IR 05000315/1999007
| ML17326A138 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/16/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17326A137 | List: |
| References | |
| 50-315-99-07, 50-316-99-07, NUDOCS 9909240030 | |
| Download: ML17326A138 (34) | |
Text
U.S. NUCLEAR REGULATORYCOMMISSION
REGION III
Docket Nos:
License Nos:
50-315; 50-316 DPR-58; DPR-74 Report No:
50-315/99007(DRS); 50-316/99007(DRS)
Licensee:
Indiana Michigan Power Company Facility:
Donald C. Cook Units 1 and 2 Dates:
July 12-30, 1999 August 16, 1999 Inspectors:
E. Duncan, Team Leader J. Gavula, Mechanical Engineering Inspector B. Fuller, Donald C. Cook Resident Inspector C. Baron, Mechanical Engineering Contractor F. Baxter, Electrical Engineering Contractor R. Cooney, Instrumentation and Controls Contractor O. Mazzoni, Electrical Engineering Contractor Approved by:
John M. Jacobson, Chief Mechanical Engineering Branch Division of Reactor Safety 9909240030 9909%6 PDR ADQCK OM003i5
EXECUTIVESUMMARY Donald C. Cook, Units 1 and 2 NRC Inspection Report 50-315/99007(DRS); 50-316/99007(DRS)
The purpose of this inspection was to validate the discovery phase of the licensee's Expanded System Readiness Review (ESRR) effort. In particular, the NRC evaluated whether the ESRR effort was effective in determining whether plant systems were capable of meeting their safety and accident mitigation functions as defined in the design and licensing bases.
To complete this obIective, the team independently reviewed the conformance of the residual heat removal (RHR) and 250 Volt direct current (Vdc) systems to the design and licensing bases, and compared those results to the results obtained through the ESRR effort. The inspection focused on other support systems as well.
~En ineerin Overall, the implementation of the ESRR process was considered effective. The scope of the review areas was broad and generally consistent with the purpose of the review effort to confirm the performance of system safety functions.
Further, the breadth and depth of material reviewed was appropriate overall and resulted in the identification of substantive issues.
Although some new issues were identified by the team which were within the scope of the ESRR effort, these issues represented isolated implementation weaknesses and not broad deficiencies in the ESRR process.
(Section E1.1)
There were several instances where the team identified issues in programmatic areas which were outside the ESRR review scope.
In addition, the team determined that a number of problems had only been broadly identified by the ESRR reviews, and the corrective action process was being relied upon to identify specific deficiencies.
As a result, the team concluded that although the ESRR reviews of the most safety-significant systems were complete, the total discovery effort was ongoing at the end of the inspection.
(Section E1.1)
The team identified an example where the licensee did not exercise appropriate sensitivity to the potential impact on operability of fuse control deficiencies identified during the ESRR review of the 250 Vdc system.
This was of particular concern since a problem regarding the sensitivity to operability of equipment required for Modes 5 and 6 was identified by the NRC following two electrical faults which occurred on April 19, and April24, 1999. Two examples of operability evaluations which failed to adequately address structural deficiencies were also identified. Collectively, these examples indicated that continued management attention is warranted in the operability evaluation area.
(Section E1.1)
~Back round Re ort Details Between July and September of 1997, the NRC conducted an architect engineering inspection coolin and at Donald C. Cook Units 1 and 2. The inspection focused primarily on th
'
and component cooling water systems and identified concerns regarding long-term core cooling during a design basis accident.
Based on questions from the NRC, American Electric Power (AEP) declared the recirculation sumps inoperable and shut down both units in accordance with technical specifications on September 9, 1997.
At the conclusion of the architect engineering inspection, AEP management committed to address and resolve, prior to plant restart, specific issues identified during the inspection.
Subsequently, the NRC issued a Confirmatory Action Letter identifying additional actions that would be taken by AEP prior to plant restart.
By early 1998, some progress had been made to resolve the Confirmatory Action Letter issues.
However, several new issues were identified which led to an AEP management decision to implement a more rigorous assessment of plant readiness for restart through the performance of System Readiness Reviews.
Also, the NRC established a D.C. Cook Restart Panel in that its r accordance with NRC Manual Chapter 0350, "Staff Guidelines for Re t rt A I,"
a i s review efforts were appropriate and provided objective measures of restart readiness.
To review the effectiveness of the System Readiness Reviews, the licensee conducted a Safety System Functional Inspection (SSFI) of the auxiliary feedwater system.
The SSFI identified further design and licensing basis issues which had not been detected through the System Readiness Review effort. Additional reviews identified additional design and licensing basis issues, potential weaknesses in engineering programs and processes, and potential vulnerabilities in the scope of systems reviewed during the System Readiness Reviews.
Accordingly, in January 1999, licensee management undertook additional actions to confront the scope of the problems at Donald C. Cook Units 1 and 2. These actions included the performance of Expanded System Readiness Reviews (ESRRs) to provide reasonable assurance that plant systems were capable of meeting their safety and accident mitigation functions as defined in the design and licensing bases.
r The ESRR program consists of four phases:
1) Initial Expanded System Readiness Reviews, 2) Restart ActivityMonitoring, 3) Final Expanded System Readiness Reviews, and 4) Startup and Power Ascension.
As discussed in NRC Inspection Report 50-315/99002(DRS);
50-316/99002(ORS), the NRC reviewed the ESRR program methodology and concluded that weaknesses identified in the original System Readiness Review process had been corrected and that the ESRR procedure described a systematic approach that was a substantial improvement over the original review guidelines implemented in 1998. As discussed in NRC Inspection Report 50-315/99003(DRS); 50-316/99003(DRS), the NRC performed reviews of the imp ementation of the ESRR program through a review of ESRR findings and concluded that the licensee was effectively identifying system deficiencies through the ESRR process, the threshold for identification of problems was conservatively low, and system walkdowns appeared effective at identifying design and configuration deficiencies.
The purpose of this inspection was to perform an independent validation of the d'sco h
e iscovery p ase o
he ESRR effort. To complete this objective, the team independently reviewed the
conformance of the residual heat removal (RHR) and 250 Volt direct current (Vdc) systems to the design and licensing bases, and compared those results to the results obtained through the ESRR effort.
ill. En lneerln E1 Conduct of Engineering E1.1 Review. of Residual Heat Removal S stem and 250 Vdc S stem a.
