IR 05000440/1996015
| ML20129H461 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/25/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20129H446 | List: |
| References | |
| 50-440-96-15-EC, NUDOCS 9610310304 | |
| Download: ML20129H461 (35) | |
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l U.S. NUCLEAR REGULATORY COMMISSION REGION lil
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Docket No:
50-440 License No:
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Report No:
50-440/96015(DRS)
Licensee:
Cleveland Electric lliuminating Company Facility:
Perry Nuclear Power Plant Location:
Parmly at Center Road Perry, OH 44081 Date:
October 11,1996
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Inspector:
Eric Duncan, Reactor Engineer Approved by:
M. A. Ring, Chief, Lead Engineering Branch Division of Reactor Safety Meetina Summarv Four apparent violations identified during the inspection were discussed, along with the corrective actions taken or planned by the licensee. Two of the apparent violations involved technical specification limiting condition for operation action requirements which were exceeded. 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 where the emergency closed cooling system was inoperable, and the actions required by Technical Specification 3.0.3 were not completed. The remaining two apparent violations involved corrective action issues, in one case corrective actions were neither timely nor adequate to address a leaking safety-related motor-operated butterfly valve. In a second case, corrective actions failed to correct an adverse condition regarding the ability to operate safety-related control room complex chillers at low lake temperatures.
9610310304 961025 PDR ADOCK 05000440 g
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Report Details 1.
Persons Present at Conference Centerior Service Company
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L. Myers, Vice President, Perry N. Bonner, Director, Perry Nuclear Engineering Department W. Kanda, Director, Perry Nuclear Assurance Department R. Collings, Director, Perry Quality Assurance Department F. Kerney, Plant Operations Superintendent, Perry H. Oates, Design Engineer, Perry
U. S. Nuclear Reaulatorv Commission W. Axelson, Acting Deputy Regional Administrator, Rlli G. Grant, Director, Division of Reactor Safety, Rlli M. Ring, Chief, Lead Engineering Branch, Rlli R. Lanksbury, Chief, Division of Reactor Projects, Rlli D. Kor.loff, Perry Senior Resident inspector J. Hopkins, Perry Project Manager, NRR B. Burgess, Enforcement Director, Rlli P. Pelke, Enforcement Specialist, Rlli B. Berson, Legal Counsel, Rlll 2.
Pre-decisional Enforcement Conference A Pre-decisional Enforcement Conference was held in the NRC Region lli Office on October 11,1996. Apparent violations of NRC regulations were discussed. The inspection findings were documented in NRC Inspection Report No. 50-440/96008, which was transmitted to the licensee by letter dated September 27,1996. The purpose of this conference was to discuss the apparent violations, root causes, contributing factors, and the licensee's corrective cctions.
The licensee's presentation included a discussion of the apparent violations, their safety significance, the circumstances which caused the apparent violations, and an outline of corrective actions taken or planned.
The NRC representatives questioned the licensee to clarify the extent of the licensee's investigation and corrective actions.
A copy of the handouts used during the presentation is attached to this report.
Attachment:
As stated
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Perry Xuclear Power Plant
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Pre-Decisional Enforcement Conference Presentation i
to the i
U.S. Xuclear Regulatory Commission j
October 11,1996
i CENTERIOR
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ENERGY
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b Discussion Topics
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Introduction Lew Myers i
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Emergency Closed Cooling (ECC)
Russ Kearney
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System Overview
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ECC Valve Leakage Neal Bonner/
Russ Kearney
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ECC System Temperature ~ Control Harry oates
Modification i
Corrective Action Program Issues Rick Collings/
Bill Kanda Summary Lew Myers
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Introduction
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Both issues involve the Perry Emergency Closed Cooling system;
however, the circumstances are unrelated.
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Opportunities to understand the issues and take correc~tive actions did
exist.
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History of corrective action issues are well documented.
- The implications of management controls were significant.
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The potential loss of the Emergency Closed Cooling system was
significant; however, after both issues were identified, risk was minimized.
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Obtain a common understanding of the facts.
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i Provide a common understanding that timeliness of both issues was
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appropriate for the known design bases.
L Provide a common understanding of the safety significance.
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f Provide assurance that the Perry Management Team is dedicated to
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identification, communications, and prompt corrective actions for i
station problems.
The NRC will determine that escalated enforcement is not necessary for
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these issues.
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I ECC System Overview
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Presented by:
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Presented by:
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Neal Bonner Russ Kearney l
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e ECC Valve Leakage
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Sequence of Events
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Root Causes
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Corrective Actions
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Sequence of Events
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7/1/93 ECC "A" surge tank overflowed.
