IR 05000315/1994007
| ML17331B407 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/26/1994 |
| From: | Burgess B, Gill C, Salehi K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17331B405 | List: |
| References | |
| 50-315-94-07, 50-315-94-7, 50-316-94-07, 50-316-94-7, NUDOCS 9406070080 | |
| Download: ML17331B407 (16) | |
Text
U.S.
NUCLEAR REGULATORY COHHISSION
REGION III
Report Nos.
50-315/94007(DRS);
50-316/94007(DRS)
Docket Nos.
50-315; 50-316 Licensee:
Indiana Nichigan Power Company 1 Riverside Plaza Columbus, OH 43216 License Nos.
D. C.
Cook Nuclear Plant, Units 1 and
Inspectio'n At:
Stevensville, Michigan Inspection Conducted:
April 11 through 22, 1994 Inspectors:
.
Gs ate ae i at Approved By:
urge s, Operational Programs Section ate Ins ection Summar Ins ection on A ril ll throu h 22 1994 Re ort Nos.
50-315 94007 DRS 50-316 94007 DRS
d: ill*i p i
i, d,
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inspection of engineering activities, using selected portions of Inspection Procedure 37700, to determine if design changes, engineering support, and corrective actions were effectively controlled and implemented.
Results:
Mithin the areas inspected, a violation of NRC requirements and several weaknesses were identified.
The apparent violation was failure to perform a required safety evaluation of a proposed change in the facility, as described'in the Updated Safety Analysis Report (USAR), to ascertain whether the proposed change involved an unreviewed safety question.
Overall, your engineering activities were adequate.
However, the inspectors identified weaknesses regarding communications, engineering assessments, and inattention to detail (see Section 3.3).
9406070080 940527 PDR ADQCK 05000515 PDR
DETAILS 1.0 Persons Contacted American Electric Power Service Com an AEPSC E.
E. Fitzpatrick, Senior Vice President-Nuclear Generation H. S. Ackerman, Engineer P. A. Barrett, Hanager-guality Assurance Department.
S.
D. Benes, Hechanical Systems Support Section Hanager A.
Feliciano, Engineer E. V. Gilabert, Engineer G.
D. Hines, Engineer S.
P.
Hodge, Division Hanager-Design Engineering W. T. HacRae, Senior Engineer D. H. Halin, Nuclear Safety,and Fuel Section Hanager R.
S.
Papps, Senior Engineer S.
H. Steinhart, Division Hanager-Nuclear Engineering Support K. J. Toth, Senior Engineer L. H. VanGinhoven, Project Superintendent J.
S. Wiebe, equality Assurance and Control Superintendent Indiana Hichi an Power Com an IH L. S. Gibson, Assistant Plant Hanager-Projects K. R. Baker, Assistant Plant Hanager-Production H.
E. Barfelz, Nuclear Safety and Assessment Supervisor T. H. Bestrom, Operations-Production Supervisor B.
R. Burgess, Scheduling Superintendent D. 0. Horey, Chemistry Superintendent R. S.
Ptacek, Licensing Coordinator J.
L. St.
Amand, Plant Engineering-Hechanical Engineering Supervisor G. A. Weber, Plant Engineering Superintendent J.
D. White, Hanager-Site Nuclear Services U.S. Nuclear Re ulator Commission NRC W. J.
Kropp, Chief, Projects Section 2A B. L. Burgess, Chief, Operational Programs Section J.
A. Isom, Senior Resident Inspector D. J. Hartland, Resident Inspector The above individuals attended the exit meeting on April 22, 1994.
The inspectors also contacted other licensee employees during the inspection and by telephone through Hay 13, 1994.
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2.0 Licensee Action on Previous Ins ection Findin s
Several problems and concerns identified in previous NRC inspection reports were reviewed for appropriate corrective actions.