Ins ection Sco e 93801 The team independently reviewed the conformance of the RHR and 250 Vdc systems to the design and licensing bases, and compared those results to the results obtained through the ESRR effort. To accomplish this objective, the team focused on the identification of the design and licensing bases of the RHR and 250 Vdc systems, and verified conformance with these requirements through a review of documentation such as modifications, calculations, operability evaluations, and surveillance tests; completion of system walkdowns; and interviews with plant staff. The inspection focused on other support systems as well.
~ b.
Observations and Findin s b.1 The team reviewed the Emergency Core Cooling System (ECCS), RHR-Shutdown Cooling, and 250 Vdc ESRR Initial Reports and determined that the ESRR effort had identified a number of significant issues related to the RHR and 250 Vdc systems which included the following:
The RHR pumps and motors had been operated for an extensive number of hours without effective preventive maintenance or condition monitoring.
Operational procedures for the RHR system during reduced inventory operation were not supported by design calculations and/or testing.
Calculations associated with the 250 Vdc system did not support the design and licensing basis requirements for accident mitigation with respect to battery sizing,
- load coordination, voltage drop, and ampacity.
Additionally, a clear basis for the N-train battery could not be identiTied.
The programmatic failure of the fuse control program resulted in the installation of numerous unqualified fuses in the 250 Vdc system.
Direct current component load control deficiencies had adversely impacted calculations, testing, and surveillance procedures.
Technical Specification surveillance requirements for the 250 Vdc system conflicted with the licensing and design bases.
Operational procedures for the 250 Vdc system conflicted with design basis calculation b.2 Review Results The team identified issues that were within the scope of the ESRR, but were not found during the ESRR review effort. The team also identified issues which were outside the scope of the ESRR process.
Issues Identified Within the ESRR Sco e
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Safety Injection System Flow Balance Procedure. Deficiencies Procedure 12 EHP 4030.STP.208SI,
"Unit 1 and Unit 2 ECCS Flow Balance-Safety Injection System," veriTied adequate safety injection system flow to meet the requirements of Technical Specification 4.5.2.h, "Emergency Core Cooling System - Tavg Greater Than or Equal to 350 F." The team identified that the procedure did not account for the potential differences in the safety injection system suction and discharge water elevation between the test configuration and the postulated post-accident configuration.
In accordance with procedure prerequisite 3.3, refueling water storage tank (RWST) level was required to be at least 19 percent to ensure adequate pump net positive suction head during the flow balance test.
Procedure step 5.1.13 addressed maintaining the water level in the reactor cavity during the test, but did not include any specific limits. However, the minimum and maximum flow acceptance criteria were based on the flow limits of Technical Specification 4.5.2.h, which included an allowance for test instrument uncertainty, but did not address RWST and reactor vessel level. Therefore, since the initial RWST and reactor vessel level during a design basis accident could significantly differ from the test configuration, this could result in a safety injection system flow rate not meeting Technical Specification 4.5.h.2 flow limits.
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
requires that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances.
The failure to account for the effects of RWST level and reactor vessel level in safety injection pump flowtest procedure 12 EHP 4030.STP.208SI was an example where the requirements of 10 CFR 50, Appendix B, Criterion V, were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-01(DRS);50-316/99007-01(DRS)),'onsistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as condition report (CR) P-99-19540.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NC RHR System Walkdown Deficiencies During a walkdown of the Unit 2 RHR system, the team noted that the 3-inch minimum flow line for the west RHR pump was poorly restrained.
In addition, the team identified chipped concrete at the structural attachment for RHR minimum flow line pipe support 2-GRH-L808. Although the ESRR team noted that the strut associated with this support and the stanchion-plate associated with an adjacent support were slightly misaligned, they had not identified the chipped concrete.
In addition, when structural support deficiencies are viewed in the aggregate, it is apparent that portions of the RHR system had experienced a hydraulic transient or some other event which caused these misalignments.
During followup discussions with operations personnel, it appeared to be common knowledge that the RHR system had historically experienced hydraulic transient events when realigning the system from shutdown cooling to ECCS standby readiness.
In addition, Procedure 01-OHP 4021.017.003, "Removing Residual Heat Removal Loop From Service," Revision 9, dated November 12, 1998, added precautionary steps intended to prevent hydraulic transients during realignment from shutdown cooling to ECCS standby readiness.
10 CFR 50, Appendix B, Criterion XVI,"Corrective Action," requires that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and nonconformances, are promptly identified and corrected.
The failure to identify a hydraulic transient on the Unit 2 RHR system was an'example where the requirements of 10 CFR 50, Appendix B, Criterion XVI,were not met and was a violation. However, this Severity Level IV violation is being treated as a Non-Cited Violation (50-315/99007-02(DRS);
50-316/99007-02(DRS)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR P-99-20026.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NCV.
Also during the walkdown, the team identifie that a nut was missing on an anchor bolt for a structural beam clip angle in the Unit 2 east RHR heat exchanger room. Although primarily used to support the personnel access platform, the structural beam was also used to support a 2-inch safety-related pipe. The licensee generated CR P-99-20014 to identify this issue for entry into the corrective action progra Valve Inservice Testing Program Classification Errors The team reviewed the RHR and safety injection system portions of the "Valve Inservice Test Program - Third Ten Year Interval," Revision 1, and questioned the safety function of several valves.
The program identified the safety function of motor-operated valves IMO-340 and IMO-350 as "0/c", indicating an active safety function in the open direction and a passive safety function in the closed position. These normally closed valves isolate the centrifugal charging pump and safety injection pump suction headers from the RHR pump discharge headers.
During the transfer from'CCS injection mode to recirculation mode the valves would be opened prior to depleting the refueling water storage tank to establish a flow path from the containment sump.
The active safety function in the open direction appeared appropriate.
However, Section 6.1.1, "Engineered Safety Features Criteria," of the Updated Final Safety Analysis Report specified the capability of the ECCS to withstand a single passive failure during recirculation. Therefore, these valves also had an active safety function to close to isolate a passive ECCS failure during the recirculation mode.
The program identified the safety function of check valves Sl-104N and Sl-104S as "0", indicating an active safety function in the open direction only. These 3/4-inch check valves are located in the safety injection pump minimum flow lines.
Downstream of these check valves the minimum flow lines join together.
Since these check valves were required to open to provide minimum flow protection for the pumps during ECCS injection mode operation, the active safety function in the open direction appeared appropriate.
However, these check valves also had an active safety function to close to isolate the discharge of an operating safety injection pump following a passive ECCS failure at the other safety injection pump during the recirculation mode.