- Temporary Instruction (TXI) - 162 written to determine source and rate of i
leakage.
7/1/93 Motor operated butterfly valve OP42-F0295A determined to be leaking
greater than 250 gpm. This was the first leak test performed on this valve.
i Valve OP42-F0295A adjusted and retested; returned to zero leakage.
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Field Clarification Request (FCR) 17843 initiated to determine allowable ECC system leakagerate.
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i 7/6y@3 Condition Report (CR)93-132 initiated (first day after Holiday
weekend); determined to not be a Technical Specification violation.
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7/7/93 CR 93-132 initially determined by Regulatory Affairs Section to not be
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reportable; the scenario necessary for ECC train / system inoperability was not known at that time.
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Sequence of Events (cont.)
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8/25/93 CR 93-132 root cause investigation completed. Initial cause
investigation inadequate; independent external review of reportability being performed.
12/1/93 CR 93-132 inve'stigation approved by Engineering management.
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12/15/93 External review received which determined ECC could have failed to
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12/30/93 Issue was reported to the NRC as a loss of safety function of the ECC
system after internal review of contractor's report and a historical reconstruction of issue sequence.
j 7/23/94 LER 93-021, " Loss of Safety Function for Emergency Closed Cooling
System A," was issued.
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Safety Significance l
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Scenario assumptions.
- Loss of Offsite Power (LOOP)/ Loss of Coolant Accident (LOCA)- Nuclear Closed Cooling (NCC) unavailable due to loss of stub bus.
Division 2 diesel generator inoperable - ECC "B" fails to function.
Make-up capability unavailable.
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ECC system was only vulr.rable for approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> due to Emergency Service Water (ESW) work. During a portion of this time, ESW was only " administratively" inoperable.
Probability Safety Assessment.
- LOOP /LOCA-5x104
- LOOP /LOCA coincident with division 2 diesel generator inoperability - 1.6x10-7
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NOTE: ECC "A" would have isolan:1 during LOOP /LOCA.
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Root Causes
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Initial review and classification of ECC/NCC isolation valve
(0P42-F0295A) was as an ASME Code,Section XI, Category "B" valve.
Subsequent evaluation and review of FCR 17843 failed to adequately address design bases leak requirements.
The need to reclassify the valve was recognized; however, this action was not adequately resolved in the CR evaluation and corrective actions.
Improper Settings of ECC/NCC isolation valve (0P42-F0295A).
- Combination of personnel error and weak procedural direction for setting l
MOV limit switches and mechanical stops in 1993. This maintenance was i
performed to correct a previously identified concern.
i Leakage test was not performed since leakage criteria had not been
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established.
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Root Causes (cont.)
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ECC leakage corrective action issues.
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FCR evaluation was inappropriately used as design input.
Inadequate initial work history review (extent of condition).
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Flawed cause analysis aid inadequate management review.
Work and corrective action backlogs adversely affected the timeliness of some actions.
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Corrective Actions
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Actions Taken Prior to July 1996.
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Reviewed other safety-related systems for similar conditions, reviewed I
wcrk/ testing history, and ensured properly set limit switches. No other similar i
concerns were identified.
Appropriate Maintenance personnel trained; lessons learned were shared with
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site personnel.
Post-maintenance testing practices have been, and continue to be, enhanced;
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procedural direction wu revised.
l FCR processes and controls revised.
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Valve has been leak tested within frequency specified for Category "A" valves as a result of Generic Letter 89-10 and preventive maintenance activities.
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Actions to reduce backlogs were implemented.
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Corrective Actions (cont.)
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Other corrective actions taken since issue
identification in July 1996.
Valve leakage criteria has been established.
The In-Service Testing program has been revised to identify these valves as Category "A." Associated testing procedures are being revised.
i The applicable Alarm Response and Off-Normal Instructions have been j
revised to delineate required operator actions.
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An extent of condition review is being re-performed.
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Issue Summary i
Self identified.
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Actions taken at the time were reasonable.
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Corrective actions have.resulted in improved
Engineering processes.
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Risk significance was minimized.
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Modification
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ECC Temperature Control Modification
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- Sequence of Events i
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- Root Causes
- Corrective Actions
- Issue Summary
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Sequence of Events
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2/86 Control Complex Chiller trips on low refrigerant temperature due to low
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lake water temperatures.
t 4/86 Design Change Package (DCP) 86-0224 initiated to alleviate low lake l
water temperature concerns.
