Findinas
The team reviewed the current revisions of the licensee's design control procedures.
The procedures adequately addressed the following: design input, design verification,
control of design output documents, post modification testing, control of field changes, l
training requirements, and 10 CFR 50.59 reviews. The procedures provided good
controls for maintaining the design basis and for implementation of design changes. The
team noted that the procedures specifically excluded using unverified information such
as that obtained in telephone conversations or verbal communications, or other
undocumented sources from use as design inputs in design documents / calculations.
Procedure VPAP-3001 provides the instructions for completing the 10 CFR 50.59 safety
review. The procedure requires completion of a safety evaluation for physical changes
to the plant, computer software changes, and setpoint changes that have the potential
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for making a chcnge to a structure, component, or system different from that specified in
an existing controlled design document. The procedure defines use of furmanite,
scaffolding placement, tube plugging, and installation of shielding (lead blankets) as
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changes which require a safety evaluation. A detailed 12 page safety evaluation form
and checklist is included as Attachment 3 to VPAP-3001. The team determined that the
instructions / checklist for preparation of safety evaluations were indepth and thorough.
The procedure specified detailed training requirements for preparers, reviewers, and
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approvers of safety evaluations. All safety evaluations are required to be independently
reviewed by a qualified reviewer, and approved by the cognizant supervisor.
Procedure VPAP-3001 also provides instructions for performance of an activity
screening. This process documents that an activity does not require a safety evaluation
and is required for all activities which do not require a safety evaluation. A scree 5ing
need not be completed if it determined that a safety evaluation is required for an activity.
The procedure specified training and experience requirements for personnel who
perform the activity screening. The experience and training requirements for screeners
are less than those specified for safety evaluators. The team noted that activity
screening do not require an independent review, or approval of the cognizant supervisor.
The team noted that if the activity screener incorrectly concludes that an activity does not
require a safety evaluation, since no independent review is perfo.med for the screening,
there is a potential that a change could be implemented without performing a safety
evaluation. The team noted that less experienced personnel ofhn perform the activity
screening.
c. Conclusions
The design control procedures complied with the requirements of 10 CFR 50.59 and '
10 CFR 50, Appendix B, Criterion 111. The design control program / procedures are good. !
The instructions / checklist for preparation of 10 CFR 50,59 safety evaluations were
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indepth and thorough. The licensee's process for screening potential changas to '
determine if a 10 CFR 50.59 safety evaluation was required does not include an ,
independent review of the initial safety screening. l
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E1.2 Electrical Desian Review
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a. Inspection Scope
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The team reviewed control circuits, protective devices, calculations, equipment ratings, )
as well as electrical separation and isolation in relation to the design basis requirements.
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b. Observations and Findinas
The team found all the control circuits reviewed correctly implemented the system
operation as described in the UFSAR. Control device contacts in these control circuits
were applied within their make and break ratings. The team also verified the overload
relay rtyle and set point associated with the discharge valves in the normal flow path of
both units in a walkdown inspection.
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The protective relay and overload relay settings were established by calculations. The
team found these calculations were completed in accordance with published industry
guidance developed by the IEEE. The team examined the installed overload relay for
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MOVs 1008,100D,2008 and 200D, and compared the installed relays to the set point
sheets and the calculations. The team concluded that all these were consistent,
indicating good configuration control. The team found that protective relays, molded-
case breakers and overload relays were properly set, except as described below. The 1
one case where a relay may not have the optimum set point is the ground fault relay at
the stub bus main breaker. The stub bus feeds the component cooling pump and the
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residual heat removal pump. The ground fault relay at the main breaker is an inverse i'
time IAC51 similar to the type used in a residual current connection. The ground fault
relays at the two motor feeder breakers are plunger type PJC relays used with a
doughnut current transformer. When relay operating times and breaker operating time
were factored in, the coordination margin was 8 cycles at the maximum fault current of
1420 A. The PJC relay has low impedance, and will not cause saturation of the 50 - 5 A
current transformer. The licensee did have two calibration points on the IAC51 curve.
The engineer responsible for establishing relay setpoint agreed the margin was tight. He i
stated that the settings of the relays would be reviewed. The team determined this issue
had minor safety significance since the potential consequences of miscoordination in this l
case were very limited.
The team found that the motor horsepower rating and feeder cable size for the AFW
pump motor were adequate. The feeder cable ampacity was independentiy determined
by the team through reference to published standards, and the cables had a good deal of
margin. The walkdown of the Unit 1 cable route verified that fire retardant material was i
not applied to the cables and no derating was necessary. The team observed that
adequate physical separation was maintained between the train A and B AFW pump
motor feede cables.
With regard to reviev) of the single failure criterion, the AFW System had three flow paths
which were very nearly independent. The team developed a diagram of the three flow
paths with the associated piping and power supplies indicated. The piping and power i
supplies were confirmed by reference to the official detail design drawings. The only
commonality of power supplies among the three paths was that the 3B AFW pump is i
powered from train B and the discharge valve for the turbine driven AFW pump is
powered from train B. The team then postulated various power supply failures together I
with piping breaks at various points. Using this inspection methodology, the team found
the design met the single failure criterion. In addition, the team observed that the valves :
in the flow path were globe valves which are not vulnerable to pressure locking and I
thermal binding.
The team questioned licensee engineers to determine if all relevant cases were covered
by the emergency diesel generator loading analysis. The team questioned whether the
l motor starting transient analysis analyzed the case of safety injection signal or l containment depressurization signal occurring shortly after (say 25 seconds) the voltage
relay actuation resulting from the loss of offsite power. The team found that this same
issue had been raised by the Electrical Distribution System Functional Inspection
(EDSFI) conducted in July / August 1991 which was documented in NRC Inspection
i
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_ _ _ _. ._ _ _ _ _ . _ . _
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Report numbers 50-338,339/91-17. In their February 28,1992 response to the EDSFI,
for this finding, number 91-17-08, the licensee stated: "The need to develop diesel
loading calculations specifically for a DBA occurring after a loss of offsite power was !
identified during the April 1991 self assessment. We will analyze this condition by
-
i
February 1993 for enhancement to our design." l
,
Licensee engineers were not able to provide documentation during the inspection
showing that this delayed DBA case had been analyzed. The specific concern with the ;
delayed DBA signal arises from the fact that the sequencing of the various loads onto the 1
diesel generator are controlled by independent timers and process variables that are not i
necessarily coordinated in time. Loads such as service water and component cooling
pumps are controlled by timers initiated by voltage relays. Loads such as safety injection
and auxiliary feedwater pumps are controlled by voltage relays. The low head saTety
injection pump is controlled by the DBA signal. The quench spray pump is controlled by l
containment pressure signal and possibly a timer. The team postulated that the low
head safety injection pump (250 HP), quench spray pump (250 HP) and service water
pump (500 HP), for a total of 1000 HP could start simultaneously if the DBA signal, high
containment pressure and the service water timer actuation occurred simultaneously. It
was not clear that this case was bounded by the analysis which modeled tripping and
starting of the safety injection pump (900 HP) after all DBA loads had been sequenced
on. The former case does have 100 more horsepower but less preexisting load.