The program did not include normally open manual valves Sl-103N and SI-103S, located in the safety injection pump suction header.
As shown on flow diagram OP-1-5142-34, "Flow Diagram - Emergency Core Cooling," Revision 34, these valves were provided with reach rods.
Licensing personnel stated that these valves would be manually closed to isolate the suction of an operating safety injection pump from a passive failure occurring at the other safety injection pump during recirculation mode. Therefore, these valves had an active safety function to close.
10 CFR 50, Appendix B, Criterion XVI,"Corrective Action," requires that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and nonconformances are promptly identified and corrected.
The failure to identify that several valves were not appropriately categorized in the inservice testing program was an example where the requirements of 10 CFR 50, Appendix B, Criterion XVI,were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-03(DRS);50-316/99007-03(DRS)),
consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR P-99-18746.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit
in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts.
underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NCV.
Emergency Diesel Generator (EDG) Fuel Oil Tank Sizing Issues The team reviewed EDG Fuel Oil Storage Tank calculation C-7, "Emergency Diesel Generator Fuel Consumption Rate." The calculation determined that the minimum fuel oil storage tank volume required to support one EDG operating continuously at design power output for 7 days was 43,240 gallons. The team identified that the calculation did not account for the unuseable tank volume due to suction line location. The licensee generated CR P-99-19773 to identify this issue for entry into the corrective action program.
The team also identified that in the event of a loss-of-coolant-accident (LOCA) on one unit with a loss of offsite power (LOOP) affecting both units, and without operator intervention, both EDGs on both units were designed to start and operate continuously.
The team questioned the licensing basis of the plant since the EDG fuel oil storage tanks were only designed to be capable of supplying one EDG operating at design power output for 7 days.
The design of the EDGs and associated fuel oil system was in accordance with Regulatory Guide 1.137, "Fuel Oil Systems for Standby Diesel Generators,"
Revision 1. Section C of Regulatory Guide 1.137 endorsed American National Standards Institute (ANSI) N195-1976 which defined the requirements for fuel oil systems that have components which are shared between units. This was applicable to the D.C. Cook facilitysince there are two EDG fuel oil storage tanks on site and each tank serves one EDG for each unit. Standard ANSI 195-1976 required that the EDG fuel oil tanks be sized with the capability to mitigate a design basis accident on one unit, operate the equipment necessary to safely shutdown the other unit, and maintain both units in a safe shutdown condition.
In order to meet these requirements, operator action would be required to secure one EDG on each unit following the LOCNLOOP event. The licensee generated CR P-99-19705 to determine whether adequate procedure guidance was provided regarding this required operator action.
Subsequently, the team determined that in the event of a single failure on each unit associated with an EDG fuel oil storage tank, both remaining EDGs would be required to operate from the other EDG fuel oil storage tank and therefore may not be able to meet the requirements of ANSI 195-1976 and Regulatory Guide 1.137. At the end of the inspection, the licensee had not determined whether the licensing basis required that a single active failure be assumed for each unit or the entire site. The licensee generated CR P-99-19773 to identify this issue for entry into the corrective action progra This is an Unresolved Item (50-315/99007-04(DRS); 50-316/99007-04(DRS))
pending NRC review of revised EDG fuel oil storage tank volume calculations and resolution of plant licensing basis questions.
600 Volt Alternating Current (Vac) Cable Sizing The team reviewed the 600 Vac system and identified that the system was established in an ungrounded configuration, and that power cables were insulated cables rated at 600 Vac. The team also reviewed Section 8, "Electrical Systems,"
of the Updated Final Safety Analysis Report and identified that motors and electrical switchgear enclosures were procured to conform to the specifications issued by the National Electrical Manufacturers Association (NEMA).
The team reviewed NEMA Publication WC 5-1968, "Thermoplastic-Insulated Wire and Cable for the Transmission and Distribution of Electrical Energy," and NEMA Publication WC 3-1969, "Rubber-Insulated Wire and Cable for the Transmission and Distribution of Electrical Energy," and identified certain cable insulation sizing requirements for ungrounded systems.
Both standards required that the selection of the cable insulation level to be used in a particular installation shall be made on the basis of the applicable phase-to-phase voltage and the following general system categories:
1) 100 percent level - cables in this catego'ry may be applied where the system is provided with relay protection such that ground faults will be cleared as quickly as possible, but in any case within 1 minute; 2) 133 percent level - cables in this category may be applied in situations where the clearing time requirements of the 100 percent category cannot be met, and yet there is adequate assurance that the faulted section willbe de-energized within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; and 3) 173 percent level -,cables in this category should be applied in systems where the time required to de-energize a grounded section is indefinite.
The team discussed these standards with licensee personnel.
As a result, CR P-99-19650 was generated to perform a review of the 600 Vac ungrounded system design to determine whether the cable insulation levels were adequate.
At the end of the inspection, the licensee determined that the 600 Vac cable insulation thickness procured was equivalent to the 173 percent level specified by NEMA.
This is an Unresolved Item (50-315/99007-05(DRS);50-316/99007-05(DRS))
pending NRC review of information regarding cable insulation levels.
600 Vac Safety-Related Bus Single Breaker Failure Vulnerability The team identified that in the event that either of the normally open cross-tie breakers associated with 600 Vac safety-related buses 11A and 11C, or 11B and 11D, were to fail to the closed position with the EDGs supplying power to the buses, both EDGs would be synchronized out-of-phase, resulting in the loss of both EDGs.
The team subsequently determined that although the ESRR effort failed to identify this vulnerability, it had been independently identified as part of another licensee initiativ ~
Net Positive Suction Head (NPSH) Calculation Assumption Error The team reviewed calculation ENSM970128AF, "ECCS Pumps Available NPSH,"
Revision 2. This calculation determined the NPSH available to the RHR, containment spray, safety injection, and centrifugal charging pumps during ECCS recirculation mode and assumed a containment pressure of 14.7 pounds per square inch absolute (psia) during ECCS recirculation operation.
In accordance with Safety Guide 1, "NPSH for ECCS and Containment Meat Removal Pumps,"
dated November 2, 1970, NPSH should be based on the minimum containment pressure present prior to the postulated LOCA. Technical Specification 3.6.1.4,
"Containment Systems - Internal Pressure," allowed a minimum containment pressure of -1.5 pounds per square inch gauge (13.2 psia) during normal plant operation.
