ML16154A637
| ML16154A637 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/26/1994 |
| From: | Chou R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16154A636 | List: |
| References | |
| 50-269-94-20, 50-270-94-20, 50-287-94-20, NUDOCS 9408090091 | |
| Download: ML16154A637 (10) | |
See also: IR 05000269/1994020
Text
ev REG&,
UNITED STATES
05
NUCLEAR REGULATORY COMMISSION
REGION II
o-
101 MARIETTA STREET, N.W., SUITE 2900
ATLANTA, GEORGIA 30323-0199
Report Nos.:
50-269/94-20, 50-270/94-20 and 50-287/94-20
Licensee:
Duke Power Company
422 South Church Street
Charlotte, NC
28242
Docket Nos.:
50-269, 50-270,
and 50-287
License Nos.:DPR-38, DPR-47,
and
Facility Name:
Oconee Nuclear Station Units 1, 2 and 3
Inspection Conducted:
June 27 - July 1, 1994
Inspector:
0
7 41
R. Chou
Date Signed
Approved by:
7
4
J.
lake, Chief
Date Signed
Ma eri als and Processes 'Section
En ineering Branch
Division of Reactor Safety
SUMMARY
Scope:
This routine, announced inspection was conducted in the areas of Inservice
Inspection, modifications, and previous open items involving safety-related
piping systems.
Results:
In the areas inspected, violations or deviations were not identified.
Two unresolved items and one inspector followup item were closed. The
licensee had weaknesses in documentation and quality control.
One support
drawing was not reverified to check its validity with the controlled document
within 14 days of installation. Two rewelds re-entered by the welder in the
weld log form on a later date were not reverified and resigned by the craft
supervisor and the welding inspector.
9408090091 940726
ADOCK 05000269
a
1.
Persons Contacted
REPORT DETAILS
Licensee Employees
- M. E. Bailey, Regulatory Compliance Engineer
- S. W. Baldwin, System Engineer
- D. W. Dalton, Mechanical Engineer
- J. M. Dave, Engineering Manager
- B. Dolan, Safety Assurance Manager
- J. W. Hampton, Vice President on Site
- R. Harris, System Engineer
- D. Kelly, Mechanical/Civil Engineer Supervisor
- F. Linsley, Mechanical Engineer
- B. L. Peele, Station Manager
Other licensee employees contacted during this inspection included
craftsmen, engineer, operators, mechanics, technicians, and
administrative personnel.
NRC Resident Inspectors
- P. Harmon, Senior Resident Inspector
G. Humphrey, Resident Inspector
L. Keller, Resident Inspector
- Attended preliminary exit Interview on June 30, 1994
- Attended final exit Interview on July 1, 1994
- Attended both exit Interviews
2.
Document Review on Inservice Inspection on Units 1 and 3 (73755)
At the time of this inspection, Oconee Unit 1 just restarted from a
refueling outage. During Inservice Inspection (ISI) this outage, the
licensee only had two inspection findings which required engineering
disposition.
Support No. 1-03A-1-0-439A-H23 was originally inaccessible for
inspection of items 1, 5, 6, 7, 8, 9, 10, 11 & 12 of the Bill of
Materials. Later, the licensee engineers and inspector reached the
support through an air duct. (Part of the support which could not be
inspected originally is inside of the air duct.) They still could not
inspect the support completely. Since a complete VT-3, Visual
Inspections could not be performed on this support, the piping system
was analyzed for the worse case scenario, which is, that this support
carries no load. This analysis demonstrated that the piping stresses
remained within design allowables, but loads on eight supports
increased. These eight supports were evaluated for the load increases
and all were found to remain within their design allowables. Based on
this evaluation, this support is acceptable for continued service
without additional inspection.
2
Support No. 1-53B-0-435B-DE067 was found to have a discrepancy as
documented in NRC Inspection Report 50-269, 270, 287/92-29. The
deficiency was that this support had a strut which was skewed beyond the
5-degree tolerance. The licensee issued Work Request 92054351 to repair
the strut during the previous outage and also requested a reinspection
to make sure that the problem is not reoccurring during this outage.
