ML20138J622
| ML20138J622 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 02/04/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20138J615 | List: |
| References | |
| 50-440-96-16, NUDOCS 9702070390 | |
| Download: ML20138J622 (9) | |
See also: IR 05000440/1996016
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U. S. NUCLEAR REGULATORY COMMISSION
REGION lll
Docket No.:
50-440
License No.:
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Report No.:
50-440/96016(DRS)
Licensee:
Cleveland Electric lliuminating Company
Facility:
Perry Nuclear Power Plant
Location:
P. O. Box 97, A200
Perry, OH 44081
Dates:
October 28,1996 through January 24,1997
Inspectors:
David Butler, senior Reactor Engineer
Doris Chyu, Reactor Engineer
Approved by:
Ronald Gardner, Chief
Engineering Specialist Branch 2
Division of Reactor Safety
9702070390 970204
ADOCK 05000440
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EXECUTIVE SUMMARY
Perry Nuclear Station Unit 1
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NRC Inspection Report 50-440/96016(DRS)
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This regional inspection reviewed the licensee's efforts to address Appendix R hot short
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vulnerabilities and corrective actions to violation 50-440/94006-06B. The following
strengths and weaknesses were identified:
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Enaineerino
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The licensee's initial evaluation of Information Notice 92-18 adequately utilized the
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information available at that time. However, during the August 1994 refueling outage,
the opportunity to identify MOV susceptibility to hot shorts was missed because the
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design effort did not consider Appendix R requirements. Since re-identification of this
concern in early 1996, the licensee has takers adequate steps to resolve this issue
(Section E2.1).
Plant Suooort
The licensee's identification of discrepant fire seal configurations was good. An
unresolved item was identified pending the licensee's evaluation of the qualification of the
as built fire seals (Section F2.1).
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Report Details
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jll. Enaineerina
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Engineering Support of Facilities and Equipment
E2.1
Motor-Ooerated Valve (MOV) Performance Followino a Control Room Fire
a.
Insoection Scoce
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On July 18,1996, the licensee identified that a postulated fire in the control room
could render several safe shutdown related MOVs inoperable due to hot shorts.
The licensee issued LER 50-440/96006, Revision 0, dated August 19,1996,and
Revision 1, dated December 2,1996, documenting this finding. The inspectors
reviewed the following documents:
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Licensee responses to information Notice (IN) 92-18, " Potential for Loss of
Remote Shutdown Capability During a Control Room Fire,"
Safe Shutdown Capability Report,
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Electrical drawing for proposed MOV modifications,
ONI-C-61, " Evacuation of the Control Room," and
Integrated Operating Instruction (101)-11, " Shutdown From Outside Control
Room."
b.
Observation and Findinas
On February 28,1992, the NRC issued IN 92-18. This IN described an unanalyzed
condition regarding fire protection and a plant's safe shutdown capability when
reactor operators were forced to evacuate the control room. This fire could cau e
hot shorts, such as short circuits between motor-operated valve control circuit
conductors and their control power source, to initiate spurious operation of certain
MOVr before the operators shifted control of the valves to the remote / alternate
shutdown panel. Motor thermal overload (TOL) protection may be bypassed, set
high cr set with a longer tripping time to allow for additional valve duty cycles
and/or reversing of the MOV during stroking. The IN identified that MOV torque
and limit switches would not electrically disconnect a stroking valve due to the hot
short bypassing the limit and torque switches. This had the potential to cause
mechanical damage to the valve and/or damage the motor.
in April 1992, the licensee's response to IN 92-18 concluded that the electrical
design would protect the MOV against hot short and/or physical damage to the
valves. The design used dual-element time delay fuses, which were normally in
circuit, for MOV electrical protection. These fuses were sized to provide motor
overload protection and to allow sufficient margin for MOV operation. In addition,
based on design stem factors and motor efficiencies available from the MOV
manufacturer during this time period, the licensee concluded that the stall thrust
values were smaller than the weak link
'it values. Therefore, a valve could
withstand a locked rotor condition withe
valve mechanical damage until the fuse
opened during the hot short condition. F .swever, the licensee did identify that
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several fuse current ratings could be lowered to optimize the opening time (15-20
seconds) during a locked rotor condition, if a fuse opened, operators could replace
the fuse at the MCC and regain control of the MOV.