The team questioned whether the NPSH analysis should be based on a containment pressure of 13.2 psia to meet Safety Guide 1, instead of 14.7 psia.
Followup discussions determined that although other errors associated with this calculation had been identified by the ESRR team, this particular issue had not been discovered.
Since the plant had not been historically operated at a negative containment pressure, the team concluded that this error was of only minimal safety significance.
The licensee generated CR P-99-18718 to identify this issue for entry into the corrective action program.
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RHR Pump Start/Stop Tracking Deficiencies The team identified that although Section 4.1.3.2 of design basis document DB-12-RHRS and Westinghouse document E-677125, "AuxiliaryPumps,"
Revision 1, stated that the RHR pumps were designed for 50 starts/stops per year, starts/stops had not been tracked.
The licensee generated CR P-99-18323 to identify this issue for entry into the corrective action program.
In addition, DB-12-RHRS stated that the RHR pump motors were designed to be shutdown and restarted once, providing the motor was allowed to coast down.
The team identified that during a LOCA followed by a LOOP, sufficient coastdown may not occur and pump damage could result. The licensee generated CR P-99-19156 to determine whether a LOCA followed by a LOOP was within the design and licensing bases.
However, since RHR pumps were not frequently started and stopped, and interviews of plant operations personnel indicated that in the event of a LOOP, pump restart would not occur until after coastdown was completed, the team concluded that this issue was of only minimal safety significance.
Operator Training on Emergency Operating Procedure Revisions The team identified that operator training to address ESRR concerns regarding Emergency Operating Procedure (EOP) ES-1.3, "Transfer to Cold Leg Recirculation," was classified as post-restart although the desired operator response to a design basis accident was revised.
Subsequently, EOP ES-1.3 revision training was re-classified as restart-required.
In addition, the licensee verified that other EOP revision training plans were appropriately classified as restart-required.
The team concluded that since the mls-classification only impacted one EOP, the issue was of only minimal safety significance.
Licensee personnel stated that the issue was classified as post-restart by the ESRR System Readiness Review Board because the issue did not fit into any of the specific Restart Criteria. Subsequently, the team questioned whether the Restart Criteria employed to classify issues as restart-required or post-restart were adequate, since this restart-required issue was not appropriately classified using the Restart Criteria that existed.
The licensee generated CR P-99-19264 to identify this issue for entry into the corrective action program.
Issues Identified Outside the ESRR Review Sco e
During the inspection, the team identified a number of issues which were outside the scope of the ESRR review. These included technical issues as well as operability evaluation program implementation concerns.
Technical Issues During the inspection, the team identified a number of issues which were outside the ESRR review scope since these issues were within the programmatic area review effort or were expected to be identified as part of the corrective actions to address previously-identified broad scope problems.
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Motor-Operated Valve Testing Concerns The team reviewed calculation HPX910726EVG, "RHR MOV[Motor-Operateg Valve] Delta-P Calculation," dated August 9, 1991. This calculation was performed to determine the maximum differential pressure across motor-operated valves in the RHR system and included cooldown isolation valves IMO-128 and ICM-129, as well as containment sump isolation valves ICM-305 and ICM-306.
The calculation for maximum opening differential pressure across IMO-128 and ICM-129 did not consider the potential for fluid trapped between these valves to become pressurized due to an increase in temperature.
In addition, the evaluation of containment sump isolation valves ICM-305 and ICM-306 did not account for the containment pressure expected during an accident.
Although the specific issues identified by the team had not yet been identified, issues concerning the Generic Letter 89-10 motor-operated valve program and inservice testing program had been broadly identified and the extent of the deficient condition in the motor operator valve program would have been conducted pursuant to the Corrective Action Program and may have identified this issue.
10 CFR 50, Appendix B, Criterion!II, "Design Control," requires that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
The failure to adequately consider the effects of elevated containment temperature and pressure on the differential pressure across the RHR system valves discussed above was an example where the requirements of 10 CFR 50, Appendix B, Criterion III, were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-06(DRS); 50-316/99007-06(DRS)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR P-99-18607.
Appendix C of the Enforcement Policy requires that
for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NCV.
EDG Loading Calculation Review The team reviewed calculation PS-EDGE-002, "Computer Simulation Study to Evaluate Dynamic Performance of Cook Emergency Diesel Generator System,"
Revision 2. Table 2.8 of the calculation indicated that a block load was assumed to start 10 seconds into the EDG starting sequence.
The team questioned the conservatism of this assumption since many of the individual loads incorporated into the overall block load were each controlled by individual process signals and therefore could start at any time during the EDG starting sequence.
'I Licensee personnel indicated that this calculation issue was outside the scope of the ESRR effort since calculations were not reviewed in detail.
In addition CR P-99-15977 had been generated to establish the design basis for the EDG, and it was expected that this issue would have been identified.
10 CFR 50, Appendix B, Criterion III, "Design Control," requires that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
The failure to fullyconsider the potential EDG loading sequence of various plant loads during EDG operation is an example where the requirements of 10 CFR 50, Appendix B, Criterion III,were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-0?(DRS);50-316/99007-07(DRS)),
consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR P-99-18433.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention. Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NCV.
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High Energy Line Break (HELB) Barrier Concern During a walkdown on June 23, 1999, the team noted that auxiliary building door 1-DR-AUX-392 was propped open and one half of the double door had been removed.
This door was identified as both a fire barrier and a HELB barrier.
There was a sign indicating that the fire barrier was open and that a roving fire watch had been posted.
However there was no indication that breaching of the HELB barrier had been recognized.
Subsequently, the team was provided a copy of Procedure 12 PMP 4030.001.002, "Administrative Requirements for Ventilation Boundary and High Energy Line Break Barriers," Revision 1. This procedure stated that HELB doors were only'required to be functional when either of the units was in Modes
through 4, and included a requirement that the HELB doors be walked down prior to ascending into Mode 4. The team found these aspects of the procedure to be appropriate.
However, Section 4.6.1 of the procedure allowed the use of a temporary plywood barrier in place of a HELB door that was determined to be non-functional or had been removed to perform work. The team questioned the technical basis of this temporary installation and requested the analysis which verified the adequacy of the temporary plywood barrier.
Licensee personnel stated that an analysis for the temporary plywood barrier could not be located, and that this procedure would be revised as part of the HELB program review.
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
requires that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances.