The items reviewed and the inspectors'valuations of the actions are discussed in this sectio.1 Closed Unresolved Item 315 94002-06 DRP This item was related to a Unit 1 motor-driven auxiliary feedwater (HDAFW)
pump frozen mini-flow line which provided dead-head protection for and was shared by both HDAFW pumps.
While in Hode 3 on February 12, 1994, soon after both pumps were started in preparation to secure the main feedwater pumps, elevated pump casing and discharge piping temperatures were noted.
The elevated temperatures were caused by a blockage of the miniflow line that resulted in both pumps being dead-headed.
The cause of the blockage was a
frozen section in the mini-flow piping to the Condensate Storage Tank (CST)
due to a blown fuse in a heat-trace circuit.
Although the problem was promptly corrected and the pumps were undamaged, the licensee root cause and corrective action for the frozen pipe, including an assessment of the actions taken to correct certain similar previous incidents, was an Unresolved Item pending further review by the NRC.
To resolve this matter, the inspectors reviewed completed licensee Condition Report No. 94-0227 and related documentation, and interviewed cognizant personnel regarding the followup investigation, root cause analysis, and corrective actions.
The investigation concluded that adequate backup systems were available to compensate for potential HDAFW pump failure during extended operation at low flow and that under analyzed design basis accident conditions, the automatic system realignment would have allowed the pumps to accomplish their safety function of maintaining steam generator level.
The root causes of the event were determined to be failure of the heat-trace power cable insulation (when the cable grounded to the conduit, the distribution panel overloaded and the fuse blew), failure of the
amp breaker to trip and isolate the failed heat-trace (causing the 30 amp panel fuse to blow instead),
and the alarmastat was out of calibration (causing the degraded heat-trace con'dition to go undetected in the control room).
During the investigation, it was also revealed that the Chemistry Department had a problem with obtaining a
CST sample on February 7,
1994.
It was later recognized that the apparent failures of two sample point heat-trace loads were both the cause and a
symptom of the failed panel fuse, also due to the failure of the
amp breaker to isolate the fault.
,In addition, a frozen instrument line for the primary water storage tank level alarm transmitter was suspected on February 3, 1994.
The licensee concluded that if the February 3 and 7, 1994 events had been more thoroughly or quickly investigated, the February 12, 1994 event may have been avoided.
The inspectors concluded that the licensee's investigation, root cause analysis and subsequent corrective action to prevent recurrence were appropriate.
This item is closed.
2.2 Closed Violation 316 94002-10 DRP This item was related to the failure to accomplish maintenance on a dump valve, 2-HRV-241, for main steam stop valve (HSSV) 2-HRV-240 in accordance with procedural requirements and failure to designate the maintenance procedure as in-hand in accordance with Plant Hanager Instruction procedural requirements.
The inspectors reviewed the licensee response dated Hay 6, 1994, associated licensee Condition Report No. 94-0328, and interviewed licensee representatives responsible for the resultant corrective action to verify that appropriate actions had been taken to correct the. problem.
The inspectors concluded that all necessary corrective action had been completed to prevent recurrence and that the licensee expected to complete the re-
evaluation of the other 30 maintenance procedures which are presently not designated as in-hand procedures by July 1, 1994.
This item is closed.
2.3 0 en Unresolved Item 315 94002-12 DRP This item was related to the inspector concerns regarding the Operations Department Superintendent's permission to raise the CCW supply temperature from its USAR limit of 95'F to 105'F.
This matter was further reviewed during this inspection as documented in Section 8.0 of this report.
This item remains open pending NRC review of the licensee's response to the enclosed Notice of Violation and other requested information, and a determination whether further NRC enforcement action is warranted.
2.4 0 en Ins ection Followu Item 315 94002-13 DRP This item was related to inspector concerns regarding a January 27, 1994 event that resulted in damage to a
CCW RCP bearing caused by low oil level. This matter was further reviewed during this inspection as documented. in
, Section 8..0 of this report.