During the August 1994 refueling outage, the licensee modified several MOVs to
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increase their operational capability in response to NRC Generic Letter (GL) 89-10.
In addition, several of the fuses identified in 1992 were replaced. At that time, the
licensee had obtained site-specific data for stem factors from MOV testing. Using
site-specific stem factors and the available motor efficiencies, the licensee
determined that most of the stall thrust values for the modified MOVs were still
smaller than the weak link limit values. For the modified MOVs which had thrust
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values greater than the weak link limit values, the licensee determined that the
pressure boundary was not violated if a stall condition occurred. However, the
licensee did not look at the functionality (if the valves could be manually operated)
of the valves after experiencing a stall condition. The MOVs were modified
according to approved design change processes; however, the licensee failed to
consider Appendix R program requirements when addressing GL 89-10.
In early 1996, the licensee started to reevaluate the applicability of IN 92-18 due to
hot shorts concerns that were identified at other plants. In addition, other utilities
had identified through their testing that motor efficiencies were higher than
predicted in the past. Perry recalculated the stall thrust values. Of 54 MOVs
required to shut down the plant during a control room fire,27 valves were
identified to be susceptible to hot short conditions. Of the 27 valves,21 valves
had stall thrust values greater than the new weak link limit values.
The licensee used an ASME code screening methodology and determined that 12 of
the 21 valves would have stall thrust values greater than the new weak link limit
values. The licensee reperformed the weak link analysis for the 12 valves using
realistic factors such as stem nut engagements, torque-to-thrust conversion factors,
temperature factors, etc. At the completion of this effort, eight valves still had stall
thrust values greater than the new weak link limit values. The affected valves
were:
1E12-F0023, Residual Heat Removal (RHR) Isolation to Reactor Vessel Head
Spray Valve,
1E12-F0024 A, RHR Return to Suppression Pool Isolation Valve,
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1821-F0019, Main Steam isolation Valve Before Seat Drain Valve,
1E51-F0063, RCIC Steam Supply Containment isolation Valve,
1E51-F0064, RCIC Steam Supply Containment isolation Valve,
1G33-F0004, Reactor Water Cleanup Suction Containment isolation Valve,
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1P57-F0015A, Safety-Related instrument Air (IA) Containment Isolation
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Valve, and
1P57-F0020A, Safety-Related lA Containment Isolation Valve.
The licensee planned to modify these valves during the next refueling outage
starting September 1997 or the next outage of sufficient duration. In addition, the
licensee provided guidance and training to the control room operators of actions to
be taken in case of a hot short condition. A review of the licensee's compensatory
actions was performed by the NRC on September 10,1996, as documented in NRC
Inspection Report 50-440/96006. That review concluded that the compensatory
measures were acceptable.
The potential spurious operation with mechanical damage to certain Appendix R
designated valves could result in the loss of safe shutdown capability during a
control room fire.10 CFR 50, Appendix R, Section Ill.G 2, required, in part, that
alternative shutdown capability be provided where the protection of systems whose
function is required for hot shutdown does not satisfy the requirements of
Section lit.G.2.10 CFR 50, Appendix R, Section Ill.G.3, required, in part, that for
cables or equipment that could prevent operation or cause maloperations due to hot
shorts of systems necessary to achieve and maintain hot shutdown conditions
which are located within the same fire area outside of primary containment be free
of fire damage. From August 1994 through July 1996, the licensee failed to
provide adequate protection for equipment necessary to achieve and maintain safe
shutdown conditions during a control room fire. The failure to meet Appendix R
requirements for alternative shutdown capability is considered an apparent violation
of 10 CFR 50, Appendix R, Sections Ill.G.2 and Ill.G.3 (eel 50-440/96016-01).
c.
Conclusions
The inspectors concluded that the licensee's initial evaluation of IN 92-18 was
based upon the use of fuses and at-that-time available design values for valve r. tem
friction coefficients. The inspectors determined that during the August 1994
refueling outage, the opportunity to identify MOV susceptibility to hot shorts was
missed. Although several MOVs were modified due to GL 89-10 issues, the design
effort did not consider Appendix R requirements. Since re-identification of this
concern in early 1996, the licensee has taken adequate steps to resolve this issue.