The failure to have an adequate procedure regarding the installation of temporary HELB barriers is an example where the requirements of 10 CFR 50, Appendix B, Criterion V, were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-08(DRS); 50-316/99007-08DRS)),
consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR P-99-19850.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicit in that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue will be dispositioned as an NCV.
~
RHR Piping Drawing Deficiencies The team reviewed flowdiagrams OP-1-5142-34, "Flow Diagram - Emergency Core Cooling," Revision 34, and OP-1-5144-33, "Flow Diagram - Containment Spray Unit 1," Revision 33. These drawings identified the 2-inch safety injection pump minimum flow line as inservice inspection (ISI) Class 3 between valve IMO-263 and the RWST nozzle. This was inconsistent with the remainder of the ECCS and the RWST piping which was ISI Class 2.
Licensee personnel stated that this ISI Class 3 classification was incorrect and generated CR P-99-19833 to identify this issue for entry into the corrective action program. This CR also identified several other examples of incorrect ISI boundary classifications that had been previously documented.
Because the piping identified by the team was exempt from inservice inspection due to the small size involved, this issue was of only minimal safety significance.
It was also outside the scope of the ESRR effort since the ISI program review was expected to identify these issues and had not been completed when this issue was identified.
Thread Engagement Issues During a walkdown of the 250 Vdc system, the team identified less than full thread engagement on three anchors associated with the Unit 1 N-train battery rack.
However, since a generic thread engagement issue had been previously identified, and extensive system walkdowns were planned to identify additional thread engagement deficiencies, these particular deficiencies were outside the scope of the ESRR review. The licensee generated CR P-99-16668 to identify this issue for entry into the corrective action program.
UFSAR Discrepancy Regarding N-Train Battery Alarm During a review of Updated Final Safety Analysis Report (UFSAR) Section 8.3.5, the team identified that a "Battery Charger Contactor Tripped" alarm for the N-train battery was described to be present in the control room although no such alarm actually existed.
The licensee generated CR P-99-19945 to identify this issue for entry into the corrective action program.
The licensee stated that although the ESRR effort was not intended to identify UFSAR discrepancies, it was within the scope of the UFSAR validation project which was in progress at the end of the inspection.
0 erabili Evaluation Pro ram Concerns The team identified examples where the staff did not exercise appropriate sensitivity to the potential impact on operability of identified deficiencies.
Although this was unrelated to the ESRR effort, it was of particular concern since a problem regarding the sensitivity to operability of equipment required for Modes 5 and 6 was identified by the NRC following two electrical faults which occurred on April 19, and April24, 1999. The following specific issues were identified:
Operability Evaluations Concerning Fuse Coordination and Fuse Replacements On June 22, during a presentation to the team regarding the overall findings and conclusions from the 250 Vdc system ESRR review, licensee personnel stated that there was insufficient assurance that the 250 Vdc system was capable of meeting its safety and accident mitigation functions for design basis accident mitigation and station blackout in its existing configuration in Modes 1 through 4.
This conclusion was based primarily on a lack of design basis system calculations, load control deficiencies, equipment qualification concerns, Regulatory Guide 1.97 compliance concerns, and fuse control problems.
Problems related to fuse control included fuse coordination concerns and the installation of Gould Shawmut fuses which differed from the Bussman fuses tested during plant construction.
The licensee also concluded that based on battery performance and other administrative controls, there was reasonable assurance that the 250 Vdc system was capable of meeting its safety and accident mitigation functions in Modes 5 and 6. The team requested an operability evaluation to support this conclusion since it appeared that the fuse control problems which contributed to conclusions made with regard to Modes 1 through 4 could also apply in Modes 5 and 6. The licensee determined that an operability evaluation for Modes 5 and 6 had not been performed and generated CR P-99-16829 to identify this issue for entry into the corrective action program.
Subsequently, the licensee completed an operability evaluation which the team reviewed.
The team was concerned regarding the depth of review to determine operability since it relied on a baseline auxiliary electrical coordination study that was performed in 1988 and had been identified as potentially inaccurate.
The team reviewed the coordination study further and identified that the study did not consider the following issues:
~
System Voltage Characteristics In general, Gould Shawmut fuses were rated by the vendor at 250 Vdc.
However, the 250 Vdc system float voltage was normally maintained at 262 Vdc and in an equalizing condition was specified in plant operating procedures to be maintained at 275 to 280 Vdc. Since the Gould Shawmut vendor only tested these fuses at 250 Vdc, operability at higher voltages had not been evaluated.
Fuse Failure Characteristics in the 3-30 Ampere Range The most recent Gould Shawmut direct current (DC) voltage rating information sheet identified that fuses in the 3-30 ampere range were qualified for up to 160 Vdc. Voltages above this limitcould result in casing and end cap damage when interrupting a fault. Calculation PS-FUSE-001, "AC [Alternating Current] Rated Fuses in DC Applications,"
Revision 0, stated that the fuse casing failure was acceptable in the DC system due to the fuse panel configuration and nonexplosive environment these fuses were applied in. However, the team identified that the effects of molten solder and ionized gases emitted that may cause damage to surrounding circuits had not been considered.
~
Fuse Interrupting Time Consideration Although the similarity between melting time curves in AC and DC applications was recognized, the total arc time differences between AC and DC applications which contributes to the overall fuse clearing time following a fault was not considered.
At the end of the inspection, the licensee generated CR P-99-19746 to identify these deficiencies and planned to re-perform the operability evaluation using the new information identified by the team.
Operability Evaluation Practices Regarding Piping Support Deficiencies After identifying that the RHR minimum flow line did not have adequate seismic supports, as discussed in the "RHR System Walkdown Deficiencies" section of this report, the team reviewed the bases for the operability of the system in the current configuration. Licensee personnel provided a copy of letter AEP:NRC:1238C1, dated November 15, 1996, which referred to an operability screening specification documented in ES-PIPE-1002-QCN, "Operability Screening Guideline for Pipe Support Conditions and Discrepancies Found by In-Service Inspection," Revision 0, dated November 25, 1998. The specification stated, through references to other correspondence, that a documented operability evaluation was not required ifthe identified support discrepancies met the operability screening criteria. For the condition that these discrepancies represented, the operability screening criteria state'd, "forany combination of conditions/discrepancies on any one support or on any two consecutive supports or on any non-consecutive supports, the continued operability is not impaired."
These criteria did not require consideration of support types or directions and was excessively broad to justify not needing a formal operability evaluation.
The licensee generated CR P-99-20046 to identify this issue for entry into the corrective action program.