This item remains open pending NRC review of the licensee's response to the enclosed Notice of Violation and other requested
'nformation, and a determination whether further NRC enforcement action is warranted.
3.0 3.1 En ineerin and Technical Su ort Introduction The purpose of this inspection was to evaluate the effectiveness of engineering and technical support (EATS) organizations in the performance of routine and reactive site activities, including identification and resolution of technical issues and problems.
This inspection focused on system engineering functions, modifications, temporary design change activities, technical problem resolution, and engineering support.
The criteria used to assess the ESTS performance was quality of technical work produced, understanding of plant design, and active involvement in preventing and solving plant problems.
3.2 ESTS Or anizational Structure and Res onsibilities Engineering activities for the D.C. Cook plant were performed by several organizations including Nuclear Engineering, Site Engineering, Plant Engineeri'ng, and Maintenance Engineering.
Nuclear Engineering and Site Engineering are under the corporate organizational structure.
The portion of the Nuclear Engineering Department located at corporate headquarters in Columbus, Ohio provided design expertise for technical issues and design changes, and had lead engineering responsibilities for major modifications (Requests for Change)
and some of the minor modifications.
Site Engineering, located at the plant, provided limited design expertise (site design) for minor modifications and plant modifications (non-safety related),
and onsite technical and engineering support for installation of modifications (project engineering).
The Plant Engineering Department provided system engineers, test engineers, and other onsite engineering support.
Maintenance Engineering, as part of the Plant Engineering Department, provided technical support for maintenance activities.
The licensee infrequently used consultants or contractors for engineering wor.3 E&TS Effectiveness Although the EATS organizations were amply staffed and the engineers were often well trained and experienced, the inspectors identified weaknesses regarding communications, engineering assessments, and inattention to detail.
The inspectors'onclusions were based on reviews documented elsewhere in this report and interviews with a wide cross-section of engineers.
Specific examples were as follows:
Section 8.0 on CCW system concerns documented weaknesses in all three of the above categories.
Although some of the apparently improper operational decisions were made by the Operations Department, the basis for those decisions were based on engineering input, judgement, and communications.
Acceptance'y the plant staff of unsubstantiated engineering judgements, rather than a rigorous review or formal evaluation, was a contributing=cause for exceeding CCW system operational limits.
Section 6.0 on RFC-4138, replacements of diesel starting air compressors, documented that several steps of this modification were determined to be difficult, inefficient, and cumbersome.
The contributing factors for these difficulties included weaknesses in communications=,
engineering assessments, and inattention to detail.
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Although the violation stated in Section 2.2 was closed in this inspection report, the problem with the subject HSSV dump valve occurred during a voluntary four-hour LCO during power operation.
Because of problems with the repair of the valve and the rush to meet the LCO time limit, the reinstalled valve failed which resulted in a reactor scram.
In January 1993, the same four-hour LCO was entered during power operation three times for the same reason.
Only after the more significant fourth event and the resultant violation, did the licensee take effective corrective action.
The root cause and contributing factors encompass the above three weakness categories.
4.0 Observation of Plant Conditions The inspectors performed plant tours and system walkdowns to observe the material condition, indication of equipment problems, housekeeping and other unusual conditions.
The inspectors noted various water leaks in the lower level of, the turbine building.
The fact that both units were in an outage, and 'numerous repair and maintenance activities were in progress, were considered contributors to the identified leaks.
Some isolated foreign material control problems existed in that debris and. tools were found in several locations in the auxiliary and turbine buildings.
Also, there were several examples of work orders tags found, relating to water and oil leaks, which dated from 1991.
The licensee responded to these concerns in a timely manner and took appropriate corrective action, as needed.
5.0 Tem orar Hodifications The inspectors did not identify any concerns with the temporary modification process.
Their review concluded that the following four temporary modifications were adequately controlled with the appropriate levels of engineering involvemen ~
Tem orar Modification 2-92-013:
This modification deleted stellite from the bonnet backseat area on gate valve 1-ICH-321 (located in the west RHR heat exchanger room) per the valve manufacturer's engineering recommendation letter.