IV. Plant Suppgr1
F2
Status of Fire Protection Facilities and Equipment
F2.1
Qualification of Fire Seal Material
a.
Insoection Scone
On October 22,1996, licensee quality assurance personnel identified that
construction gap fire seals and compartmentalized fire seals were not installed as
depicted on design drawings and in the qualification test report. The inspectors
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toured several of the affected areas and reviewed Potential Issue Form (PlF)
No. 96-3243; Fire Test Configuration for BISKO Three Hour Fire Seal Report
No. 3001-03-B; Design Guide UOO1 and Specification SP-2000, " Penetration Seals,
Raceway Fire Barriers and Radiant Heat Energy Shields;" and associated design
drawings.
b.
Observations and Findinas
The quality assurance personnel identified through a routine audit that the
construction gap fire seals and compartmentalized fire seals were not installed
according to BISCO design drawing Nos. 4549-07A-005-2 and 4649-07A-015-3.
According to a general note on Drawing No. 4549-07A-005-2, construction gaps
wider than 4 inches shall be divided longitudinally by using marinite boards. The
licensee identified that the marinite boards for two construction gaps were installed
latitudinally rather than longitudinally.
In addition, according to Drawing No. 4549-07-015-03, for an opening of greater
than 6.25 square feet, the compartment framing materials were to be anchored to
the surrounding concrete walls. The framing materials mainly consisted of unistruts
and unistrut brackets. The licensee identified that the fire wall separating the
intermediate Building and the Fuel Building on elevation 574 was not
compartmentalized according to the drawing. The entire opening was divided into
smaller compartments and the framing materials were not anchored to the wall.
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The test configuration documented in BISCO Three Hour Fire Seal Report
No. 3001-03-B used a 48 by 48 by 12 inch concrete tesi slab with a 30 by 30 inch
opening. A 4 by 24 inch cable tray with 40 percent cable fill was installed in the
opening. The opening and cable tray were filled with BISCO SF 20 silicone foam to
a depth of 9 inches. The entire assembly was then subjected to a three hour fire
test and a hose stream test. The test results were satisfactory; however, this
tested configuration was different than the as-built configuration.
The licensee initiated a PlF and engineering analysis to evaluate the as-built
configuration. This item is considered an unresolved item (URI 50-440/96016-
02(DRS)) pending NRC review of the licensee's evaluation.
c.
Conclusions
The inspectors concluded that the licensee's effort in the identification of this
discrepancy was good.
F8
Miscellaneous Fire Protection issues
(Closed) 50-440/94006-06B(DRS).: Slow corrective actions for inoperable
Appendix R required emergency lights. The inspectors reviewed the open PlF list
and work requests for Appendix R required emergency lights and identified no
outstanding items. In addition, since December 1994, the licensee had assigned a
system engineer to oversee the Appendix R emergency light program. The system
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engineer rewrote several surveillance and maintenance procedures to provide better
guidance for determining operability of Appendix R lights.
The inspectors reviewed Periodic Test Instruction (PTI)-R71-P0003,4, 5, and 6,
"Self-Contained Emergency Lighting Unit Discharge Test," Revision 1. However,
the procedure did not contain a quantitative acceptance criterion for determining
battery operability. The vendor's recommendation for final battery terminal voltage
for a 12-volt emergency light battery was 87.5 percent (10.5 volts) of the initial
voltage following a 8-hour discharge test. If the result was less than 10.5 volts, a
vendor representative was to be contacted. The only acceptance criterion in the
procedures required that the lamps be verified "on" at the conclusion of the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
discharge test.
The licensee indicated that PTI-R71 series procedures were developed according to
EPRI Nuclear Maintenance Application Center (NMAC) guidance. In Section 5.3.2
of the NMAC document,it discussed the acceptance criterion as a verification that
all lamps were still energized at the end of the discharge test. The guidance also
took into effect that the low-voltage cut out protection was in the circuit to protect
the battery. The NMAC document indicated that if the low voltage cut out
actuated (the lamps would be off) before the specified test duration, a battery
capacity problem existed and should be corrected before the test was repeated, in
Section 6.1.2, the document discussed a low voltage cut out setpoint of
75 percent 15 percent (9.6 to 8.4 volts). However, discharging to this voltage
could result in cell reversal and battery damage. In both Secticas 5.3.2 and 6.0,
the NMAC document stated that the manufacturer's literature should be reviewed
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for specific guidance because the information in the NMAC document may not fully
address the requirements of a particular type or model of emergency lights. The
inspactors concluded the guidance set forth in the NMAC document was reasonable
for maintaining emergency lights.