This support was reinspected by ISI examiner and was found it to be
acceptable.
The inspector also reviewed the engineering dispositions on Item Nos.
F1.01.140 & F1.01.141 for the Unit 3 ISI inspection during the last
refueling outage around February 1994. Both items are in 3B Steam
Generator support skirt. The bolting material and gusset plate welds
were found to have a "heavy accumulation of boron, rust, and debris."
The bolting material inspected is the anchor bolts for the support
skirt, which are partially embedded in concrete. There are 48, two-inch
diameter bolts around the skirt; the material specification for the
bolts is ASTM A-490. The licensee's engineers determined that the skirt
welds were acceptable and the corrosion on the bolts did not reduce the
material below the minimum thickness requirements due to the size and
quantity of the bolts. Work Request 94003674 was written to clean and
inspect the bolt during the next refueling outage.
No violation or deviations were identified in this area.
3.
Document Review on Modification on Unit 1 (73755)
The inspector reviewed two modification packages which were completed
during this outage; the two packages were NSM# 12921 and 12971.
a.
NSM # ON-12921, Rev. 0
This modification was classified as an Urgent Modification.The
purpose of this modification was to replace the existing carbon
steel piping and valves to the Turbine Driven Emergency Feedwater
Pump (TD-EFWP) Bearing Cooler Jacket with stainless steel
materials because the piping was seriously degraded. The
modification consisted of replacing the piping downstream of
1LPSW-136 to the TD-EFWP Bearing Cooler Jacket; replacing the
piping used for backup cooling water downstream of 1HPSW-248 to
the TD-EFWP Oil Cooler; replacing the carbon steel piping from the
High Pressure Service Water (HPSW) backup to the pump jacket; and
tie into a second, new connection on the 16" HPSW header. Low
Pressure Service Water (LPSW) will no longer service the cooler.
An orifice and flow measurement instrumentation were also added
downstream of 1LPSW-137 for bearing cooling flow measurement.
This modification included replacement of about 150 feet of
corroded 2" carbon steel piping and nine valves.
The construction and QC inspection record package was reviewed by
the inspector. The review included signatures and dates by
workers, craft supervisors, QC inspectors and managers for
3
materials and tools checked out, hold points, weld inspections,
installation completions, procedure approvals, etc. There were no
problems identified by the inspector.
b.
NSM #ON-12971, Rev. 0
This modification was to correct problems which the licensee had
identified in the Low Pressure Service Water (LPSW) piping on the
discharge side of each of the three High Pressure Injection (HPI)
pump motor bearing coolers. Four Problem Investigation Process
(PIP) reports had been generated and one additional problem was
found concerning this line. The following lists a brief
description of the problems which were to be corrected by this
modification:
(1) PIP #93-0801 - Piping downstream of coolers is non-seismic.
The LPSW piping immediately downstream of the HPI pump motor
bearing coolers is Duke Class G (non-seismic) and has a
normal back pressure of 35 psig. During a seismic event,
one train could be broken and cause the other two trains to
be inoperable. The disposition of this PIP will upgrade the
piping to Duke Class F (seismic) to a common discharge
(2) PIP #93-0868 - Piping design temperature is too low.
The design temperature of the LPSW piping immediately
downstream of the HPI pump motor bearing coolers is 1600F.
OSC-6015, "Operability Evaluation for PIP 0-093-0660,"
indicated that this temperature could reach approximately
200'F during accident conditions. This PIP requires this
line to be upgraded to at least 2000F.
(3) PIP #93-0694 - Low LPSW flow through the HPI motor coolers.
This PIP was written for blockage concerns in the LPSW
piping. The modification will resolve the problem.
(4) Existing Rotameters are unreliable and are easily clogged.