Preliminary analysis prepared during the inspection indicated that the three simultaneous
motors starts could not occur due to the way the various timere function. The licensee
indicated that the case in question may be beyond the design basis of the plant. The
team indicated that the case was probably within the design basis, basically because the
postulated delay in the DBA signal was only about 25 seconds, which could credibly be
within an evolution stemming from one initiating event. Inspector Followup item 98-05-
04, Diesel Generator Loading Transient Analysis for the Delayed DBA Case, was l
identified to the' licensee pending further NRC review of the analysis. Licensee
management indicated that they had committed to perform the delayed DBA diesel
generator loading analysis in their response to the EDSFI finding, and that they would
provide documentation demonstrating that they had fulfilled that commitment, or would
perform the required analysis.
I
c. Conclusions
Design and installation of AFW electrical equipment, including control circuits, electrical ;
protective devices, pump motors, and cable separation were in accordance with good
industry practices, NRC requirements, and the licensing basis. The AFW System met
the single failure criterion. The electrical calculations reviewed were found to be I
accurate and consistent with licensing commitments. The quality of the electrical
calculations was good. The licensee was unable to provide documentation for one
subpart of the diesel generator loading analysis that had been identified as an issue
during the Electrical Distribution System Functional Inspection. Preliminary analysis and
engineering judgement indicated there was no immediate operability concern.
1
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E1.3 Review of instrumentation Setooint Calculations
a. Inspection Scope
The team reviewed a set point calculation for instrumentation associated with the AFW
System to assure that the design basis for the plant was being maintained. I
b. Observations and Findinos -
The team reviewed the calculation of instrument loop uncertainty, operator logs
associated with readings of tank volume, and the alarm response procedure for the
emergency condensate storage tank (ECST). The team found the calculation of loop i
uncertainty for the condensate storage tank level instrumentation was performed -
according to the licensee's corporate design guides. The team found that the design
guide was essentially consistent with current recommended practices published by the
International Society for Measurement and Control. The alarm set point correctly
allowed for the alarm loop uncertainty. The acceptable level given in the operator logs
correctly allowed for the indicator loop uncertainty. Information on the annunciator
response procedure was consistent with the other relevant documents and specified
actions statements were correct. ;
c. Conclusions
t
The set point calculation used an approved rnethodology and considered appropriate
sources of instrumentation inaccuracies. However, some potential problems were
identified regarding the ECST tank volume during review of mechanical design
calculations. These are discussed in Section E1.4. !
E1.4 Mechanical AFW Desian Calculations
i'
a. Inspection Scope (93809)
The team assessed the quality of mechanical design calculations which support the
design and licensing basis for the AFW system. A review was performed of selected
calculations, analyses, the UFSAR, tne System Design Basis Document (SDBD), and
other engineering documents used to support system performance during normal and
accident or abnormal conditions. Interviews were also conducted with the licensee's
staff.
b. Observations and Findinas
Many of the active calculations were developed by the architect / engineer, Stone &
Webster (SWEC), during the initial design and licensing of the station. SWEC i
maintained design engineering responsibility until 1989 when the licensee assumed all
engineering activities. Because of this, many original calculations of record were ,
completed by SWEC up to the transition to the licensee's staff. In some cases, an
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original SWEC calculation was replaced by a new calculation, however, the old l
calculation was not removed from the active status. As a result, several calculations
;
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12
provide inconsistent information regarding a specific design or licensing requirement. In ,
other cases, design changes were not incorporated into existing calculations. This I
inconsistent information could impact the accuracy of a later operability evaluation or l
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design change, particularly with respect to the available margins assumed in the !
analysis. 1
i
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l The licensee's integrated review. team assessment (IRT) identified some calculation
l
deficiencies. The significance of the deficiencies was evaluated and entered into a i
corrective action tracking program. The deficiencies were identified as IRT open items. l
\
The NRC Team identified examples of calculation inadequacies and discrepancies ;
between the calculations, the UFSAR, and the AFW system design basis document. l
-
Examples were identified where the design basis of the AFW system was not !
i implemented consistently. The SDBD gives several values for single parameters such l l as minimum and maximum temperatures and volume of the water in the ECST. Some of l
the values are specific to a condition or accident, while others were the result of design l
l
changes that were not completely implemented in calculation affecting the AFW system. !
It was found that calculations affected by design changes were readily available from l
document control and did not include any revision or indication to show that they had i
l been superseded. Examples identified by the team of calculation inadequacies and ; l
discrepancies between various design documents are discussed below. :
NPSH Reauirements
The licensing basis for North Anna specifies that no operator action can be assumed to !
occur for 30 minutes after a postulated accident. Calculation numbers 19096.0510, M-1, .;
l Rev 1 dated October 5,1989 and ME-0441, Rev 0 dated July 27,1995 were reviewed to ! l determine if the net positive suction head available (NPSHA) to the turbine driven j j auxiliary feedwater (TDAFW) pump was sufficient for 30 minutes (prior to any operator i
action) after a main feedwater line break (MFLB) or a main steam line break (MSLB). l
l The team noted that the conclusions from these calculations were affected by the results ; j from calculation 19096.0510, M-4, Rev 0 dated November 17,1989. This calculation
.
!
was completed in 1989 when the licensee discovered an error in the orifice sizing
calculation which resulted in an increase of the TDAFW pump flowrate from 900 to 970 ,
;
,
gpm. The NPSHA calculations did not take into account the correct pump discharge
flowrate based on the installed orifice in the pump discharge piping nor the correct {
design basis storage tank water temperature.
:
In order for the team to determine the NPSHA, it was necessary to extrapolate the !
results from calculation 19096.0510, M-4. The team concluded that 30 minutes after a
'
MFLB or MSLB, the NPSHA for the TDAFW pump would be insufficient. Discussions ,
with licensee engineers and review of documentation disclosed that this problem was :
previously identified and was the subject of a justification for continued operation (JCO).
The JCO concluded that increased flow to the faulted steam generator did not adversely l
affect the accident analyses, and that the potentialloss of the TDAFW pump due to
! inadequate NPSH prior to operator intervention, would have no impact on the accident : l analyses. It further concluded that adequate NPSH would be available for the motor {
!
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driven AFW pumps, and that the TDAFW pump would be secured by the operator after ;
30 minutes at which time the TDAFW pump would no longer be required. )
:
- -
The team concluded that the potentialloss of the TDAFW pump in this scenario did not )
have an adverse impact on accident mitigation. Instead of revising the NPSH i
calculations to account for the increased TDAFW pump flowrate, the licensee relied on a !
memo from SWEC which contained the JCO. There was no link between the above
listed calculations and the JCO memorandum. The licensee's integrated review team ,
(IRT) did r.ot identify this discrepancy in the calculations. This was identified as an l
example of a weakness in the mechanical design calculations.
Backuo Water Source for AFW ,
-
!