The licensee determined that the ernergency lights were equipped with non-
adjustable low-voltage cut out protection. The setpoint for such protection was
based upon battery and relay coilimpedance, and circuit component tolerances. A
nominal setpoint for a lead acid battery was 8 to 9 volts. The vendor indicated that
a lead acid battery would become susceptible to cell reversal and potential damage
if the battery approached to 2.5 volts. Based upon the low-voltage cut out
protection, the inspectors concluded that the existing acceptance criteria for the
8-hour discharge test was acceptable.
The inspectors consider the corrective actions for this item to be adequate. This
item is closed.
V. Manaaement Meetinas
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Exit Meeting Summary
On October 31,1996 and January 24,1997, the inspectors presented the
inspection results to licensee management. The licensee acknowledged the findings
presented.
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The inspectors asked the licensee whether any materials examined during the
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inspection should be considered proprietary. No proprietary information was
identified.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
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D. Haviland, Civil / Structural Design Lead
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K. Jury, Compliance Supervisor
L. McGuire, Electrical Unit Supervisor
J. Perry, Lead Auditor, Quality Assurance
INSPECTION PROCEDURES USED
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IP 37551: Onsite Engineering
ITEMS OPENED, CLOSED, AND DISCUSSED
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Ooened
50-440/96016-01
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Appendix R required MOVs were susceptible to a hot short
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condition
50-440/96016-02
Differences between the as-built fire sea! configurations and
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tested configuration
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Glpsed
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50-440/94006-06B VIO
Inadequate Breaker Lubrication and Slow Corrective Actions
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for Inoperable Appendix R Required Emergency Lights
LIST OF ACRONYMS USED
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American Society of Mechanical Engineering
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Escalated Enforcement item
Electrical Power Research Institute
GL
Generic Letter
IN
information Notice
101
Integrated Operating Instruction
LER
Licensee Event Report
Motor-Operated Valve
NMAC
Nuclear Maintenance Application Center
PlF
Potential Issue Form
PTl
Periodic Test instruction
Reactor Core Isolation Cooling
Reaidual Heat Removal
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Unresolved item
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Attachment A
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D. C. Cook Calculation ENSM 961213AF, Revision O
Allowable Centrifugal Charging Pump Degradation
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1.
Please provide the basis for the assumption that the CCP miniflow paths are
isolated when the suction is aligned to the RWST.
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2.
Please provide the basis for the assumption that control valves ORV-200 and ORV-
251 are fully open.
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3.
Although this calculation accounts for pressurizer pressure instrument uncertainty,
it does not appear to account for the uncertainty in the instruments used to record
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the data. Please provide additional information regarding this issue.
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4.
Please provide the Unit 2 pre-1990 operability review results.
5.
Please provide additional information regarding piping configuration input into the
Proto-Flo code.
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Please provide additional information regarding fluid viscosity inputs into the Proto-
Flo code.
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7.
Please provide additional information regarding initial RWST level assumptions.
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8.
Please discuss the sensitivity of flowrate to developed head and how this was
factored into the calculation.
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Attachment B
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D. C. Cook Calculation RD-96-02, Revision 0
Offsite and Control Room Thyroid Doses From
Containment Bypass Associated With a Charging Pump in ECCS Mode
1.
Please provide the basis for the assumption that the leak persists for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Please provide additional documentation which supports the chosen operating point
of filtered and unfiltered control room inleakage used in the calculation.
3.
Please provide additional discussion of the purpose and effect of doubling the
"LEAKRATE" term in the code.
4.
Please discuss whether the contribution from ESF leakage was included in the
control room thyroid dose calculation.
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5.
Please provide RD-94-01, "Offsite Doses Due to FCCS Leakage."
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6.
Please provide RD-88-01, Revision 2, " Control Room Dose to Operators Following a
LOCA."
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