ML20129E979
| ML20129E979 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/27/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20129E950 | List: |
| References | |
| 50-440-96-08, 50-440-96-8, NUDOCS 9610040013 | |
| Download: ML20129E979 (20) | |
See also: IR 05000440/1996008
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Docket No:
50-440
License No:
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Report No:
50-440/96008(DRS)
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Licensee:
Cleveland Electric Illuminating Company
facility:
Perry Nuclear Power Plant
Location:
P. O. Box 97, A200
Perry, OH 44081
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Dates:
August 26 through September 11, 1996
Inspector:
E. Duncan, Lead Engineering Assessment Person
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Approved By:
M. A. Ring, Chief
Lead Engineering Branch
Division of Reactor Safety
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9610040013 960927
ADOCK 05000440
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EXECUTIVE SUMMARY
Perry Nuclear Power Plant
NRC Inspection Report 50-440/96008
This inspection report contains the findings and conclusions for a special
inspection conducted from August 26 through September 11, 1996 to review
Licensee Event Report (LER)93-021, " Loss of Safety Function for Emergency
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Closed Cooling System A," and LER 94-005, " Loss of Both Trains of Control Room
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Emergency Recirculation Due to Low Emergency Closed Cooling Temperature."
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Enaineerina
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From June 2 to July 2,1993, two apparent violations of technical
specifications occurred. One of these involved a 45 hour5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> period when
the Emergency Closed Cooling (ECC) system was inoperable, and the
actions required by Technical Specification 3.0.3 were not completed.
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Identification of long-term corrective actions to address excessive
leakage of motor-operated butterfly valve OP42-F295A was slow and failed
to include periodic leak rate testing as required by the inservice
testing program. An apparent violation of 10 CFR 50, Appendix B,
Criterion XVI was identified.
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LER 93-021 failed to provide an adequate assessment of the safety
consequences and implications of the event. A violation of 10 CFR 50.73
was identified.
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Corrective actions to address the inability to operate the control room
complex chillers under certain environmental conditions were slow and
failed to fully evaluate the design requirements of the ECC system
against all potential plant configurations, including those present
during periodic testing. The failure to perform an adequate post-
modification test following installation of a design change to address
the problem was a contributing factor. An apparent violation of 10 CFR 50, Appendix B, Criterion XVI was identified.
Summary of Open Items
Violations:
Identified in Sections E1.1.b.7, E1.1.b.8, and E1.2.b.7
Unresolved Items: None Identified
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Inspector Follow-up Items: None Identified
Non-cited Violation: None Identified
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INSPECTION DETAILS
III. Enaineerina
El
Conduct of Engineering
El.1 LER 93-021. " Loss of Safety Function for ECC System 6"
a.
Insoection Scope
The inspector reviewed LER 93-021, " Loss of Safety Function for
Emergency Closed Cooling (ECC) System A."
b.
Observations and Findinas
1.
Description of the Event
On July 1,1993, during testing to verify ECC "A" heat exchanger
performance,10-inch butterfly motor-operated valve (MOV) OP42-F295A was
identified as leaking 250 gallons per minute (gpm). An initial review
also determined that the allowable ECC "A" train leakage was 6.6 gpm.
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The valve limit switches and mechanical stops were subsequently re-
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adjusted on July 2, and a condition report (CR) was written to document
the event.
On December 17, the licensee determined that maintenance was last
performed on the valve on March 19, 1993, and therefore the leakase
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existed from March 19 until July 2.
A review of operating logs also
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determined that the division 2 diesel generator (DG) was out-of-service
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(00S) from June 14 to June 15 for about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />.
Given a loss-of-offsite-power (LOOP) with the division 2 DG 00S, the
OP42-F295A valve would fail to close on a loss of power and the "B"
train would also fail to function due to the loss of the division 2 DG.
Given the leak rate of the OP42-F295A valve, within 1 minute of receipt
of the surge tank low level alarm, the ECC "A" surge tank and suction
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standpipe would have drained to a point where the ECC "A" pump would
cavitate and fail. This scenario would result in a complete loss of
ECC.
2.