The planned restoration was to replace the bonnet with a new bonnet.
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Tem orar Modification 1-93-007:
This modification was to drill, tap, and plug a through wall leak on the bonnet of steam dump test loop shutoff valve 1-TBP-123.
The planned restoration was to remove the plug and make a weld repair to the bonnet.
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Tem orar Modification 1-93-010:
This modification was to install a
plug in the furmanite injection hole in the stuffing box on 1-TBP-133 to stop steam leak/air inleakage during steam dump valve testing.
The planned restoration was a weld repair.
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Tem orar Modification 1-93-013:
This modification was to change the wiring configuration inside the ILRT test rack, LTR-l, for.TSC computer RTDs in accordance with existing plant drawings.
The planned restoration was to enter containment to restore RTDs wiring per the drawings rather than at the test rack.
6.0 Ha 'or Modifications The inspectors reviewed selected portions of two open major modification (Request for Change)
packages and suppoi ting documentation, interviewed personnel involved with the design change evaluations and modification installations, verified that the physical installations reflected the documented design changes, and observed ongoing installation work activities.
The'escriptions of the packages reviewed and the-inspectors'onclusions follow:
RFC No. DC-12-3076:
Functionally Replace Westinghouse Radiation Monitors with Eberline RHS Compatible Equipment This RFC did not include removal of the Westinghouse monitors.
This removal was to be accomplished by a separate design change which would not commence until the functional replacements installed by RFC-3076 were operable and had, prior to being declared operable, shown reasonable functional reliability.
This design change was to comply with commitments to NRC Generic Letter 89-06, regarding replacement of obsolete equipment.
The RFC was divided into seven tasks to be completed over several outages, involving installation of 16 new radiation monitoring systems.
The inspectors reviewed the three RFC tasks with work in progress during the ongoing dual unit outage.
The inspectors did not identify any problems with the design change evaluations, work package instructions, the knowledge level of the engineering staff.involved, observation of work in progress, or the comparison of the installation drawing details with the installed components.
RFC No. DC-12-4138:
Replace Diesel Starting Air Compressors, Units
and
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7.0 A total of eight air compressors were to be replaced with the new compressors to include aftercoolers and coalescing filters to improve air quality.
This modification was initiated in response to the preventive maintenance program to replace aging parts.
However, the existing air compressors parts were no longer being manufactured.
In response, it was decided that all air compressors would be replaced.
Improved performance, added reliability, and response to environmental consideration were other driving forces to chan'ge the design of this system by replacing all air compressors.
Recent failure of an air compressor and apparent impending failures of additional air compressors prompted the licensee to expedite the modification process.
The interface between engineering staff at Columbus and site project engineering became a focal-point in this modification.
During the installation process, several steps of this modification were determined by the licensee to be difficult, inefficient, and cumbersome.
The inspectors concluded that during the design phase, this modification should have been better coordinated to improve installation efficiency.
Self-Assessment and Trendin of En ineerin Activities Assessment of engineering activities at the D.C.
Cook plant consisted of audits and surveillance of modifications and engineering support, including corrective actions.
Overall, the various assessments covered the spectrum of engineering support activities.
Corrective actions for engineering activities were also trended.
In March 1994, the Safety and Assessment (SM) Department reorganized.
The new department retained quality control (QC) functions and added the onsite quality assurance (QA) group.
The new QA/QC Department reported directly to corporate headquarters.
However, the Nuclear S&A group was removed from this department and reported directly to the Assistant Plant Hanager - Projects.
This provided the Plant Engineering Department with self-assessment capabilities.
Because the new organization was not fully implemented, it was too early to assess its effectiveness.