~
Operability Evaluation Deficiency While reviewing the resolution of support 2-GRH-L808 misalignment identified in CR-99-06828, the team noted that the operability evaluation discussed pipe support and pipe stress margins from the latest piping analysis.
However, a number of pipe supports had been added in the analysis to resolve design basis stress deficiencies which had not been installed in the field. Therefore, the current operability was based on a future configuration and not on the existing configuration. The licensee generated,CR P-99-20045 to identify this issue for entry into the corrective action program.
10 CFR 50, Appendix B, Criterion XVI,"Corrective Action," requires that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and nonconformances are promptly identified and corrected.
The failure to perform satisfactory operability evaluations for the examples discussed abov'e was an example where the requirements of 10 CFR 50, Appendix B, Criterion XVI,were not met and was a violation. However, this Severity Level IVviolation is being treated as a Non-Cited Violation (50-315/99007-09(DRS);50-316/99007-09(DRS)),
consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CRs P-99-16829, P-99-20046, and P-99-20045.
Appendix C of the Enforcement Policy requires that for Severity Level IVviolations to be dispositioned as NCVs, they be appropriately placed in the licensee's corrective action program.
Implicitin that requirement is that the corrective action program be fullyacceptable.
The D.C. Cook corrective action program was not adequate and has been the focus of significant attention.
Although the licensee and the NRC have not yet concluded that the corrective action program is fullyeffective, the improvement efforts underway are captured in the D.C. Cook Restart Plan which is under the formal oversight of the NRC through the NRC Manual Chapter 0350 process, "Staff Guidelines for Restart Approval." Consequently, this issue willbe dispositioned as an NCV.
c.
Conclusions Overall, the implementation of the ESRR process was considered effective. The scope of the planned review areas was broad and generally consistent with the purpose of the review effort to confirm the performance of system safety functions.
Further, the breadth and depth of material reviewed was appropriate and the ESRR teams were effective at identification of substantive issues.
Although some new issues were identified by the team which were within the scope of the ESRR effort, these issues represented isolated implementation weaknesses and not broad deficiencies in the ESRR process.
There were several instances where the team identified issues in programmatic areas
'which were outside the ESRR review scope.
In addition, a number of problems had only been broadly identified by the ESRR reviews, and the corrective action process was being relied upon to identify specific deficiencies.
As a result, the team concluded that although the ESRR reviews of the most safety-significant systems were complete, the total discovery effort was ongoing at the end of the inspection.
The team also concluded that the implementation of an effective corrective action program was essential to ensure that problems that have already been discovered are corrected as well as to identify and correct specific problems that have as yet only been broadly identified.
" The team also identified an example where the licensee did not exercise appropriate sensitivity to the potential impact on operability of fuse control deficiencies identified during the ESRR review of the 250 Vdc system.
This was of particular concern since a problem regarding the sensitivity to equipment required for Modes 5 and 6 was identified by the NRC following two electrical faults which occurred on April 19, and April24, 1999.
Two examples of operability evaluations which failed to adequately address structural deficiencies were also identified. Collectively, these examples indicated that continued management attention is indicated in the operability evaluation area.
V. Mana ement Meetin s X1 Exit Meeting Summary The team presented the inspection results to members of licensee management at the conclusion of the inspection on August 16, 1999. The licensee acknowledged the findings presented.
Although some proprietary information was reviewed during the inspection, none of the information resulted in issues which are discussed in this report.
M. Finissi S. Greenlee S. Lacey J. Pollock R. Powers M. Rencheck L. Thornsberry K. Van Dyne A. Zarechnak PARTIALLIST OF PERSONS CONTACTED Assistant Director - Plant Engineering Manager - Design Engineering Director - Engineering Restart Electrical Design Engineering Senior Vice President Vice President - Engineering Manager - System Engineering Regulatory Affairs Expanded System Readiness Review INSPECTION PROCEDURES USED IP 93801 Safety System Functional Inspection
~Oened ITEMS OPENED, CLOSED, AND DISCUSSED 50-315/99007-01;50-316/99007-01 NCV 50-315/99007-02;50-316/99007-02 NCV 50-315/99007-03;50-316/99007-03 NCV 50-315/99007-04;50-316/99007-04 URI 50-315/99007-05;50-316/99007-05 URI 50-315/99007-06;50-316/99007-06 NCV 50-315/99007-07;50-316/99007-07 NCV 50-315/99007-08;50-316/99007-08 NCV 50-315/99007-09;50-316/99007-09 NCV Closed 50-315/99007-01;50-316/99007-01 NCV 50-315/99007-02;50-316/99007-02 NCV 50-315/99007-03;50-316/99007-03 NCV 50-315/99007-06;50-316/99007-06 NCV 50-315/99007-07;50-316/99007-07 NCV 50-315/99007-08;50-316/99007-08 NCV 50-315/99187-09;50-316/99007-09 NCV Discussed None Deficient Safety Injection Flow Balance Procedure RHR System Walkdown Deficiencies RHR System Walkdown Deficiencies EDG Fuel Oil Storage Tank Licensing Basis 600 Vac Cable Sizing Motor-Operated Valve Testing Concerns EDG Loading Calculation Concerns High Energy Line Break Procedure Deficiencies Operability Evaluation Deficiencies Deficient Safety Injection Flow Balance Procedure RHR System Walkdown Deficiencies RHR System Walkdown Deficiencies Motor-Operated Valve Testing Concerns EDG Loading Calculation Concerns High Energy Line Break Procedure Deficiencies Operability Evaluation Deficiencies
AC AEP ANSI CFR CR DC DRS ECCS EDG EOP ESRR HELB ISI kv LOCA LOOP MOV NCV NEMA NPSH psia RHR RWST SSFI UFSAR URI Vac Vdc LIST OF ACRONYMS USED Alternating Current American Electric Power American National Standards Institute Code of Federal Regulations Condition Report Direct Current Division of Reactor Safety Emerg'ency Core Cooling System Emergency Diesel Generator Emergency Operating Procedure Expanded System Readiness Review High Energy Line Break Inservice Inspection Kilovolt Loss-of-Coolant-Accident Loss of Offsite Power Motor-Operated Valve Non-Cited Violation National Electrical Manufacturers Association Net Positive Suction Head Pounds Per Square Inch Absolute Residual Heat Removal Refueling Water Storage Tank Safety System Functional Inspection Updated Final Safety Analysis Report Unresolved Item Volts Alternating Current Volts Direct Current
PARTIALLIST OF DOCUMENTS REVIEWED The following is a list of licensee documents reviewed during the inspection.