Emeraency Closed Coolina (ECC) System Description
The purpose of the ECC system is to provide safety-related components
with a reliable source of cooling water, ensuring proper operation of
these components during normal, as well as accident conditions.
The ECC system is divided into two independent loops designated loop A
and loop B.
Each loop consists of a pump, heat exchanger, 660-gallon
surge tank, motor-operated valves, and interconnecting piping.
The ECC
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heat exchangers are cooled by the Emergency Service Water (ESW) system.
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The ECC system consists of two identical loops.
Each ECC pump takes a
suction on its loop suction header and discharges through the shell of
an ECC heat exchanger to the system loads. After serving the system
heat loads, the cooling water returns to the pump suction. The suction
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header of each pump is provided with a surge tank to ensure that an
adequate suction head is available to facilitate system fill and make-
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up.
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The ECC system provides cooling to the following loads:
Looo A
Low Pressure Core Spray (LPCS) Room Cooler
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Reactor Core Isolation Cooling (RCIC) Room Cooler
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Residual Heat Removal (RHR) "A" Room Cooler
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Residual Heat Removal "A" Pump Seal Cooler
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Hydrogen Analyzer Cooler "A"
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Control Complex Chiller Unit "A"
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Looo B
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Residual Heat Removal "B" Room Cooler
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Residual Heat Removal "B" Pump Seal Cooler
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Residual Keat Removal
"C" Room Cooler
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"C" Pump Seal Cooler
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Hydrogen Analyzer Cooler "B"
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Control Complex Chiller Unit "B"
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During loss-of-Offsite-Power (LOOP)/ Loss-of-Coolant Accident (LOCA)
conditions, cooling water supply to the control complex chillers is
transferred from the Nuclear Closed Cooling (NCC) system to the ECC
system. Automatic isolation between the NCC and ECC systems is provided
by valves OP42-F295A/B and OP42-F290.
3.
Seauence of Events
7/1/93
During performance of Periodic Test Instruction (PTI) P42-
P0001, " Emergency Closed Cooling System A Heat Exchanger
Performance Test," while transferring the "A" control
complex chiller cooling water supply from NCC to ECC, the
"A" ECC surge tank overflowed.
7/1/93
Temporary Instruction (TXI) TXI-162 was written to determine
the source and rate of valve leakage.
7/2/93
Motor-operated butterfly valve (MOV) OP42-F295A was
identified as leaking in excess of 250 gallons per minute
(gpm). Work Order (W0) 93-2502 was initiated to determine
and correct the cause of the 250 gpm ceat leakage.
Subsequent investigations under WO 93-E502 identified OP42-
F295A to be stroking past its full closed position as a
result of mis-positioned limit switches and mechanical
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stops. The limit switches and mechanical stops were re-
adjusted in accordance with GEI-014, "Limitorque
Limit / Torque Switch Adjustment." The valve was
satisfactorily re-tested per surveillance instruction (SVI)
SVI-P42-T2001, "ECC System Pump and Valve Operability Test."
In addition, TXI-162 was then re-performed and verified no
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significant seat leakage existed.
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7/2/93
Field Clarification Request (FCR) 17843 was initiated and
requested the allowable ECC system leakage rate.
Based on
initial indication of a problem upon receipt of an ECC surge
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tank low level alarm and a 15-minute operator response time
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to establish makeup from the safety-related emergency
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service water (ESW) system, the maximum allowable OP42-
F0295A valve leakage rate was determined to be 6.6 gpm.
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7/6/93
Condition Report (CR)93-132 was initiated to document the
"as-found" 250 gpm leakrate on OP42-F295A.
Shift Supervisor
review section of CR 93-132 notes no Technical Specification
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(TS) violation.
Remarks section of CR 93-132 indicates that
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the Licensing and Compliance Section (LCS) to determine
reportability.
7/7/93
CR 93-132 review by LCS concluded that the event was not
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reportable and that ECC could have performed its intended
function if required.
Subsequently, a request to a
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contractor, Gilbert / Commonwealth, was initiated to evaluate
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the effects of the 250 gpm leakage condition of OP42-F295A
for reportability.
8/25/93
CR 93-132 root cause investigation completed and forwarded
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for approval.