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The licensing basis and SDBD specifies use of either the fire protection or service water
systems as a backup source of water to the AFW system. Paragraph 5.2 of the SDBD
l
and UFSAR Section 10.4.3.3 state that the fire protection and service water systems
provide an emergency backup supply of water to the AFW system. Makeup water from
the main condensate storage tank cannot be relied on since it is a non seismic, non
safety related tank. The team determined that no hydraulic calculations have been
completed to verify AFW performance from the backup (alternate) water sources. The
alternate sources of auxiliary feedwater, other than from the normal source of I
Emergency Condensate Storage Tank, cannot be flow tested since the ~ water quality
from these alternate sources is not preferred except in emergency conditions. The
design basis relies on the availability and the acceptability of these alternate sources. :
The team questioned licensee engineers regarding the basis that the alternate sources,
once aligned, will perform acceptably. The licensee responded that calculations are not i
required to demonstrate the flow capability of the alternate AFW supply sources since
the accident analyses assumed that the missile protected, safety related emergency !
condensate storage tank (ECST) was available to provide sufficient AFW supply to cool ;
the reactor down to 35* F (RHR entry point) in 6 hours or to maintain the reactor at hot i
shutdown for 8 hours, as determined in calculation SM-1152, Rev. O, including
Addendum A. The failure to complete a calculation to document the design basis of the ;
'
emergency backup water supply to the AFW system was identified as an example of a
weakness in the mechanical design calculations.
ECST Volume ;
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The team reviewed Calculation SM-1152, dated May 18,1998, and Addendums A and
B. This was the calculation of record which was prepared to demonstrate that the i
'
volume of water in the ECST was sufficient to remove total decay heat to maintain hot
standby conditions. However, this calculation failed to consider that during a MFLB with
the faulted steam generator aligned to the TDAFW pump, a condition would exist in
which a significant portion of the ECST water volume could be used such that a j
realignment to an alternate water source would be necessary. During a MFLB, l
approximately 50,000 gallons of the ECST volume would be used up in the first 30
minutes. Assuming that the turbine driven pump is aligned to the affected steam ;
generator, about 30,000 gallons of the 50,000 gallons would be lost from the break.
Therefore, the conclusions stated in the calculation that the ECST is available to provide
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14
sufficient AFW to cool RCS to 35' F, without relying on one of the emergency AFW
backup sources may not be substantiated in the case of a MFLB. However, the team
'
concluded that sufficient water volume was available in the ECST for the first 30 minutes
of the accident without requiring any operator actions to swapover to the backup
sources.
Calculation SM-1152, and addendum, was compared to calculation SM-0372.
Calculation SM-0372 demonstrates that the volume in the ECST is sufficient for '
maintaining safe shutdown. Calculation SM-0372 was available in the document control
system. There was no indication noted in this calculation that it was not the latest
calculation. Calculation SM-0372 showed that 8,445 gallons of water below the top of
the suction line to the turbine driven pump was not available (i.e., not useable in the
postulated event). Calculation SM- 1152 showed 6,910 gallons of water which was not
available (below the elevation of suction line) to the turbine driven pump. The reason for
the 1,535 gallon discrepancy was not immediately apparent. When questioned
regarding this discrepancy, licensee engineers were not able to explain the reason for
the differences. The team later determined that 1,535 gallons was the volume of water
between the elevations of the center of the suction line and the top of the suction line.
Licensee engineers acknowledged that this volume, which is more than twice the
available margin for ECST volume, would ie unavailable. The calculation also
accounted for approximately 3,500 gallons of unusable water associated with vortexing.
However, it was acceptable to consider vortexing to occur at the center of the suction
line because that is where the maximum flow velocity occurs.
The various discrepancies in the design documents, discussed above, were identified as
additional examples of a weakness in the mechanical design calculations.
Pipino Desion Pressure and Temperature
-
The team reviewed EWR 93-009 which raised the design pressure of the turbine driven
pump discharge piping from 1400 psig to 1480 psig. This was done to alleviate spurious
lifting cf the discharge relief valve RV-100/200 during pump surveillance testing at
recirculation flow. Calculation ME-435 implied that the entire AFW system downstream
of all three AFW pumps was rated at a design temperature and pressure of 100 degrees i
and 1480 psig respectively. However in discussions with licensee engineers, the team '
determined that piping between the motor driven pumps and the check valves were still ;
rated for design pressure of 1400 psi. It was not clear from the EWR change l
documentation and applicable calculations what the affected boundary was for elevating l
the system design pressure. Additionally, the review of the system P&lD showed that l
this drawing indicated that the relief valve was still set at 1400 psig. This was identified i
as a weakness.
ECST Water Temperature
A calculation, ME-0393, was completed to demonstrate that the temperature of the water
in the ECST would not rise above 100 F. The originator and engineer who performed
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the design verification for calculation ME-0393 failed to identify an error in the )
calculation which resulted in a rounding off error which reduced the calculated
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15
temperature by 3* F from the maximum calculated ECST water temperature. The
results of the calculation indicated that the water would not get increase above 94* F due
-
to the rounding off error. The correct value should have been 97* F. This error has no
safety significance, but was another example of a weakness in the licensee's
calculations.
In UFSAR section 10.4.3.1, the temperature of the feedwater supplied from the ECST
was assumed to be 120" F when considering feedwater enthalpy. Calculation SM 1152
with Addendum A and B used 110' F for the temperature of the water in the ECST when
considering enthalpy. When questioned about the discrepancy the licensee stated that
the design temperature of the ECST had been reduced from 120" F to
100* F by EWR 93-009 which was completed in 1993. The team noted that use of a
temperature value of 110* F when considering enthalpy was conservative since the
maximum ECST water temperature was 100* F.
The team concluded that failure to update the UFSAR with regard to the maximum
operating temperature of water in the ECST was an NRC-identified violation of 10 CFR
50.71(e). However, according to NUREG 1600, " General Statement of Policy and
Procedures for NRC Enforcement Actions (the Enforcement Policy), as revised on May
13,1998, the NRC may refrain from issuing a Notice of Violation when certain criteria are
met for issues considered old design issues. Discretion may be considered if, in the
Staffs view, the licensee would have identified the vialation inlight of the defined scope,
thoroughness, and schedule of the licensee's initiative (UFSAR review program). The
scope of the licensee's UFSAR review program is defined in a letter from VEPCO to the
NRC dated May 23,1997, on the subject of an integrated configuration management
program.
The inspectors also examined the methodology and data bases utilized in the UFSAR
review program, andlound that documentation was detailed and extensive. The
inspectors concluded the licensee had a good UFSAR review program. The inspectors
also noted that the UFSAR review program had not completed the AFW System at the
time of the inspection. In summary, the UFSAR discrepancy identified by the team did
represent a violation of the requirement to update the UFSAR. However, the NRC is
exercising discretion in accordance Section Vll.B.3 of the Enforcement Policy and
refraining from issuing a citation for this Severity Level IV violation.
The SDBD states the maximum ECST temperature for safe shutdown maintenance is -
120* F. Calculation SM 1152, with addendum, was the calculation which demonstrated
that the volume of water in the ECST is sufficient to maintain safe shutdown. This
calculation used 110' F as a design input for the water temperature in the ECST. The
licensee stated that the SDBD would be revised to clarify the maximum temperature of
the water in the ECST for safe shutdown maintenance.