The existing Rotameter flow switches (1LPSFS 0009,
1LPSFSO010, 1LPSFS0011) are unreliable. As a result,
Operations has to perform a quarterly "bucket" test. The
modifications will install reliable flow switches.
(5) PIP #93-0695 - Piping upstream of 1LPSW-771 is non-seismic.
The piping upstream of valve 1LPSW-771 providing backup
cooling water to the HPI pump motor bearing coolers is Duke
Class G (non-seismic). A seismically induced break in this
4
portion of the line would prevent the coolers from receiving
the required cooling water flow. The modification will
install a check valve on the class F side of 1LPSW-771 to
eliminate this problem. The modification included:
(1) Replacing/upgrading the piping and associated pipe
supports
(2) Eliminating the unused flow control valves and
thermometers
(3) Replacing three existing rotameters
(4) Adding new flow switches with a flow orifice and a
QA-1 check valve (ILPSW-931)
The inspector reviewed the installation and QC inspection package.
A problem was identified on page 1 of 1, MP/O/A/1810/014,
Enclosure 13.3- Weld Log and Piping Surface Inspection Form. This
form initially contained two welds which had been completed and
signed off by the welders on April 24 & 26, 1994, and accepted by
a QC inspector. On the bottom of the form, required craft
supervisor and welding inspector were signed and dated April 26,
1994. On May 21 and 22, 1994, two rewelds were completed and
entered by the welders and accepted by the QC inspector. The
signatures and dates of craft supervisor and welding inspector on
the bottom portion of the form were not re-entered. This is
considered a weakness in the area of documentation and quality
control.
c.
Support Calculations Review
Six pipe support calculations from the above two modification
packages were randomly selected for review. The design
calculations were partially reviewed and evaluated for
thoroughness, clarity, consistency, and accuracy. The review
included formulas, theories, assumptions, displacements,.member
sizes, stress checks, weld sizes and symbols, bolt sizes, and
standard component capacity. In general, the design calculations
were of good quality. The calculations reviewed are listed below.
Support No.
Calculation No.
Rev. No.
NSM No.
1-14B-403A-H4191
OSC-1237-00-0033
0
12921
1-14B-403A-H4195
OSC-1237-00-0037
0
12921
1-14B-403A-H4199
OSC-1237-00-0037
0
12921
1-14B-435K-H5641
OSC-0967-14-0004
0
12971
1-GH-RS-7273-04
OSC-1619-01-1005
4
12971
1-14B-5100-NS-2004
OSC-1239-10-1015
3
12971
d.
Results and Conclusions
The modification packages and calculations inspected were
acceptable except for the weakness on documentation and quality
control as stated above.
No violations or deviations were identified.
4.
Special Event Review (92700)
PIP Serial No. 1-094-0866 was reviewed. This PIP described an instance
when the LPSW, a safety-related system, was overpressurized and caused
24 gauges to be over-ranged. The licensee suspected that the gauge
damage might be caused by the overpressure or by a water hammer.
After the licensee modified the portion of LPSW and removed a check
valve, the LPSW system was hydro-tested on May 22, 1994, before it was
returned to service. The LPSW system was pressurized by High Pressure
Service Water (HPSW) to a maximum pressure of approximately 115 psig
(HPSW pressure). Some of the 100 psig pressure gauges (the gauges are
capable of sustaining 125 psig) were broken. After disassembly of a
broken gauge, the cause of failure was determined to be the result of
failure of mechanical linkages due to system vibration.
The licensee's Instrumentation and Electrical (I&E) department tested
one of the gauges and determined that the deformation of the mechanical
linkages would require approximately 500 psig, which was well above the
115 psig applied. The licensee concluded that the LPSW hydro-test did
not result in over-pressurization of the LPSW system.
On May 31, 1994, a flow test of the LPSW was performed with a flow of
about 2000 gpm through the Reactor building Cooling Units (RBCU) and
three Rosemount flow transmitters were damaged. Per the manufacturer's
catalog, the maximum range of the Rosemount flow transmitters is 750
inches of H20 (about 4000 gpm) which is well above the test flow rate of
2000 gpm. It appears that the damage to the Rosemount flow transmitters
did not result from the LPSW system flow test.