Root cause investigation incorrectly
concluded that maintenance had not been performed on the
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valve since initial plant construction and that instructions
for initially setting MOVs did not contain sufficient detail
to adequately perform the activity.
12/1/93
CR 93-132 approved.
12/15/93
Gilbert / Commonwealth evaluation completed and concluded that
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the condition was reportable because it would have resulted
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in a condition that could have prevented the fulfillment of
the safety function of the ECC system.
In the event of a
LOOP with a single active failure of the division 2 diesel
generator (DG), ECC "A" would have lost fluid at a rate
greater than 250 gpm. Within 1 minute the ECC "A" surge
tank and suction standpipe would have drained down to a
point where the ECC "A" pump would begin to cavitate on low
net positive suction head (NPSH). Within a few minutes, the
ECC pump would probably fail from cavitation.
In addition,
with a LOOP and a division 2 DG failure, ECC "B" would also
be inoperable, resulting in a loss of the entire ECC system.
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In addition, Gilbert / Commonwealth reviewed the licensee's
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allowable leak rate calculation and extended the operator
response time from 15 minutes to 30 minutes in accordance
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with ANSI /ANS 58.8 - 1984, " Time Response Design Criteria
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for Nuclear Safety-Related Operator Actions." As a result,
the allowable leak rate was reduced from 6.6 gpm to 3.3 gpm.
12/17/93
A review of the work history for OP42-F295A as part of a
compliance review of CR 93-132 identified that the limit
switches and mechanical stops had been adjusted on March 19,
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1993 under WO 92-0495 contrary to CR 93-132 findings. A
post-maintenance test which consisted of manual and remote
valve strokes to ensure proper operation with no binding was
performed satisfactorily.
A leak rate test was not
performed.
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Since no other work had been performed on the valve until
the date of this event, OP42-F295A was considered to have
been leaking from March 19, 1993, to July 2, 1993.
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12/23/93
The allowable ECC "A" system leakage was formally revised
from 6.6 gpm to 3.3 gpm.
12/30/93
The licensee determined that the event was reportable since
the inoperable valve could have prevented the fulfillment of
the safety function of the ECC system. A 10 CFR 50.72
report was initiated.
1/23/94
LER 93-021, " Loss of Safety function for Emergency Closed
Cooling System A," was issued in accordance with
4.
Root Cause
The licensee attributed the root cause of this event to per sonnel error
in incorrectly setting the limit switches and mechanical stops for OP42-
F295A on March 19, 1993.
5.
Safety Sionificance
In the event of a LOOP /LOCA with a single active failure of the division
2 diesel generator (DG), ECC "A" would have lost fluid at a rate greater
than 250 gpm. Within 1 minute of receipt of the low level alarm on the
ECC "A" surge tank, the tank and suction standpipe would have drained
down to a point where the ECC "A" pump would begin to cavitate on low
net positive suction head (NPSH).
Within a few minutes, the ECC pump
would probably fail from cavitation.
In addition, with a LOOP and a
"B" would also be inoperable, resulting in a
loss of the entire ECC system.
The inspectors also determined that from June 2, 1993, to July 2, 1993,
Unit I was in operational condition 1, 2, or 3 and vulnerable to the
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event as described above.
In addition, from June 14 to June 15, for
about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, the division 2 DG was 00S.
Therefore, in the event of a
LOOP /LOCA during this period, the ECC system and the systems supported
by ECC may have been unable to mitigate the consequences of the
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accident.
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6.
Licensee Corrective Actions
As part of the licensee's immec kte corrective actions for this event,
the limit switches and mechanical stops were re-adjusted on valve OP42-
F295A and post-maintenance leak testing was performed to verify that the
valve would function as required.
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The following long-term corrective actions were also accomplished:
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An engineering evaluation was. initiated to determine if the
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potential for an incorrectly set butterfly valve limit switch or
mechanical stop on other systems could cause a similar problem.
As a result of that review, the licensee determined that although
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incorrectly setting limit switches could cause leakage in other
systems, none were identified that would cause an adverse impact
on safety.
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A review of mair.tenance instructions and vendor information was
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performed to determine a positive means to identify when a
butterfly valve is full open or full closed in order to ensure
that the limit switches and mechanical stops are properly set.