7.1 Audits and Surveillance The inspectors reviewed recent Qu'ality Assurance (QA) audit and surveillance records and interviewed personnel to determine the effectiveness of the licensee's self-assessment of engineering activities.
The QA audit and surveillance records indicated that these activities were generally performance based, effective in finding engineering weaknesses, and adequately, covered engineering activities.
Surveillance activities complimented the audit program in an appropriate manner.
Findings and recommendations noted were significant and appropriate.
The importance placed on corrective action effectiveness was commendable.
7.1.1 ualit Assurance Audits Comprehensive audits of engineering were normally conducted yearly with additional. audits of supplemental engineering activities and design change controls conducted
'as needed.
Records of three QA audits of engineering,
including major modifications or other engineering related activities, were reviewed.
These records were:
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gAVP 93-03:
Design control (RFC 12-4122: Modification of Spare Containment Penetration CPN No. 71)
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gAVP 93-05: Nuclear Engineering Department and Design Division Software Audit
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gAVP 93-06: Design Control (RFC 12-2985:
Change Foxboro-H-Line Electronic Instrumentation Modules)
The 1994 Onsite and Offsite guality Verification Audit Schedule and Plans were also reviewed.
The audit schedule for engineering activities appeared comprehensive and appropriate.
7.1.2.
Surveillance Surveillances were used by gA to supplement audits in the assessment of engineering performance.
The inspectors reviewed a selected sample of the engineering related surveillance reports issued in the last year.
These reports indicated that the surveillance program, complemented the audit program in an appropriate manner.
The program was performance-based and broad in scope but was focused on suspected or potential problem areas.
The six surveillance reports reviewed were:
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SURV 93-05: Condition Report Investigations and Corrective Actions in the Areas of Civil Engineering, Chemical Engineering, and Structural and Analytical Design SURV 93-06: Special Topic Surveillance in Response to an NRC Inspection Concern to Ensure That LER Corrective and Preventive Action Commitments are Reflected in Condition Reports, Tracked, and Trended
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SURV 93-08: Technical Evaluations in the Areas of Piping, Valves, Power Systems, and Human Factors
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SURV 93-10:
NRC Inspection Report Followup:
EDG Operability Concerns
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SURV 93-11: Condition Report Investigations and Corrective Actions in the Areas of Nuclear Design-Elect'rical, Power Systems, and Human Factors
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SURV 93-14: Condition Report Investigations and Corrective Actions in the Areas of Nuclear Design-Mechanical; Heat Exchangers, Pumps
& Turbines; and Piping, Valves, HVAC, and Fire Protection 7.2 Corrective Action Assessment and Trendin The inspectors reviewed the results of engineering corrective action assessment and trend report for the first six months of 1993. (released on December 6, 1993).
The trend report identified that human performance related issues constituted the majority of the issued corrective action reports.
In
comparison, equipment related issues constituted a considerably lower percentage.
The inspectors did not identify specific corrective actions in response to specific trend report recommendations.
However, during the exit interview, the licensee stated that internal memoranda had been issued to various supervisors for evaluation of trend report results.
To ensure consistency of the trend report data, the licensee opted to audit the results of the second six month trend report before implementing specific corrective actions.
The inspectors determined that the delay in responding to corrective action recommendations made in the first trend report constituted a weakness in that an opportunity to address human performance issues was missed.
The continued poor trends related to'human performance in the preliminary results of the second trending report supported this conclusion.
8.0 Com onent Coolin Water CCW S stem Concerns 8.1 Introduction The inspectors fol.lowed up on several concerns regarding Unit
.CCW system operational and engineering decisions which were i'dentified'y the resident inspectors in Inspection Reports No. 50-315/94002,.(DRP)
and No. 50-316/94002 (DRP).
After further review, the inspectors concluded that these concerns in aggregate may have represented a significant weakness in communications, engineering assessments, and inattention to detail.
In response to these concerns and the composite weaknesses, the licensee initiated'a Condition Report to investigate the concerns and take corrective actions, as needed.