Inclusion on this list does not imply that NRC inspectors reviewed the documents in their entirety, but rather that portions or selected portions of the documents were evaluated as part of the overall inspection effort. NRC acceptance of the documents or any portion thereof is not implied.
Procedures PMP 5030.001.002, Rev. 3, Control of Critical Parameters PMP 7030.CAP.001, Rev. 1, Corrective Action Program Process Flow PMP 7030.OPR.001, Rev. 2, Operability Determinations PMP 7200.RST.004, Rev. 5a, Expanded System Readiness Review Program PMP 7200.RST.009, Rev. Oa, Programmatic Restart Readiness PMP 7200.RST.010, Rev. 1a, Functional Area Restart Readiness PMP 6065.ISP.001, Rev. 1, Plant Instrument Setpoint Program 01-OHP 4021.082.011, Rev. 7, Locating 600 Vac Grounds 01-OHP 4021.017.003, Rev. 9, Removing Residual Heat Removal Loop from Service 01-OHP 4023.ES-1, Rev. 7, Loss of Reactor or Secondary Coolant 01-OHP 4023.ES-1.1, Rev. 9, Safety Injection Termination 01-OHP 4023.ES-1.2, Rev. 5, Post LOCA Cooldown and Depressurization 01-OHP 4023.ES-1.3, Rev. 5, Transfer to Cold Leg Recirculation 02-OHP 4030.STP.053A, Rev. 13, ECCS Valve Operability Test - Train A 12-OHP 4021.082.012, Rev. 2, Location of 250 Vdc Grounds 12-EHP 4030.STP.208SI, Rev. 2, ECCS Flow Balance - Safety Injection System 12-EHP 5070.ISI.017R, Rev. 3,Section XI Centrifugal Charging Pump Performance Verification 12-EHP 5040.DES.003, Rev. 1, Calculations 12-PMP 4030.001.002, Rev. 1, Administrative Requirements for Ventilation Boundary and High Energy Line Break Barriers Modifications DC-12-2154, Provide Interlocking for the RHR Pump Motors to Cause Tripping on Low-Low RWST Level when Suctions are Lined up to Tank, dated February 23, 1978.
DC-12-2387, Revise Safety Injection Activation Logic from Low Pressurizer Pressure Coincident
, with Low Pressurizer Level (1/3) to Low Pressurizer Pressure (2/3), dated March 26, 1980 DC-12-2395, Modifythe Safeguards Actuation and Reset Circuits to Prevent Inadvertent Over-Riding of Safety Actuation Signals, dated September 14, 1979.
DC-01-2629, Repairs Performed on the Unit 1 North Safety Injection Pump, dated January 13, 1983.
DC-12-2651, Replace the Safety Injection Pump Minimum Flow Orifice Assemblies with New Orifice Assemblies Designed to Pass Increased Flow of 60 gpm, dated January 17, 1984.
12-DCP-183, Remove Auto-Close Feature From Valves IMO-128 and ICM-129, dated December 31, 1998.
RFC DL-12-2154, Provide Interlock for RHR Motor Trip on RWST Low Level, dated October 17, 1977.
RFC DL-12-2387, Revise Safety Injection Actuation Logic from Low Pressurizer Press/Low Level to 2/3 Low-Low Pressure, dated April 12, 1980.
12-DCP-0853, Modification to ILS-950 and -951, dated September 24, 1997.
RFC 1-2186, Unit 1 AuxiliaryFeedwater Changes, dated July 1979.
RFC 12-1729, Plant Batteries - Add Overvoltage Relays, dated November 10, 1986.
RFC 12-2602, Relocate Undervoltage Detection on Plant Batteries, dated August 7, 1984.
RFC 12-2843, Plant Batteries 1-AB, 1-CD, 2-AB, 2-CD Undervoltage Relay Settings, dated February 6, 1987.
RFC 1-2222, Unit 1 Turbine Driven AuxiliaryFeedwater Valves to DC Power, dated January 12, 1980.
RFC 1-2927, Unit 1 Replacement of AB Battery Chargers, dated March 23, 1989.
02-MM-134, Broken Conductor 2-13351-2 on Valve 2-ICM-306, dated July 13, 1999.
02-MM-166, Install an ArtiTicialLeak By on Safety Injection Pump Discharge Check Valve 2-S1-110S, dated September 14, 1993.
Calculations ENSM961212CV, Centrifugal Charging Pump Minimum Flow LOCA, dated December 13, 1996.
ENSM970128AF, Rev. 2, ECCS Pumps Available Net Positive Suction Head.
ENSM970606JJR, Rev. 2, Refueling Water Storage Tank Vortexing.
ENSM971016AF, RHR Deadheading, dated October 25, 1997.
ENSM971208RWH, Impact of RHR Heat Exchanger Partition Plate Deformation, dated December 20, 1997.
ENSM980817QSL, Injection Flowrate Sensitivity, dated September 22, 1998.
HPX840301JN, Rev. 0, Net Positive Suction Head Calculations.
HPX881024AF, Rev. 0, Determine ifOne RHR Pump Can Supply the Safety Injection and Centrifugal Charging Pumps During the Recirculation Phase of Safety Injection Without Cavitating.
HPX901012J JR, ECCS/RHR IMO 330/331 Delta-P, dated October 18, 1990.
HPX910726EVG, RHR Motor-Operated Valve Delta-P Calculation, dated August 9, 1991.
HXP910829EVG, RHR Operational Mode Temperatures, dated September 18, 1991.
HXP921021EVG, RHR Operational Mode Temperatures, dated October 30, 1992.
NEMP950501 JEW, RHR, Safety Injection, Containment Spray - NPSH Available, dated October 22, 1997.
N931001, UAValues of Closed Cooling Water and RHR Heat Exchangers During Recirculation Phase of LOCA, dated November 1, 1993.
ECP 1-2-I3-01, Rev. 1, RHR Flow for IFC-315 and -325, dated Novemb'er 9, 1987.
1-ICP-00159, I&CChange Package 1-IFC-315 and -325, dated December 3, 1998.
2-ICP-00160, l&C Change Package 2-IFC-315 and -325, dated December 3, 1998.
ECP 1-2-I3-04, Rev. 5, RHR Flow Indicators, dated March 14, 1994.
ECP 1-2-I3-05, Rev. 2, RHR Heat Exchanger Inlet Resistance Temperature Detectors, dated March 25, 1994.