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As a result of that review, the licensee determined that a post-
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maintenance seat-leakage test is the only positive method to
ensure that butterfly valve limit switches and mechanical stops
are properly set. Therefore, GEI-0014, "Limitorque Limit / Torque
Switch Adjustment," was revised on January 27, 1996, to require
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that a post-maintenance leakage test be performed on those
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butterfly MOVs which have an established seat leakage limit.
In
addition, GEI-0014 was revised to clarify instructions for setting
limit switches and checking mechanical stop nut settings.
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A review of safety-related butterfly valves was performed to
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determine if further adjustment or testing was required to verify
proper butterfly valve closure.
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As a result, the licensee determined that OP42-F295A/B and OP42-
F325A/B fell into a unique category to isolate ECC to prevent
draining of a surge tank and required that a leak rate criteria be
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established. All other safety-related butterfly valves were
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evaluated and the licensee concluded that leakage during an
accident was not a safety concern.
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Training was provided to maintenance and system engineering
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personnel on this event and the necessity of verifying proper
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butterfly valve closure.
7.
Inspector Review
The inspectors reviewed the events and corrective actions discussed
above. The following concerns were identified:
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a) Corrective Actions
The inspectors reviewed the corrective actions taken as a result of this
event as well as the priority placed on fully understanding the
implications of the event to develop the corrective actions.
The
following issues were identified:
1)
Corrective Action Timeliness
The inspectors concluded that following the overfill of the ECC "A"
surge tank, the subsequent identification and repair of OP42-F295A was
prompt and complete.
The inspectors also concluded that the licensee had missed prior
opportunities to identify ECC system operability concerns and initiate
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corrective actions in a more timely manner.
These opportunities
included:
Condition Report (CR)93-132 which documented that butterfly valve
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OP42-F295A was leaking in excess of 250 gpm was initiated on
July 6,1993, although the condition was identified on July 2,
1993.
The inspectors concluded that the licensee's action to
document this significant condition adverse to quality was slow.
The initial Licensing and Compliance review of CR 93-132 on
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July 7,1993, concluded that the event was not reportable and that
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ECC could have performed its intended function if required,
although a calculation was performed on July 2 which indicated an
allowable leakrate of 6.6 gpm. The inspectors concluded that the
licensee's review of this event was slow and the conclusions were
incorrect although information with regard to leakage limits was
available.
The root cause investigation for CR 93-132 was not completed until
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August 25, 1993, and was not reviewed until December 1, 1993.
The
inspectors concluded that the investigation and review process was
not timely.
The results of an external followup review of the safety
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significance of the event was not completed until December 15,
1993, when the licensee received the report from Gilbert /
Commonwealth documenting their conclusions. The inspectors
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concluded that the followup review of the event, which required
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greater than 5 months to complete, was not timely.
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The Gilbert / Commonwealth report indicated that the event was
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reportable because it would have resulted in a condition that
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could have prevented the fulfillment of the safety function of the
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ECC system.
However, a 10 CFR 50.72 report was not made until 15
days later on December 30, 1993.
In light of the implications of
this event, the inspectors concluded that the licensee's review of
the Gilbert / Commonwealth report and subsequent NRC notification
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was slow.
The inspectors concluded that the licensee's response to the event was
not timely and negatively impacted the corrective action process.
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2)
Corrective Action Scope
The inspector concluded that the licensee's corrective action to perform
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post-maintenance seat-leakage testing following maintenance on butterfly
MOVs was a good initiative.
However, the inspectors noted that although valves OP42-F295A/B and
OP42-F325A/B had specific maximum allowable leak rate limits assigned,
they were classified as Category B valves in the In-Service Testing
(IST) program.
Category B valves are defined in American Society of
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Mechanical Engineers (ASME)Section XI, 1983, Article IWV-2000, as
" valves for which seat leakage in the closed position is inconsequential
for fulfillment of their function," and do not require periodic leak
rate testing.
Category A valves are defined in ASME Section XI as,
" valves for which seat leakage is limited to a specific maximum amount
in the closed position for fulfillment of their function."