The root cause investigation was being conducted by a multi-disciplinary team which planned to look generically at the control room command and control process and the use of engineering judgements in making unit operational decisions, rather than only confining the review to specific CCW concerns.
The investigation was scheduled for completion by June 23, 1994.
8.2 Back round Information In September 1993, the licensee decided to continue to operate the Unit
reactor coolant pump (RCP)
No.
4 with a low oil level in the lower radial bearing until the scheduled Unit 1 shut down on February 12, 1994.
The following list represented a chronological summary of the NRC inspector concerns.
In late September 1993, RCP-14 lower oil pot annunciator alarmed.
Although the alarm was valid, the licensee decided to continue operation with a lower radial bearing low oil level.
The Operations Department decision was based on the engine'ering judgement that it was acceptable to continue operating the pump motor in this degraded condition.
On January 17, 1994, the RCP-14 bearing temperature spiked above the alarm setpoint from 155~F to 302 F.
The temperature soon'tabilized at 2184F.
The apparent cause of this transient was a recent decrease in the CCW supply temperature of 10 to 154F which was assumed to have shrunk the oil level below that needed for adequate lubrication, resulting in bearing damage.
Although the temperature indication and alarm were valid, the licensee decided to continue Unit 1 operation with RCP-14 damaged.
The Operations Department decision.was based on the engineering judgement that it was acceptable to continue to operate the pump in this further degraded conditio ~
Par't of the reason for the engineering judgement to continue to run RCP-14 was the system engineer's belief that the high temperature reading and alarm was due to a faulty RTD.
Therefore, an installed spare RTD was used instead and indicated a temperature of 155 F.
Even after the resident inspectors, with the assistance of an I&C technician, demonstrated that the original RTD gave an accurate temperature reading and its physical location was more indicative of the bearing temperature,'he system engineer retained his original judgement that continued operation of RCP-14 was appropriate.
Also, the resident inspectors concluded that the engineering decision to switch to the spare RTD could not be supported by any available engineering data.
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In addition, on January 27, 1994, the Operations Department Superintendent gave permission to raise the CCW supply temperature from the procedural limit of 95 F to 105'F.
The reason for the increase was to reduce the amount of PCP-12 seal leakage in an effort to keep reactor coolant leakage below technical specification limits until the planned Unit 1 shutdown on February 12, 1994.
This permission was based on the engineering judgement of the CCW system engineer that the increase in temperature would not damage the system or the equipment it cooled.
8.3 Subsequently, because the CCW temperature alarm was not reset from 95~F to 105'F, the operator was not alerted that CCW temperature had increased above 105'F during a transient caused by boric acid evaporator operations on January 27 and 28, 1994.
The operator was unable to control the transient before the temperature rose to 110 F.
During the inspection, the inspectors asked the licensee to supply engineering calculational assessments to determine the safety significance of operating with a CCW supply temperatures of 105 and 110 F, and with an RCP return CCW temperature above the 120'F alarm value.
The inspectors were shown E-mail engineering judgement notes from the corporate office which stated that there was no safety significance to operating this system beyond its design data temperature alarm limits'.
The inspectors were'also told that the requested engineering calculations could not be accomplished because the licensee and the vendor did not know the design limits for the CCW system temperatures.
The only documented number for the design data upper limit for the CCW supply temperature is 95'F as stated in the USAR, operational procedure, and system description.
Likewise, the only documented value for the upper limit on the return line from the RCPs is 1204F, which was exceeded on RCP-14.
NRC Conclusions Pending the results of the licensee's ongoing root cause team investigation, Unresolved Item No. 315/94002-12(DRP)
and Inspection Followup'tem No.
315/94002-13(DRP),
regarding the above concerns will remain open.
However, a
violation will be issued associated with exceeding the CCW supply temperature of 95'F without the required safety evaluation.