UFSAR Table 10.4-1
Calculation ME-122, Rev 0, dated March 20,1987, used a design input value of 2941
feet for the head required for a flow of 340 gpm. Calculation 13075.01-29 established
the available margin for the turbine driven auxiliary feed pump and changed the head
. _ _ . , - _ ___ _ - . _ _ _ _ . . _ _ . _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ .
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requirement of 2941 feet to 2957 feet. However calculation ME-122 was never updated l
to incorporate this change. UFSAR Table 10.41, AFW System Design Basis, which 1
'
tabulated the pump head and head margin, also was not updated consistent with
changes made by calculation 13075.01-29. The team concluded this represented a
-
violation of 10 CFR 50.71(e), however, the NRC wi'll refrain from issuing a violation for
the same reasons stated in the above section on ECST Water Temperature.
1
Consideration of Waterhammer Loads in Pioina Analysis i
!
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During review of the corrective actions for the discrepancies identified in the AFW pipe ;
tunnels, the team noted that transient loads were not discussed in the operability i
l evaluations. Discussions with licensee engineers and review of design calculations
disclosed that the piping had been evaluated for deadicads, seismic loads, internal j
pressure, and thermal stresses. The team questioned the licensee why waterhammer ;
loads had not been included in the evaluations. The team noted inat UFSAR Section l
10.4.3.3 stated that the AFW piping had been evaluated for waterhammer loads. The !
licensee prepared Engineering Transmittal CME-98-0041 to evaluate the credibility of a I
water hammer event occurring in the AFW piping. The analysis showed that the piping ,
would not be subject to water hammer loads. Further review of the UFSAR and the
l
Safety Evaluation Report issued by NRC for licensing of the plant showed that the water i
hammer issues addressed by the UFSAR concerned potential transient loads occurring i
i in the steam generators, specifically the feedwater j-tubes on the sparger rings. The .
.
plants were licensed with License Conditions which required performance of specific l
tests to demonstrate that the steam generators would not be subjected to waterhammer i
loads. These tests were witnessed by NRC inspectors during plant startup testing. l
l !
During review of ET CME 98-0041_, the team identified a discrepancy in the analysis .
where an incorrect AFW flowrate was used to calculate the coastdown times for the AFW l
pumps. A value of 900 gpm was used. Calculation 19096.0510, M-4, Rev. O changed ,
, the flowrate from 900 gpm to 970 gpm. The team determined that using the correct j L '
value for flowrate (970 gpm) in the analysis resulted in a negligible change to the i
calculated AFW pump coastdown time. The team considered the use of the incorrect ;
j flowrate as an example of the potential consequences of failure to update the design ! !
documents to reflect the latest design basis information.
,
c. Conclusions i
;
The quality of AFW mechanical design calculations was poor. Numerous discrepancies !
were identified between various design calculations and design documents. Calculation l
quality was identified as a weakness.
Two examples of failure to update the UFSAR were identified. Discretion was exercised I
;
for this violation of 10 CFR 50.71(e) and was not cited (EA 98-500).
i I 4
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E3 Engineering Documentation (
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E3.1 AFW System Design Base Document
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! a. Inspection Scope (93809) i .
:
l The team reviewed the AFW system licensing basis and the system design basis i j ' document (SDBD). ! I' i
b. Observations and Findinas ;
j !
The licensee has initiated an integrated configuration management project. The purpose !
of this project is to perform a detailed review of the UFSAR, design basis documents,
5
and the improved technical specifications to improve design and licensing basis l
documentation. The overall approach is a complete validation of plant configuration j
documents, the UFSAR, and implementing procedures on a system by system basis. l
Each system is reviewed by an integrated review team comprised of engineering, ,
operations, and licensing personnel. Prior to this inspection the AFW system was j
reviewed by an IRT. ;
i
The current revision of the SDBD was revision 2, dated July 13,1998. This revision l
incorporated resolution of some of the deficiencies identified by the IRT during their !
. integrated assessment of the AFW system. Findings which have not yet been l
incorporated into_ Revision 2 of the IRT were listed as Open Items. The SDBD was !
comprehensive in its listing of design and licensing criteria and functional requirements l
for the system and equipment. However the poor quality of some of the design !
calculations impacted the reliability of the design references listed in the SDBD. In !
addition to the discrepancies in the SBDB listed in Paragraph E1.4 above, the team also !
identified the issues discussed below. !
, !
Paragraph 3.1 of the SBDB states that the ECST was provided with nitrogen as a means j
to control dissolved oxygen concentrations. The system P&lD, drawing number 11715- '
FM-074A, sheet 3 of 4, indicated a nitrogen connection connected to a manhole pipe I
connection as well as a vacuum breaker connection. Review of ECST tank drawing 1
11715-FV-43A-6 indicates a vacuum breaker connection, an atmospheric vent I
connection, and the manhole connection. No atmospheric vent connection is shown on -{'
the P&lD. The team could not ascertain from the P&lD and the tank drawing whether the
nitrogen was being supplied to the ECST or whether the tank was vented to atmosphere.
The licensee stated that the nitrogen was no longer in use and that the tank was an
atmospheric tank.
An apparent discrepancy appears to exist between the Pipe Specification NAS-1009,
Rev 18, dated 12-20-95, Pipe Class Sheet 601 and the AFW system design pressure &
temperature. Pipe Specification NAS-1009, Rev.18, Class 601 lists the permitted
temperature & pressure for pipe class 601 to be 200 degrees and 1250 psi respectively.
The P&lD invokes pipe class 601 for much of the AFW piping, specificaMy, the turbine
and motor driven pump discharge piping. It would appear that the maximum pressure
permitted by the specification is 1250 psi up to 200 degrees. Based on calculation ME-
1
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435, Rev 0, which establishes the system maximum temperature and pressure, the
current design temperature & pressure for AFW is 100 degrees and 1480 psi. It was i
'
determined that based on code allowables, the interpolated pressure and temperature l
limitation for the system piping and components was indeed 100/1480. However this !
interpolated temperature and pressure limitation was never specifically identified on the
'
.
pipe class sheet in the specification. Licensee engineers indicated that the pipe class
sheet would be revised to indicate the 100 degree pressure rating. The licensee's IRT
did not identify this discrepancy.
c. Conclusions ;
i
'
The SDBD was a comprehensive consolidation of design and licensing base critetia for
the AFW system. However, some deficiencies in design documentation references were !
identified. l
E3.2 Operability Evaluations t
a. Inspection Scope
!
The team reviewed the operability evaluation which the licensee performed to determine
operability of the AFW piping system for deficiencies identified with missing and f
degraded supports.-
;
b. - Observations and Findinas I
The licensee initiated DR -98-1881 to document and disposition a missing pipe hanger I
(gang support) which was to support seven individual pipe lines DR -98-2106 was -
'
initiated to document another missing gang support which was designed as a pipe
anchor for the three AFW supply lines. The team r6 viewed calculation numbers CE- ;
1416 and CE-1417 which were completea to determine operability of the piping systems !
for the deficiencies documented in DR -98-1881 and N-98-2106. !