When this concern was identified by the inspectors, the licensee
subsequently determined that the valves should have been re-classified
as Category A valves following identification of allowable leak rate
criteria on July 2, 1993. The licensee also determined that this
concern was previously identified on January 20, 1994, however the
recommendations were never incorporated into the planned corrective
actions for the event. At the end of the inspection, the licensee was
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in the process of re-classifying the subject valves as Category A valves
in the IST program and establishing periodic leak rate testing in
accordance with ASME Section XI requirements.
The inspectors concluded that the licensee had not implemented an
important corrective action to prevent recurrence of this event since
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periodic leak rate testing is one method to trend and/or identify
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excessive valve leakage.
b)
Licensee Event Report Review
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The inspectors reviewed LER 93-021, " Loss of Safety function for
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Emergency Closed Cooling System A," and identified the following concerns:
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The cover letter and title of LER 93-021 indicated only a loss of
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safety function for ECC "A" although ECC "B" was also inoperable
for about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> during the event.
The main body of the LER only explicitly discussed a loss of ECC
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"A".
Although the 00S condition of the division 2 DG on June 14
and 15 was discussed, neither the impact on the operability of ECC
"B" nor the consequences of a total loss of ECC were mentioned.
The inspector also noted that the 10 CFR 50.73(a)(2)(v) block was
checked which indicated that the event was reported as a loss of safety
function (both ECC trains inoperable).
10 CFR 50.73 requires that the Licensee Event Report contain an
assessment of the safety consequences and implications of the event.
The inspectors concluded that LER 93-021 failed to provide an adequate
assessment of the safety consequences and implications of the event
since the impact of the 00S condition of the division 2 DG on ECC "B"
was not explicitly discussed. This was a violation (50-440/93008-01).
8.
Anoarent Violations
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The following regulatory concerns were identified:
From June 2,1993, to July 2,1993, the plant was in operational
conditions 1, 2, or 3.
In addition, from 3:13 a.m. on June 14, 1993, to
11:05 p.m. on June 15, 1993, a period of about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, the division 2
DG was inoperable. As a result, the following apparent violations were
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identified:
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Technical Specification 3.0.3 requires that when a limiting
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condition for operation (LC0) is not met, except as provided in
the associated action requirements, within I hour action shall be
initiated to place the unit in an operational condition in which
the specification does not apply by placing it, as applicable, in:
1.
At least startup within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;
2.
At least hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
3.
At least cold shutdown within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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From June 14 to June 15, for about 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, with the ECC system
inoperable which required entry into TS 3.0.3, the licensee failed
to initiate action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit in at least
startup within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, at least hot shutdown within the
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and at least cold shutdown within the
subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This is an apparent violation.
Technical Specification 3.7.1.2 requires, in operational
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conditions 1, 2, 3, 4, and 5, that with an emergency closed
cooling loop inoperable, declare the associated systems or
components inoperable and take the action required by the
applicable specification.
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Technical Specification 3.3.7.5 requires in operational
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condition 1, 2, and.3, with only one primary containment
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hydrogen concentration analyzer operable, restore the
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inoperable channel to operable status within 7 days or be in
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at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in-cold
shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Technical Specification 3.7.3 requires in operational
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conditions 1, 2, and 3, with the reactor core isolation
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cooling (RCIC) system inoperable, restore the RCIC system to
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operable status within 14 days or be in at least hot
shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce steam dome
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pressure to less than or equal to 150 pounds per square inch
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gauge (psig) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Technical Specification 3.7.2 requires that in operational
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conditions 1, 2, or 3, with one control room emergency
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recirculation system inoperable, restore the inoperable
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subsystem to operable status within 7 days of be in at least
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hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'and in cold shutdown
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Technical Specification 3.5.1 requires, in operational
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condition 1, 2, and 3, with ECCS division 1 inoperable and
ECCS divisions 2 and 3 operable, with the LPCS system
inoperable and LPCI subsystem "A" inoperable, restore at
least the inoperable LPCI subsystem."A" or the inoperable
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LPCS system to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, otherwise be
in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in
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cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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From June 2 to July 2,1993, ECC "A" was inoperable, which
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rendered primary containment hydrogen concentration analyzer "A",
control room emergency recirculation subsystem "A",
LPCI "A",
RCIC, and LPCS inoperable.