After reviewing the licensee's response to the violation and the licensee's conclusions of the ongoing root cause team investigation, the NRC will determine whether further NRC enforcement action is warranted.
8.3. 1
A arent Violation of 10CFR 50.59 10 CFR 50.59 requires, in part, that changes made to the facility as described in the safety analysi.s report be evaluated in accordance with 50.59(a) to determine, in part, if an unreviewed safety question exists.
Section 9.5.3 of the D. C.
Cook Nuclear Plant Updated Safety Analysis Report (USAR) states that the Component Cooling Water (CCM) system component design data are listed in Table 9.5-3.
Table 9.5-3 states that the shell side CCW heat exchanger outlet design water temperature is 95'F.
Contrary to the above, the Unit
CCW heat exchanger outlet water temperature as described in Table 9.5-3 of the USAR was authorized to be exceeded, and was actually exceeded, on January 27, 1994, without the required evaluation to determine if an unreviewed safety question existed.
Specifically, the Operations Department Superintendent authorized increasing the temperature limit to 105 F at 1915 hours0.0222 days <br />0.532 hours <br />0.00317 weeks <br />7.286575e-4 months <br /> on January 27, 1994, and subsequently the operator increased the temperature above 954F.
The temperature was allowed to increase to 110'F.
(315/94007-01 (DRS))
8.3.2 Violation Concerns and Contributin Factors The violation was for failure to perform a required safety evaluation of a proposed change in the facility, as described in the Updated Safety Analysis Report (USAR), to ascertain whether the proposed change involved an unreviewed safety question.
The violation is of concern because the USAR upper-limit design temperature of 95 F was authorized by the Operations Department Superintendent to be increased to 105>F for Unit 1 without a required safety evaluation and without a revision to the CCW operating procedure, which specified the same 95'F limitation.
Subsequently, an operator increased the CCW supply temperature beyond 954F and because of operator inattention, the CCM supply temperature increased to 110F during a transient induced by intermittent boric acid evaporator operation on January 27 and 28, 1994.
Formal calculations to evaluate the.safety significance of operating the CCW system at
F beyond the design temperature specified by the USAR and the operating procedure was not initiated by the licensee until this concern was identified by the inspectors, and the evaluation was not completed by the end of the inspection.
Subsequent to the inspection, the inspectors were informed by the licensee on Hay 10 and 13, 1994 of the extent the licensee's planned investigation and that the calculations could not be performed until the CCW
.design basis temperature could be determined.
Contributing factors to exceeding the CCW specified supply temperature included inadequate communications between the CCW system engineer and the operations department, and the failure of the operations department to reset the CCW supply (heat exchanger outlet) temperature alarm setpoint from 95~F to 105 F (inattention to details).
The system engineer should have informed the operations department that although engineering judgement indicated that 105'F CCW.supply water would not significantly damage the CCW system or the equipment it cooled, this temperature was beyond the approved design operational temperature of 95'F.
Also, the operations department should have been informed that the engineering department had not formally evaluated by calculational assessment the acceptability limits of operating the CCW system beyond the USAR stated acceptance criterion of 95 F.
In addition, if the operations department had reset the CCW supply temperature alarm to 105~F, the
operator would have been alerted that a
CCW system transient was occurring and might have controlled the situation before the temperature rose to 110'F.
E 9.0
~Ei II An exit meeting was conducted on April 22, 1994, at the D.
C.
Cook Nuclear Plant to discuss the major areas reviewed during the inspection, the apparent violation of NRC requirements, the weaknesses identified, and the other observations made during the inspection.
Licensee representatives and NRC personnel in attendance at this exit meeting are documented in Section 1.0 of this report.
The inspectors also discussed the likely informational content of the inspection report with respect to documents reviewed by the team during the inspection.
The licensee did not identify any documents or processes as proprietary.
Subsequent to the inspection, the inspectors continued discussions with licensee representatives by telephone through Hay 13, 1994.
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