On June 11, 1994, the 181 Reactor Building auxiliary cooler supply
header was found to be leaking and a 3" header end cap was separated
from the supply header after a loud noise was heard. All of the
pressure gauges on the 1BI auxiliary cooler were also found to be broken
at this time. Meantime, operations was performing PT/O/A/0160/03,
"Component Test of ES Channels 5 & 6."
It was suspected that the
performance of PT/O/A/0160/03 resulted in a water hammer in the
auxiliary coolers because the pressure in IB1
auxiliary cooler, due to
the higher elevation of the cooler, might be below the system vapor
pressure resulting in a water hammer due to vapor cavities.
6
On June 13, 1994, two of the Rosemount flow transmitters could not be
recalibrated and were replaced. A third Rosemount flow transmitter was
recalibrated; all of the Bailey flow transmitters could be recalibrated.
At the same time, the pressure gauges on the 1A, 1C, and 1D Reactor
Building auxiliary coolers were also found to be damaged and were
replaced. The licensee concluded, at that time, that a pressure
transient had not occurred in the LPSW system.
On June 15, 1994, insufficient LPSW flow on the "lA" decay heat removal
train was discovered. The key connecting the manual operator to the
valve stem was found to be sheared on 1LPSW-254. The failed key was
determined to be due to excessive stress. Therefore, the insufficient
LPSW flow shown above could have been the result of a water hammer.
On June 18 and 21, 1994, Reactor Cooling Pump (RCP) 1A2 and IBI cooling
coils were found to be leaking. The licensee concluded that the RCP
cooling coil leaks did not result from a water hammer in the LPSW system
because the valves were not in service during that period.
Based on the above events, the licensee suspected that PT/O/A/0160/03
might result in a water hammer. To determine if PT/O/A/0160/03 resulted
in a water hammer in the LPSW system, TT/1/A/0251/48, "ES channels 5&6
LPSW Functional Test" was performed to simulate the valve manipulations
of PT/O/A/0160/03. Upon performing TT/1/A/0251/48, there were no
indications of a water hammer.
Therefore, the licensee concluded that the damage to pressure gauges was
due to vibration and would bring vibration experts to investigate the
problem. The root cause of the damaged Rosemount transmitters and the
leaking 1B1 auxiliary supply header is still unknown. A possible
explanation is that the damaged components may have been subjected to an
air pocket. The licensee will continue to observe the system during
operation.
The inspector agreed with the licensee's decision to continue to observe
the system operation. No violations or deviations were identified in
this area.
5.
Previous Open Items (92701)
a.
(Closed) Unresolved Item (UNR) 50-269,270,287/91-05-01, Improper
Gap Between Washer and Snubber Rod Bearing
Excessive gaps were found between washers and rod bearings for
snubbers and sway struts during a previous inspection. A %" gap
on Support No. 3-07A-6-0-2400A-H72 was found. The licensee did
not inspect the gaps and did not have a procedure to inspect them.
Conceivably, with an excessive gap, the snubber or sway strut
movements, or impact due to the pipe movement, could cause damage
to the rods or pins. Per the manufacturers catalog the
construction tolerance for the gap should be 1/16-inch.
7
The inspector discussed the gap problem with the licensee
engineers and reviewed the information provided. The licensee
sent a letter on March 8, 1993, to Grinnnell Corporation, 1341
Elmwood Ave. Cranston, RI 02910 and requested guidelines or an
acceptable gap tolerance for inspection. The Grinnell Corporation
provided the following reply to the licensee in a letter
FB/V/3163D, dated April 16, 1993:
The gap between pipe clamp halves is a concern when
evaluating potential damage to the rods or pins in the area
of the spherical bearing and washer.