However, the licensee failed to take
the actions required by TS 3.7.1.2 and declare the associated
systems inoperable and take the action required by the associated
specification. This is an apparent violation.
10 CFR 50, Appendix B, Criterion XVI, " Corrective Actions,"
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requires that measures shall be established to assure that
conditions adverse to quality, such as deficiencies and
nonconformances, are promptly identified and corrected.
In the
case of significant conditions adverse to quality, the measures
shall assure that the cause of the condition is determined and
corrective action taken to preclude repetition.
The inspectors concluded that the licensee's identification of
long-term corrective actions to address the leakage of OP42-F295A
was slow and failed to include periodic leak rate testing, an
important corrective action to prevent recurrence of this event
and a corrective action required by the IST program. As a result,
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an apparent violation of 10 CFR 50, Appendix B, Criterion XVI was
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identified.
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The inspectors determined that from March 19, 1993, to June 2, 1993, the
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plant was in operational conditions 4 or 5 and that fuel movements were
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not in progress. Following a review of applicable technical
specifications, the inspectors concluded that no regulatory concerns
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existed during this period.
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c.
Conclusions
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From June 2 to July 2,1993, two apparent violations of technical
specifications occurred including one in which for a 45 hour5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> period with
the ECC system inoperable, the actions required by TS 3.0.3 were not
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completed.
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In addition, identification of long-term corrective actions to address
excessive leakage of OP42-F295A was slow and failed to include periodic
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leak rate testing as required by the inservice testing program and
previously identified as a necessary corrective action in January 1994.
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As a result, an apparent violation of 10 CFR 50, Appendix B, Criterion
XVI was identified.
Finally, the inspectors concluded that LER 93-021 failed to provide an
adequate assessment of the safety. consequences and implications of the
event. As a result, a violation of 10 CFR 50.73 was identified.
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El.2 LER 94-005 " Loss of Both Trains of Control Room Emeraency Recirculation
Due to Low Emeroency Closed Coolina Temperature"
a.
Insoection Scope
The inspector reviewed LER 94-005, " Loss of Both Trains of Control Room
Emergency Recirculation Due to Low Emergency Closed Cooling
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Temperature."
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b.
Findinas and Observations
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1.
System Descriotion
As discussed in section El.1, the ECC system is designed to provide
safety-related components, including control complex chillers, with a
reliable source of cooling water.
Each division of ECC is cooled by a
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division of the Emergency Service Water (ESW) system. The ESW system
water supply is Lake Erie.
The design of the control complex chillers requires that inlet
condensing water temperature from ECC be greater than 55'F. At
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temperatures below 55'F the control complex chillers may trip on low
refrigerant temperature causing the associated train of control room
heating, ventilation, and air conditioning (CRHVAC) emergency
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recirculation mode to be rendered inoperable. The CRHVAC system
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provides cooling, heating, ventilation, and when required, smoke
removal, for the control room.
In addition, the emergency recirculation
mode of CRHVAC provides the necessary particulate and gaseous filtration
of the air supplied to the control room areas during emergency and other
abnormal conditions to reduce the radiation dose for personnel
protection.
When lake temperature dropped below 55'F, ESW system flow to the ECC
heat exchanger was transferred from a 14-inch supply line to a 3-inch
bypass line (winter mode) to ensure that ECC water temperature would not
decrease to below 55'F during accident conditions. This bypass was a
modification installed as a result of licensee identification of
concerns in 1986 that ECC temperature may decrease below 55'F under
accident heat load conditions with ESW temperatures significantly less
than 55'F.
However, the calculations and procedural modifications to
support this design change only considered accident heat loads with ECC
initially secured and not conditions where ECC would be initially
supporting minimum heat loads prior to an accident.
The control complex chillers provide chilled water to the cooling coils
of their respective CRHVAC train.
Normal cooling water to the control
complex chillers is supplied by the nuclear closed cooling (NCC) system.
During accident conditions, cooling water supply to the control complex
chillers is transferred from the NCC system to the ECC system and the
CRHVAC transfers from normal operation to emergency recirculation mode.