2/94 - 3/94 During Residual Heat Removal system heat exchanger
performance testing, with ESW system flow reduced by the 3" by-pass line, the
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ECC system temperature is observed drifting below 55 oF. LER 94-05 submitted as a condition prohibited by Technicsl Specifications and as any condition that alone could have prevented the fulfillment of the safety function.
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The 3" diameter bypass line sizing calculation did not consider minimum heat loads.
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Operator guidance was established to throttle ESW flow using the 3 inch bypass valve.
A modification (i.e., DCP 94-0027) was initiated to control ECC system temperatures
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with varying lake water temperatures.
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Sequence gf Events (cont.)
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10/94 Design approach shifted from ESW flow control to ECC flow control. It
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was determined that from a safety perspective, modification should not be installed while the plant was on-line. LER was revised.
$95 DCP 94-0027 was issued.
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Calculation P42-30 concluded that with minimum or no ECC loads, the valve will by-
pass 90% of system flow with stable control.
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10/95 DCP 94-0027 was revised to incorporate TXI-0230 as the post
modification test. Precautions were delineated to prevent temperature from dropping below 55 oF.
2/96 - 3/96 DCP 94-0027 implemented.
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with no system loads, and full ESW system flow, when temperature was i
observed to be decaying.
Flow induced (vortexing) heat transfer at the heat exchanger discharge piping, coupled
with potential convective heat losses, resulted in the decay.
The effect of this phenomena was not expected nor assumed in calculations.
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f Concerns do not exist during normal or accident configurations.
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Adequate heat load is available when system is in normal line-up and when
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required to perform its safety function.
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The testing configuration is the only line-up during which the ECC.
,
temperature control modification may not perform its function; low lake
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water temperature also required.
- The rate of temperature decay is slow relative to the normal duration (i.e., less than 30 minutes) in which the test evolution occurs.
Operator actions ensured the system remained operable throughout
the fifth operating cycle, while the design process proceeded in a
controlled and scheduled manner.
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Page 24
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Root Causes
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Heat transfer mechanisms associated with the new
configuration were not recognized as being
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significant during design development / review.
TXI-0230 failed to confirm the adequacy of the design.
Design assumptions inherent to the design of the modification
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were not confirmed by post modification testing.
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Corrective Actions
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Procedural direction has been developed to avoid
challenging the ECC system.
Additional actions are being evaluated:
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Post-modification testing development methodology is being i
evaluated.
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Control Complex chiller setpoint evaluation is being performed.
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Page 26 i
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Issue summary
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- Selfidentified.
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- Risk significance.
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Condition only existed during testing configurations.
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- Actions taken were reasonable. Modification did evaluate the ECC system against all potential plant configurations, including those present during j
periodic testing.
- The modification has improved system performance.
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Page 27
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Corrective Action Program Issues
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Presented by:
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i Rick Collings j
Bill Kanda i
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Page 28
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t Corrective Action Program Issues
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- Identified weaknesses in the program are well documented (1992-1993).
l Inconsistencies in program utilization and problem identification.
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Root cause determinations were sometimes incomplete or incorrect.
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Not fully effective and timely in correcting problems.
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Page 29
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Corrective Action Program Issues (Cont.)
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- Improvement actions initiated starting in Fall 1993.
Perry Course of Action / Perry Plan for Excellence.
Post modification / maintenance test manual.
Change in documentation threshold and management expectations.
Complete revision / approach to Corrective Action program.
Root cause and extent of condition emphasized, including additional formal root cause analysis training.
Establishment of a Corrective Action Review Board.
Site and industry experience reviews integrated into process.
Trend analysis and categorization process.
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Performance indicators established.
Page 30
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Summary
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Ensure a common understanding of the facts, including the timeliness of actions
associated with the Emergency Closed Cooling / Nuclear Closed Cooling valve leakage, which was based on what was known at the time.
Use of the Technical Specifications during the time of the issue did not result in entry into the action statement, due to the failure of the design to recognize that the isolation
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valves should be considered Section XI Category "A" valves. Prompt actions were taken to properly set the limit switches and leak rate tests have been, and continue to be, performed.
During the time period in question, the Emergency Closed Cooling system would not
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have failed without a design basis scenario, which includes a LOOP /LOCA and a loss of the division 2 diesel generator.
Emergency Closed Cooling temperature control modification implementation
was timely. This modification did improve the operations of the control room
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chiller system.
Opportunities existed to identify both issues. Consideration should be given for
self identification, time frame, and corrective actions taken for each issue.
The Perry Management Team is dedicated to prompt identification,
communication, and corrective actions for station problems.
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