;
Calculation CE-1416 was the piping stress reanalysis which was completed on July 24, j
1998, to evaluate the stresses in the piping for the missing supports. The stress !
reanalysis showed the allowable stresses in the piping were within code allowable
values. However the design loads exerted on the pipe supports were increased at four
supports by the revised stress analysis. The four supports were:
-
Support H-21 (Node 170, anchor for 3"-WAPD-13-601-Q3) l
l
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Support H-24 (Node 380,6"-WAPD-1-601-Q3) l
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Support FPH-WAPD-39-2 (Node 195, anchor for 4"-WAPD-39-601-Q3)
.
Support FPH-WAPD-39-2 (Node 150 for 4"-WAPD-39-601-Q3)
Calculation CE-1417 was completed on July 29,1998, to determine if the stresses in the
four supports were within code allowable values for the increased design loads due to
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19
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the missing supports. The calculation showed the stresses in three of the supports were
less than code allowable values. However, the stresses for Support (Anchor) H-21
-
exceeded code or design requirements. After further review, the licensee determined
- the support met the operability requirements of General Letter (GL) 91-18 " Resolution of
Degraded and Nonconforming Conditions and on Operability, Revision 1. Therefore,
the licensee determined that this support was operable, but degraded. The licensee will
modify the support to meet code or design requirements during the next refueling
outage.
The undersized welds, discussed in paragraph M2.2, above, and docu,uented on DR
l
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-98-2200 and DR -98-2204, were also evaluated in Calculation CE-1417. In response
to the undersized welds identified by NRC, the licensee randomly selected two supports ;
and completed as-built drawing to document measured as-built welds on the supports. ,
Calculation,
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!
! CE-1417 analyzed _ supports SHP-R-189 and -177 at local ccanections for the worst case !
conditions due to the undersized welds. These supports were selected based on field
!
reinspections performed by license engineers to document as-built conditions. The :
licensee concluded that the two supports were operable with the undersized welds. The
team concurred the results of the licensee's evaluations in both calculations. However, i
further review will be performed to determine the adequacy of the licensee's corrective -
actions conceming the undersized welds in followup on violation item 50-338,339/98-05- '
01. !
i
The team reviewed the extent of time for the licensee to complete operability evaluations ;
after the missing supports were identified, and the documentation of system operability. j
On June 16,1998, the missing pipe hanger which was to support seven pipe lines was ;
discovered to be missing. The operability evaluation for the affect of the missing hanger, l
documented in DR N 98-1881, stated that based on absence of concentrated loads and ;
low seismic acceleration values, the existing as-built configuration was acceptable and
l
would not result in any overstressed condition on the piping. It was stated on the DR that i
the Engineering Mechanics group would perform an analysis to verify that the pipes were ;
not overstressed. The stress analysis was not completed until July 24,1998. The i
support analysis which identified that a support was " degraded" was not completed until ,
July 29,1998, which was more than six weeks (43 days) after the initial missing support
was identified. The team reviewed the licensee's processes for performance of !
operability evaluations and concluded that they did not adequately define the need for
'
prompt operability evaluations, as outlined in NRC Generic Letter 91-18, Revisions 1 and ,
2. The team concluded that although the current process does not violate NRC :
regulations, it was a significant weakness in consideration that it permitted a delay of six [
weeks to determine that a support was operable but degraded. The documentation ;
associated with the DRs did not provide sufficient information to document that the piping .
systems were operable.
c. Conclusions
A weakness was identified in the licensee's process for completion of operability
evaluations. The final operability evaluation which determined the effect of missing
_eaMw4 a.= 's ,+=M e r4-4-4 -a48- 4 4a 42 A 1 hAamM 4 4- 4'e5 .--s-J4JB- a--A8-+--- d
. ,
i
20 !
)
supports on the operability of the AFW system was not completed until more than six l
weeks after the missing support was initially identified. j
'
E.7 Quality Assurance in Engineering Activities !
l
E7.1 Licensee Self Assessments
I
a. Inspection Scope (IP 93809) l
:
The team reviewed self-assessments performed within the engineering organization.
.I
b. Observations and Findinos _
The team reviewed administrative procedure VPAP-0104 which specified the licensee's l
'
program for performance of assessments by nuclear oversight, internal self-
assessments, and audits by external organizations. The team noted that the procedure i
specified that Deviation Reports were to be used to document discrepancies. However
l
the procedure contained little specific information on format of the assessment reports, i
how to define assessment findings which did not meet the definition for DRs, frequency ,
and scope of self-assessments, and followup actions.
The team also reviewed the results of 12 self-assessments performed within the I
'
engineering organization since 1996. Subjects covered by the self-assessments
included the environmental qualification of electrical equipment program, the Appendix R I
program, Regulatory Guide 1.97, drawing control, heavy loads, outages, system i
engineering, and USl A-46/IPEEE. After review of the self-assessment reports the team ,
identified the following issues: !
!
-
There was no-consistency in the report format used to document the self- l
assessment findings. For example, the self-assessment findings were
identified as strengths, weaknesses, short term improvements,
recommendations for improvements, recommendations for enhancement, i
recommended corrective actions, etc. The meaning of these terms was
not defined, and the use of the terms varied from report to report.
l
-
The self-assessments were performed in limited subject areas. It l
appeared that the licensee had not developed a plan for performance of ;
self-assessments which would address a broad subject area. !
!
'
-
Corrective actions to address self-assessment findings were frequently
delayed, some taking more than one year to complete. There appeared !
to be no followup action to determine effectiveness of corrective actions.
l
-
Only three DRs had been identified as a result of the self-assessments.
After the team had identified the above issues, the licensee informed the team that they
had performed an assessment of the overall self-assessment process which identified 1
many of the same issues. The team reviewed the report titled VPAP-0104 Self
1
J
. .
21
Assessment Program implementation, dated December 30,1997, which summarized
these findings. The conclusions of the report identified the following issues: inadequate
~
procedural guidance; inconsistent management expectations; and lack of pre-
established criteria for report findings. Four actions were recommended to improve the l
program, the most significant which was to revise VPAP-0104 to provide better
instructions for performance of the self-assessments. The due date for preparation of
the revised procedure was March 31,1998. As of this inspection date, the revised ,
procedure had not been issued. During a subsequent inspection, documented in NRC
Inspection Report numbers 50-338,339/98-06, an inspector follow-up item was identified
to review the effectiveness of the self-assessment process in a future inspection.
c. Conclusions _
t
A weakness was identified regarding the lack of progress in correcting deficiencies in the l
self-assessnw.,t process identified in December,1997. The self-assessment program
has not been effective.
'
V. MANAGEMENT MEETINGS
l
X1 Exit Meeting Summary l
l
The Team Leader discussed the progress of the inspection with licensee representatives ,
on a daily basis and presented the results to members of licensee management and staff *
at the conclusion of the inspection on July 31,1998, and during a telephone
conversation on September 14,1998. The licensee acknowledged the findings
presented.