Control complex chillers comprise about 90 percent of the total system
heat load during accident conditions.
2.
Description of the Event
During preparation for Residual Heat Removal (RHR) heat exchanger
testing on February 5, 1994, with ESW "A"
and ECC "A" running and
supplying a minimal heat load (only RHR "A" running), ECC "A"
temperature was observed to be below 55'F.
A review was performed to determine if this condition could have existed
at any time where both trains of ECC were affected.
It was determined
that from January 28, 1994 to January 29, 1994, for about 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, both
trains of ECC were affected.
Therefore, during this period both trains
of the CRHVAC emergency recirculation mode were inoperable.
During this
period, the plant was in operational condition 3, Hot Shutdown, however
Technical Specification (TS) 3.0.3 was not entered as required.
3.
Root Caug
The licensee determined that the root cause of the event was inadequate
design. When a similar problem was identified in 1986, the corrective
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actions considered only accident heat loads with ECC initially secured
and failed to consider operation of ECC with minimum heat loads at low
ESW temperatures.
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4.
Safetw Sianificance
The time for which both trains of the CRHVAC emergency recirculation
mode were inoperable did not exceed TS 3.0.3 action requirements to
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place the plant in operational condition 4 from operational condition 3.
However, since the completion of the desiga change to install the 3-inch
ESW bypass line in 1986, the potential existed that a situation similar
to the January 28 and 29, 1994, condition may have occurred.
5.
Corrective Action _1
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As part of the licensee's immediate corrective actions, control complex
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chiller "A" was re-aligned to the NCC system and all testing was
suspended.
In addition, the associated CRHVAC emergency recirculation
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mode was declared inoperable until appropriate operating procedures were
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revised to provide for throttling ESW flow to ensure that ECC
temperature was maintained above 55'F.
As part of the licensee's long-term corrective actions, the following
was accomplished:
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Procedure changes were implemented to avoid operating ECC with
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ESW.
In those operations or system initiations where ECC and ESW
were required to be run together, an operator was stationed at the
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ESW bypass valve to manually control ECC temperature.
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Engineering investigated a reduction in the vendor prescribed ECC
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minimum temperature for the control complex chillers.
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As a result of that investigation, the chiller vendor issued a
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report on October 28, 1994, in which they could not confirm that
tha control complex chiller would start at temperatures lower than
55'F.
Design change package (DCP) 94-0027 was prepared to install a
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bypass line around each of the ECC heat exchangers utilizing a
three-way electro-hydraulic modulating temperature control valve
(TCV) to control the flow between the heat exchanger and the
bypass line based on the ECC water temperature downstream of the
heat exchanger.
Both modifications were installed in Spring 1996.
Procedures were subsequently revised to remove references to
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winter mode operations since the modification was designed to
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operate with full ESW flow under all ESW temperature conditions.
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6.
Inspector Review
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The inspectors reviewed the events and corrective actions discussed
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above. The following concern was identified:
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Temperature Control Valve Modification
The inspectors conducted a walkdown of design change package (DCP) 94-
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0027 which installed a bypass line around each of the ECC heat
exchangers utilizing a three-way electro-hydraulic modulating
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temperature control valve (TCV) to control the flow between the heat
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exchanger and the bypass line based on the ECC water temperature
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downstream of the heat exchanger.
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During subsequent discussions with licensee personnel concerning the
modification, the following was identified:
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On March 7,1996, ESW "A"
and ECC "A" were in operation with no heat
load (ECC "A" pump running) in preparation for RHR "A" shutdown cooling.
In about 45 minutes, ECC "A" temperature decreased from about 64*F to
56*F at which point ESW "A" was secured to prevent ECC "A" from
decreasing below the 55'F control complex chiller operability limit.
In response to this observation, the licensee determined that although
the system was designed to maintain ECC at 70*F with minimum heat loads
and a worst case ESW temperature of 33*F, the system was not designed to
maintain temperature above 55'F under no heat load conditions.
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In addition, operating data was collected which indicated that at
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minimum heat load conditions at an ESW temperature of 37*F, ECC
temperature decreased and stabilized at about 57.5'F, much lower than
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the design setpoint of 70*F, but greater than the 55'F chiller low
temperature limit.