The functions of the gap between pipe clamp halves is to
accommodate the spherical bearing and to permit the rod eye
to rotate in a 100 cone of action. If the gap is too large
the bending stress in the stud becomes unacceptable.
Additionally, there is a concern that in the unlikely event
of bearing dislodgement, a large gap would permit
unacceptable movement of the bearing.
It is the opinion of Grinnell Corporation that once our
clamps are properly installed with all nuts locked
(particularly the load stud nuts), normal operating
conditions will not affect the gap ("S" dimension) of the
clamp. Any overloading condition would warrant a
reinspection of the clamp.
The concern that an excessive gap between the spacer washer
and spherical bearing could cause damage to the rods or pins
is not shared by Grinnell.
From the above reply to the licensee, the Grinnell Corporation
implied that excessive gaps could result in damage to the support
due to failure of the stud or through bearing dislodgement; but
they failed to specify what gap size could be tolerated.
Per discussion with the licensee engineers, .the inspector learned
that the licensee will not establish inspection tolerances for the
gap, since the manufacturer will not provide a definitive
inspection tolerance. The inspector will close this item and
refer the question to NRR.
b.
UNR 50-269, 270, 287/91-23-01, Maintenance Corrective Action
Program
An investigation process for the failure of the internal threads
on one of the anchor heads during the concrete tendon surveillance
was not initiated and the failure was not considered to be an item
reportable to the NRC. The licensee did not initiate a root cause
investigation based on the non-conformance program to investigate
the failure of the internal threads. Later, the licensee
engineers indicated that this problem would be written in Station
8
Directive (SD) 4.5.5, Problem Investigation Process, which is a
lower tier program. The more significant non-conformances were
handled by a Problem Investigation Report (PIR). The inspector
also found that this event would not be a reportable item to NRC
since the tendon was evaluated to be acceptable and operable.
The inspector discussed the problems with the licensee engineers
and reviewed the information provided. The licensee rewrote the
Problem Investigation Report (PIR) to combine all other procedures
into it and to include all non-conformances. The new PIR is
divided into four categories from the least significant non
conformance to the most significant non-conformance and were
classified and evaluated by the same people in each displine.
Based on the actions taken by the licensee, this item is
considered closed.
c.
Inspector Followup Item (IFI) 50-270/93-16-01, Pipe Support
Related Defects Found During Inservice Inspection
This matter identified three problems to be followed and the
corrective actions reviewed. The problems were a bent rod,
standing water in the Quench Tank, and a base plate gap between
the base plate and concrete.
The inspector discussed the problems with licensee engineers and
reviewed the information provided. The bent rod was replaced.
The standing water disappeared toward the end of outage during the
licensee reinspection. A shim was installed between the base
plate and concrete per Work Request 93027412 and Work Order 94024506-01.
The inspector reviewed the construction and QC inspection package
for the above work request and work order. A minor deficiency was
found in that the support drawing attached to the package for
construction was not validated within 14 days of the construction.
The drawing was initiated, obtained, and verified from the
document controlled area on March-9, 1994;-the-work was completed
between April 12 and 14, 1994. The support drawing was not
reverified and revalidated within 14 days of April 12, 1994. The
inspector considered this problem to be a minor weakness in
document control, therefore, this IFI is considered closed.
6.
Exit Interview
The inspection scope and results were summarized on June 30 and July 1,
1994, with those persons indicated in paragraph 1. The inspector
described the areas inspected and discussed in detail the inspection
results listed below. Proprietary information is not contained in this
report. Dissenting comments were not received from the licensee.
(Closed) Unresolved Item 50-269, 270, 287/91-05-01, Improper Gap Between
Washer and Snubber Rod Bearing (paragraph 5a).
9
(Closed) Unresolved Item 50-269, 270, 287/91-23-01, Maintenance
Corrective Action Program (paragraph 5b).
(Closed) Inspector Followup Item 50-270/93-16-01, Pipe Support Related
Defects Found During Inservice Inspection (paragraph Sc).