PARTIAL LIST OF PERSONS CONTACTED ;
< ,
LICENSEE:
i
D. Christian, Vice President, Nuclear Operations
W. Corbin, Supervisor Test and Inspection ,
B. Foster, Superintendent, Station Engineering ,
J. Graf, Supervisor, Electrical Engineering
L. Hartz, Vice-President, Nuclear Engineering and Services '
J. Leberstien, Technical Specialist, Licensing
'
B. Leonard, Manager, Nuclear Engineering
E. May, Supervisor, Mechanical Engineering
W. Matthews, Site Vice-President ,
J. McCarthy, Manager, Nuclear Licensing and Operations Support !
D. Sommers, Supervisor, Configuration Management '
NRC:
M. Morgan, Senior Resident inspector }
R. Gibbs, Resident inspector .
t
i
,
.
-
_ -
_
.
22
,
LIST OF INSPECTION PROCEDURES USED
"
,
IP 93809, Safety System Engineering Inspection
LIST OF ITEMS OPENED
50-338, 339/98-05-01 VIO Failure to Coristruct Pipe Supports in Accordance
with Design Requirements - Section M2.2
,
50-339/98-05-02 VIO Failure to Correct Cause and Effects of Corrosion
of Unit 2 AFW Pipe Tunnel Supports - Section M2.2
50-338,339/98-05-03 VIO Fa;lure to include AFW Tunnel Pipe
Supports in the ISI Program - Section M2.2
50-338,339/98-05-04 IFl Diesel Generator Loading Transient
Analysis for the Delayed DBA Case -
Section E1.2
-
e
$
_ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ . . _ _ _ . . . _ _ _ _ _ . - . ___
. ,
;
.
i
23 ,
i
APPENDIX 1 ,
..
TABLE
!
AFW PIPE SUPPORTS INSPECTED :
Succort No. Unit Location Discrepancies '
SHP-R-187 1 Tunnel Most welds on connections between angles or
angles and wide flanges at the two upper box _
:
frames and on the right-top connection at Detail T
measured 1/8" or 3/16". Weld sizes specified on
3
the drawings were 1/4"
SHP-R-188 1 Tunnel Welds at locations inside and near pipes at the top ['
connections of the box frames for the three 3"
diameter pipes measured 1/8". Weld sizes
specified on the drawings were 1/4". i
SHP-R-189 1 Tunnel . Similar weld deficiencies were identified as noted ;
for SHP-R-188. Additional weld deficiencies were !
'
found at for the bracing connecting to top beam.
SHP-A-634 -2 Tunnel Support base plates, anchor bolts, nuts, and
support members were corroded up to a distance of
6" above the tunnel floor.
SHP-R-36 2 , Tunnel Same deficiencies as SHP-A-634
SHP-R-35 2- Tunnel Same deficiencies as SHP-A-634
SHP-R-34 2 - Tunnel Same deficiencies as SHP-A-634. A 2" diameter
pipe supported by the top beam was not shown on
the drawings. Vertical weld at top and left
connection for the left 3" diameter pipe measured
3/16"_ The weld sizes specified on the drawings
were 1/4"
SHP-R-15 2 Tunnel Welds measured 1/4" at right side of left post at
connection to ceiling, and two sides, front and left,
of right post connection to ceiling. Weld sizes
specified on the drawings were 5/16". Support
base plates, anchor bolts, nuts, and support
members at or near floor were corroded.
. _ _ . _ _- _ . _ .__. _ __. . _ _ _ __ _. _ _ _ _ .___. _ _._. _ _ _ .
. .
!
-
i
24 '
1
WAPD-R-198 1 Pump The weld on the top of the web at the left
l
connection between the wide flange and base plate !
-
measured 3/16". The weld size specified on the !'
~ drawing was 1/4". Two stiffener plates at the left
side of the left beam and the right side of the right
beam were not shown on the drawings. i
;
WAPD-R-22 . 1 Pump Welds at the connection between the' top of
bottom flange and base plate for the top j
beam measured 3/16". Weld sizes i
specified on the drawings were 1/4". Weld {
sizes were not specified on the drawings for_ j
connections between the insides of vertical . !
members and the horizontal beams. A gap !
existed between bottom of the pipe and the ;
supporting beam. The drawing specified a !
"zero" gap. i
i
WAPD-A-18 1 Pump Welds between the stiffener and base plates !
measured 3/16". Weld sizes specified on I
,
the drawings were 1/4" i
!
WAPD-R-24 1 Pump Welds at connections between the ri0ht and l
left outside the angles and wide flanges,
and connections between the right side of
l
the stiffener and base plates shown in
j
section 1-1 measured 3/16". The weld sizes .
specified on the drawings were 1/4" !
l- < ;
!
I
i
1
i
1 ,
?
l
t , e 1 !
!
f l
,- -. -, - _ _- .. -. - . -. - _
-.
_ . _ _ _ _ . . _ _ . . _ _ _ _ _ ..__.._-.-.._.-_._._...._._.._.___.m.__._-..-..-.-.
. -, i
l
l
!
25
6P.?ENDIX 2
~
'
LIST OF DOCUMENTS REVIEWED
Administrative Procedures
" VPAP-0104, Rev. 2, dated March 10,1997, NBU Management Station Self Assessment
Program
VPAP-0302, Rev. 7, dated February 3,1998, Station Drawing, Annotation, and Revision
~
VPAP-0304, Rev.1, dated January 24,1995, Request for Engineering Assistance i
VPAP-0305, Rev. 3, dated January 22,1997, Environmental Qualification of Electrical
Equipment Program
VPAP-0307, Rev. 7, dated August 29,1997, Repair and Replacement of ASME Section XI
Components
VPAP-1408, Rev. O, dated June 1,1995, System Operability
VPAP-1501, Rev.11, dated March 31,1998 Deviation Reports
VPAP-1601, Rev.10, dated March 31,1998, Corrective Action
VPAP-3001, Rev. 4, dated June 30,1998, Safety Evaluations
,
Alarm Response Procedures
I' 1(2)-AR-E-D8, Rev.1, dated July 1,1998, CN Tank 1,110,000 Gal., Lo/LO-LO Level l 1-AP-22.5, Rev 4, dated April 28,1995, Loss of Emergency Condensate Storage Tank 1-CN- l TK-1 l L Nuclear Desian Control Proaram Enaineerina Standards !
NDCM 2.1, Configuration Control
NDCM 2.2, Indoctrination and Training
NDCM 3.1, Design inputs
NDCM 3.3, Design Verification
. ,
NDCM 3.4, Engineering Change Requests j
l
- . _ _ . _ . , - . - - - ~ , , -. ___. -
- ;
.
, e
26
NDCM 3.5, Design Document Preparation and Revision
NDCM 3.7, Calculations
-
. .
NDCM 3.11, Technical Reports
NDCM 5.1, Document Control l
Er,aineerina Procedures and Standards
STD-EEN-0304, Rev. 2, dated December 14,1993, Calculating Instrumentation Uncertainties by
the Square Root of the Sum of the Squares Method
j
STD-GN-0001, Rev.16, dated May 7,1997, Instructions for DCP Preparation l
STD-CEN-0016, Rev,1, dated October 5,1990, Pipe Stress Analysis Standard l
l
!
Maintenance Procedures i
!