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The inspectors reviewed the licensee post-modification testing program
and determined that although the modification was designed to maintain
ECC temperature above 55'F with low lake temperature and under low heat
load conditions, the licensee hid not performed a functional test to
verify that the installed modification would perform this function.
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As part of the licensee's immediate corrective actions, appropriate
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procedures were revised to ensure ECC temperature was maintained above
55*F in all environmental conditions. At the end of the inspection, the
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licensee was in the process of determining the root cause and long-term
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corrective actions.
7.
Apparent Violation
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10 CFR 50, Appendix B, Criterion XVI, " Corrective Actions," requires
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that measures shall be established to assure that conditions adverse to
quality, such as deficiencies and nonconformances, are promptly
identified and corrected.
In the case of significant conditions advern
to quality, the measures shall assure that the cause of the condition is
determined and corrective action taken to preclude repetition.
The inspectors concluded that the licensee's corrective actions to
address the inability to operate the control complex chillers under all
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environmental conditions were slow and failed to fully evaluate the
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design requirements of the ECC system against all potential.. plant
configurations, including.those present during periodic testing.
In
addition, the inspectors concluded that the licensee failed to perform
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an adequate post-modification test as part of DCP 94-0027 which
contributed to slow identification of the problem. As a result, an
]
apparent violation of 10 CFR 50, Appendix B, Criterion XVI was
identified.
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c.
[gnclusions
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The licensee's corrective actions to address the inability to operate
the control complex chillers under all environmental conditions were
slow. Corrective actions failed to fully evaluate the design
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requirements of the ECC system against all potential plant
configurations, including those present during periodic testing. The
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inspectors identified that the failure to perform an adequate post-
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modification test following installation of DCP 94-0027 was a
contributing factor. As a result, an apparent violation of 10 CFR 50,
Appendix B, Criterion XVI was identified.
V. Manaaement Meetinas
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The inspectors presented the preliminary inspection results to members of
licensee management on August 30, 1996.
In addition, the final inspection
results were presented on September 11, 1996. The licensee acknowledged the
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findings presented.
The inspectors asked the licensee whether any materials examined during the
inspection could be considered proprietary.
No proprietary information was
identified.
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PARTIAL LIST OF PERSON,S CONTACTED
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Licensee
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N.: Bonner
Engineering Director
D. Phillips
System Engineering Supervisor
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D. Drevay
Component Performance Unit Supervisor
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G._Oates
Design Engineer
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P. Roney
Design Engineer
J. Bridgens
~ECC System Engineer
K. Jury
Compliance Supervisor
L. Zerr
Compliance Engineer
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NRC
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D. Kosloff
Perry Senior Resident Inspector
R. Twigg
Perry Resident Inspector
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LIST OF ACRONYMS USED
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Alternating Current
Alarm Response Instruction
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CFR
Code of Federal Regulations
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CR
Condition Report
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Control Room Heating, Ventilation, and Air Conditioning
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Design Change Package
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-DG
Diesel Generator
Emergency Closed Cooling
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Emergency Service Water-
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'F
Degrees Fahrenheit
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FCR
Field Clarification Request
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GPM
Gallons Per Minute
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IST_
In-Service Testing
LC0
Limiting Condition for Operation
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'LCS
Licensing and Compliance Section
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Low Pressure Core Spray
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Motor-0perated Valve
Nuclear Closed Cooling
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Net Positive Suction Head
Perry Administrative Procedure
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Potential Issue Form
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Pounds Per Square Inch Gauge
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PTI
Periodic Test Instruction
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Reactor Core Isolation Cooling
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SVI
Surveillance Instruction
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Temperature Control Valve -
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TS
Technical Specification
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TXI
Temporary Instruction
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Updated Final Safety Analysis Report
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Work Order
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INSPECTION PROCEDURES USED
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Engineering
Effectiveness in Licensee Controls in Identifying, Resolving, and
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Preventing Problems
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ITEMS OPENED, CLOSED, AND DISCUSSED
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Opened
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96008-01
Inadequate Licensee Event Report
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