0-EPM-0302-03, Rev. 9, dated September 10,1996, BBC/ITE 4160-Volt Type SHK Breaker 9-
Year inspection
1-EPM-B-1816 01, Rev. O, dated May 18,1994, Functional Testing of Interlocks from Control
Circuits for Breaker 15H3, Auxiliary Steam Generator Feed Pump 1-FW-P-3A
1
1-EPM-R-1816-01, Rev.1, dated November 2,1994, Protective Relay Maintenance for Breaker
15H3, Auxiliary Steam Generator Feed Pump 1-FW-P-3A i
l
MDAP-0017 Revision 0, dated July 1,1990, Circuit Breaker and Associated Switchgear j
Maintenance Program (latest revision was also reviewed) '
Unit 1 MOV Switch Settings, Page FW-1, for valves 1-FW-MOV-100B and 100D; and same for
Unit 2
1-PT-71.10, Rev 19, dated August 7,1997,1-FW-P-2, Turbine Driven Auxiliary Feedwater
Pump and Valvo Test
Calculations
l
EE-0343, Rev.1, dated May 2,1994, Relay Settings for Safety Bus 1H, Attachment A.3, Page 3,
l for system ground relays coordination i
EE-0345, Rev. 2, dated May 2,1994, Relay Settings for Safety Bus 1H
, 4 l r i ! i
. -.. _
, . i
-
I
. j
27 ;
I
' EE-0557, Rev.0, dated March 25,1994, Evaluation of TOLs for North Anna Unit 1 Generic j
Letter 89-10 MOVs, also corresponding calculation for Unit 2 for 2-FW-MOV-200B and D i
-
EE-0100, Rev.1, dated October 28,1996, North Anna Emergency Condensate Storage Tank f
Level Uncertainty l
:
ME-0393, Rev 0, dated October 6,1993, Reduction of Auxiliary Feedwater System Piping l
Design Temperature From 12'F to 10*F
__
SM-1152, Rev 0, dated May 18,1998, Emergency Condensate Storage Tank Heat Removal l
Capacity
~
SM-1152, Addendum A, dated June 25,1998, Emergency Condensate Storage Tank Heat j
'
Removal Capacity -
SM-1152, Addendum B, dated July 21,1998, Emergency Condensate Storage Tank Heat
Removal Capacity - ,
l
19096.0510, M-1, Rev 1, dated October 5,1989, NPSH Available to the Turbine Driven AFW h
Pumps
.
ME-0441, Rev 0, dated July 27,1995, NPSH Available to the Turbine Driven AFW Pumps !
l Following A Main Steam Line Break j l 19096.0510, M-4, Rev 0, dated November 17,1989, Turbine Driven Auxiliary Feedwater Pump l Orifice Verification l 1
13075.01-29, Rev 1, dated October 31,1989, Available Margin for the Turbine Driven Auxiliary )
Feed Pump <
!
l
!
l ME-170, Rev 0, dated June 25,1988, AFW Pump Flowrate at PCV Fail Open [
: ;
ME-435, Rev 0, dated December 20,1994, Auxiliary Feedwater System Piping Design Pressure i
!
'
ME-437 Rev 0, dated' December 22,1994, Turbine Driven Auxiliary Feedwater Pump Discharge
Relief Valve Sizing and Setpoint '
SM-776, Rev 0, dated March 22,1991, Main Feedline Break Analysis for Degraded AFW. Pump
.
- ME-122, Rev 0, dated March 20,1987, Auxiliary Feedwater System Curve Determination
l ME-258, Rev 1, dated March 25,1992, Head Requirements for Motor Driven Auxiliary l
Feedwater Pumps %-FW-P-3A/3B !
9
;k
-
13075.35-03, Rev 0, dated August 26,1980, Low Level Alarm Setpoint in the Emergency
; Condensate Storage Tank
; ,
l SM-372, Rev 0, dated February 3,1986, ECST Maximum Allowable Temperature !
'
i,
1
.
;
?
a '
- -- - -
f
_ _ _ __ _
-
l
.
28 .
!
CE-1412, Rev 0, dated June 9,1998, Pipe Minimum Wall Thickness Requirements Due To
.
Longitudinal Stress For A Portion Of Several Lines Located in Unit 1 F pe Tunnel
CE-1416, Rev 0, dated July 27,1998, Review of Auxiliary Feedwater Piping for Missing Pipe
Supports identified in Deviation Reports N-98-1881 and N-98-2106
CE-1417, Rev 0, dated July 29,1998, Review of Auxiliary Feedwater Pipe Supports due to
Missing Supports and Weld Discrepancies 1
4
CME 98-0035, Rev 0, dated June 2,1998, Minimum Wall Thickness in AFW Piping
~
CME 98-0041, Rev 0, dated July 17,1998, Waterhammer Evaluation for Auxiliary Feedwater
System
l
)
I
Enaineerina Drawinas j
Stone & Webster drawing 11715-ESK-5AA, Rev.18, Elementary Diagram 4160 V Circuits l
Auxiliary Steam Generator Feed Pump 1-FW-P-3A '
Stone & Webster drawing 11715-FE-21T, Rev. 29, DC Elementary Diagram 4160 V Emergency
Bus 1H Undervoltage (for 27A power source) -
Stone & Webster drawing 11715-ESK-11D, Rev. 8, Elementary Diagram Loss of Reserve
Station Power (for 27C power source)
Stone & Webster drawing 11715-ESK-5Z, Rev. 9, Elementary Diagram 4160 V Circuits Steam
Generator Feed Pumps (circuit for start signal on main feed pump trip)
:
Stone & Webster drawing 11715-ESK-6MA, Rev.10 Elementary Diagram 480 V Circuits
Containment isolation Trip Valves Sheet 1 (close steam generator. blowdown valves)
Stone & Webster drawing 11715-ESK-6MB, Rev.11, Elementary Diagram 480 V Circuits
Containment isolation Trip Valves Sheet 2 (close steam generator blowdown valves)
Stone & Webster drawing 13075-ESK-10 BAN-A, Rev. 7, Elementary Diagram Annunciator
(circuit for AFW pump discharge PCV not open annunciator window) ;
Stone & Webster drawing 11715-ESK-6PR, Rev.17, Elementary Diagram Solenoid Operated
l
Valves - Sheet 40 (AFW pump turbine drive steam intet trip valve)
Stone & Webster drawing 11715-ESK-3E, Rev.10. Elementary Diagram Control Switch Contact
Diagrams - Sheet 5 (relates to ESK-6PR)
!
VEPCO drawing 11715-ESK-6CL, Rev.16 Elementary Diagram 480 V Circuits Motor Operated
Valves Sheet 11,01-FW-MOV-100B and 100D l
l
l
l
i
. .
29
VEPCO drawing 11715-ESK-6CL-2, Rev.14, Elementary Diagram 480 V Circuits Motor
Operated Valves Sheet 11-2,01-FW-MOV-100A and 100C
-
Stone & Webster drawing 11715-FE-90CA, Rev. 3, Appendix R Block Diagram Auxiliary
Feedwater System Sheet 1 (shows routing of feeder cables for AFW pump motors, used in
walkdown of cables)
VEPCO drawing 11715-FE-1BB, Rev. 9, One Line Diagram Electrical Distribution System (key
one-line diagram)
i
VEPCO drawing 11715-L-CN100A & CN1008, Condensate System Missile Protected
Condensate Storage Tank 1-CN-TK-1 Low Level Indicators and Alarm (loop diagram) i
!
VEPCO drawing 11715-H-FW100C, Rev. 7, Feedwater System Feedwater Control From l
Auxiliary Feed Header B" t's Steam Generator 1-RC-E-1C (loop diagram)
VEPCO drawing 11715-P-FW159A & FW159B, Rev. 5, Feedwater System Auxiliary Feed
Pumps Minimum Discharge Pressure Control (loop diagram)
VEPCO drawing 11715-TV-MS111 A, Rev. 9, Main Steam System Auxiliary Feed Pump Turbine
Drive Trip Valve TV-MS111 A Control (loop diagram)
i
VEPCO drawing 11715-FM-074A, Sheets 1 through 4, Flow / Valve Operating Numbers Diagram
Feedwater System
12050-FM-074A, Sheets 1 thru 4, " Auxiliary Feedwater System P&lD, Unit 2"
11715-FM-074A, Sheets 1 thru 4, " Auxiliary Feedwater System P&lD, Unit 1"
r
11715-FV-43A-6, "110,000 Gallon Condensate Storage Tank 1 & 2-CN-TK-1"
ITT Grinnell Mark No.1-SHP-R-187, Sketch No. 7053, Rev.1
i
Stone & Webster Sketch No.1-MFSK-4289A-3, Title SHPD-187, Rev. 3 )
ITT Grinnell Mark No.1-SHP-R-188, Sketch No. 7054, Rev.1
ITT Grinneil Mark No.1-SHP-R-189, Sketch No. 7055, Rev.1
Stone & Webster Sketch No.1-MFSK-42708, Details of New Base Plates for Hanger number 1-
SHP-R-189, Line 3 WAPD-11-601-Q3, Rev.1
ITT Grinnell Mark No.1-WAPD-R-191, Sketch No. 7057, Rev. 2
ITT Grinnell Mark No.1-WAPD-R-192, Sketch No. 7058, Rev. 2
ITT Grinnell Mark No.1-WAPD-R-193, Sketch No. 7059, Rev.1
._ . _ _ . _ . . . _ _ _ ._ _. -. _ . . _ - _ _. . _ . . . _ _ . _ _ _ _ _ _ . _ . . -
. .
1 '
'
:
30 i
VEPCO drawing 12050-PSSK-107DB.08, Rev.1, Supt. No. 2-SHP-A-634
-
,
ITT Grinnell Drawing Mark No. 2-SHP-R-36, Sketch No. 2-4235, Rev. 0
'
,
Stone & Webster Drawing Sketch No. ZFSK-2642-2,411 Modification on 2-SHP-R-36 Line 3"-
4
. wapd-409,411, &413-601-Q3, Rev. 2 -
'
'lTT Grinnell Drawing Mark No. 2-SHP-R-35, Sketch No. 2-4234, Rev.1
. '
Stone & Webster Drawing Sketch No. ZFSK-2637-2, Title Modification on 2-SHP-R-36 on 3"-
wapd-409-601-03 Safeguard Pipe Tunnel, Rev. 2
/ )
ITT Grinnell Sketch No. 2-SHP-R-34, Sketch No.2-4233, Rev. 0 l
l
VEPCO Drawing No.12050-PSSK-107DB.06, Rev.1, Support No. 2-SHP-R-34
ITT Grinnell Sketch No. 2-SHP-R-32, Sketch No. 2-4231, Rev. 0
l
l
VEPCO Drawing No.12050-PSSK-107DB.04, Rev.1, Support No. 2-SHP-R-32 -
!
l
ITT Grinnell Sketch No. 2-SHP-R-15, Sketch No.2-4214, Rev. O <
Stone & Webster Sketch No. ZFSK-2536, Rev. 2, Modification to 2-SHP-R-15 I
VEPCO Drawing No.12050-PSSK-107DB.18, Rev.1, Support No. 2-SHP-R-15
l
ITT Grinnell Sketch No. 2-SHP-R-16 Sketch No.2-4215, Rev. O
l
i
ITT Grinnell Sketch No. 2-SHP-R-18. Sketch No.2-4217, Rev. 0 '
Stone & Webster Sketch No. ZFSK-2551, Rev. 3, Modification to 2-SHP-R-18 & 2-WCMC-RH-
19
Stone & Webster Drawing No.11715-PSSK-102FA.3, Mark No.1-WAPD-R-19
Stone & Webster Drawing No.11715-PSSK-102FD.1, Mark No.1-WAPD-R-22,
Stone & Webster Drawing No.11715-PSSK-102FD.4, Mark No.1-WAPD-A-18,
Stone & Webster Drawing No.11715-PSSK-102FD.3, MK 1-WAPD-R-24
Stone & Webster Drawing No.11715-PSSK-102FB.1, MK 1-WAPD-R-29
Stone & Webster Piping Drawing No.11715-MSK-107G1, Unit 1 Yard Piping-North Reactor
Containment Rev. 5 -
,
Stone & Webster Piping Drawing No.11715-MSK-107G2, Unit 1 Yard Piping-North Reactor
Containment, Rev. 4
:
_ __. __ _. _ _. _ . _ _ . - - . _ . - - . _ - - - _ _ . - .-_ _ _ ._ -. _ _
q ,
+ ,
,. ;
;
-
\
!
31- !
i
Stone & Webster Piping Drawing No.'12050-MSK-107A1, Unit 2 Yard Piping - North Reactor !
Containment, Rev. 5
l
' !
Stone & Webster Piping Drawing No.12050-MSK-107A2, Unit 2 Yard Piping-North Reactor :
Containment, Rev. 5
!
'
.
Stone & Webster Piping Drawing No.11715-FP-2J, Steam Generator Auxiliary Feedwater
Lines, Rev.14 l'
!
Miscellaneous l
~
!
Unit 1 Control Board Log Readings for September 11,1997, Page 13, items 128 and 129 for !
ECST Level and Special Instructions for same
f
i
instrument Society of America-RP67.04, Part II, Recommended Practice, Methodologies for the j
Determination of Set Points for Nuclear Safety-Related instrumentation j
i
institute of Electrical and Electronics Engineers Std-741-1997, IEEE Standard Criteria for the
{
~ Protection of Class 1E Power Systems and Equipment in Nuclear Power Generating Stations !
:
f
SDBD NAPS-AFW, Rev 2, dated July 13,1998, " System Design Basis Document for Auxiliary I
Feedwater System" {
i
North Anna AFW Integrated Review Team Open items List.1998
l
NAS-1009, Rev 18, dated December 20,1995, " Specification for installation of Piping and !
Mechanical Equipment" !
< i
EWR-93-009, Rev 0, dated October 12,1993, * Revise Pressure Rating for Auxiliary Feedwater
l
Piping of Unit 1 & 2" ;
. -
Ultrasonic Thickness Record NDER 98-206, " Aux. F. W. Line",- Revision 4
..
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1
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1
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}}