ECP 1-I9-01, Rev. 6, RWST Temperature Alarm, Indication, and Heat Tracing, dated September 4, 1994.
ECP 1-2-I9-03, Rev. 16, RWST Levels/ RHR Pumps Interlock, dated October 24, 1997.
ECP 1-CG-39, Rev. 2, RWST Level, dated October 24, 1997.
ECP 1-CG-24A, Rev. 0, RHR Pump Discharge Temperature, dated July 29, 1993.
ECP 1-CG-24C, Rev. 0, RHR Heat Exchanger Discharge Flow, dated July 29, 1993.
ECP 1-2-CG-24D, Rev. 0, RHR Loop Return Flow, dated May 12, 1994.
ECP 1-UNC-100, Rev. 0, RHR Pump PP-35E and PP-35W Discharge Pressure, dated September 30, 1998.
ECP 1-UNC-1,14, Rev. 0, 250 Vdc Battery Voltage, dated September 30, 1998.
ECP 1-RPC-14, Rev. 2, Containment Sump Level, dated May 17, 1994.
ECP 1-2-N3-01, Rev. 4, Redundant Containment Water and Sump Water Level, dated March 12, 1994.
ECP 1-2-UNC-000, Rev. 2, Uncertainty Calculations Reference Document, dated September 4, 1998.
PS-FUSE-001, Rev. 0, AC Rated Fuses in DC Applications, dated September 24, 1994.
PS-250VL-016, Rev. 0, Composite Capacity and Voltage Drop Study, Battery 1AB, dated August 24, 1992.
PS-250VL-017, Rev. 1, Composite Capacity and Voltage Drop Study, Battery 1CD, dated April 12, 1997.
PS-250VL-018, Rev. 0, Composite Capacity and Voltage Drop Study, Battery 2CD, dated August 24, 1992.
PS-250VL-017, Rev. 0, Composite Capacity and Voltage Drop Study, Battery 2AB, dated August 24, 1992.
PS-250VL-006, Rev. 1, Unit 1 250 Volt DC System Train B Station Blackout Load Study-1AB.
PS-250VL-007, Rev. 2, Unit 1 250 Volt DC System Train A Station Blackout Load Study-1CD.
PS-250VL-008, Rev. 1, Unit 2 250 Volt DC System Train A Station Blackout Load Study-2CD.
PS-250VL-009, Rev. 1, Unit 2 250 Volt DC System Train B Station Blackout Load Study-2AB.
PS-EDGL-001, EDG 1AB Steady-State Loading and Voltage Drop.
PS-EDGL-002, EDG 1CD Steady-State Loading and Voltage Drop.
PS-EDGL-003, EDG 2AB Steady-State Loading and Voltage Drop.
PS-EDGL-004, EDG 2CD Steady-State Loading and Voltage Drop.
PS-EDGD-002, Rev. 2, Computer Simulation Study to Evaluate Dynamic Performance of Cook Emergency Diesel Generator System.
PS-4KVD-003, Minimum and Maximum Bus Faults - 4kV [kilovolt],600 Vac, and 480 Vac Systems.
PS-4KVD-002, Fault Short Circuit Calculation.
PS-600VD-001, 600 Vac Fault Calculation.
PS-EPCS-001, Electric Protection Coordination Study.
PSPKVD-001, Cook Voltage Performance Study, 1996-2000 Period.
HV-12SG19-N, Pressure Drop Calculation - Switchgear Rooms.
HV-12SG21-N, 4kV Switchgear Room Ventilation.
C-7, Emergency Diesel Generator Fuel Consumption Rate.
DC-D-02-RH-02, Piping and Pipe Support Analysis, dated May 28, 1997.
ESRR Re orts and Assessments Initial Expanded System Readiness Review Report, Emergency Core Cooling System, dated June 17, 1999.
Initial Expanded System Readiness Review Report, Residual Heat Removal Shutdown Cooling System, dated June 19, 1999.
Initial Expanded System Readiness Review Report, 250 Vdc System, dated June 21, 1999.
ESRR Assessment Input Form DSTR016, Attribute ECCS03, Criterion 6: Motor-Operated Valve Interlocks - ECCS Motor-Operated Valve Generic Letter 96-01 Applicability, dated May 21, 1999.
ESRR Assessment Input Form LLIN004, Attribute ECCS04, Criterion 2: Minimum/Maximum Required Flow - RHR Pump Runout, dated May 21, 1999.
ESRR.Assessment Input Form LLIN008, Attribute ECCS04, Criterion 3: Generic Letter 88-04 RHR Strong/Weak Pump Interactions, dated March 29, 1999.
ESRR Assessment Input Form LLIN023, Attribute ECCS06, Criterion 2: When to Terminate Charging Pumps During Recirculation, dated May 15, 1999.
ESRR Assessment Input Form LLIN038, Attribute'ECCSC10, ECCS Single/Common Mode Failure Vulnerabilities, dated May 17, 1999.
ESRR Assessment Input Form LLIN039, Attribute ECCS08, Criterion 1: Waterhammer - RHR Heat Exchanger Small Break LOCA - Injection to Recirculation Swapover, dated May 24, 1999.
ESRR Assessment Input Form LLI5040, Attribute ECCS08, Criterion 1: Waterhammer - RHR Heat Exchanger Large Break LOCA - Injection to Recirculation Swapover, dated April20, 1999.
ESRR Assessment Input Form BHOL024, Attribute RHRSDC09, Design Basis Document/Final Safety Analysis Report Change Documents Reviewed, dated May 7, 1999.
ESRR Assessment Input Form BHOL043, Attribute RHRSDC09, Instrumentation and Controls Calculations, Specifications, and Vendor Information, dated May 6, 1999.
ESRR Assessment Input Form BHOL059, Attribute RHRSDC04, Calculation Review, dated April 26, 1999.
ESRR Assessment Input Form LLIN007, Attribute ECCS04, Minimum/Maximum Flow Technical Data Book Errors.
Condition Re orts CR P-93-00655 CR P-94-01221 CR P-94-01729 CR P-95-00358 CR P-95-01015 CR P-96-01321 CR P-96-01576 CR P-96-01753 CR P-96-01989 CR P-97-00232 CR P-97-00765 CR P-97-00925 CR P-97-01098 CR P-97-01313 CR P-97-02122 CR P-97-02157 CR P-97-02158 CR P-97-02413 CR P-97-02449 CR P-97-02450 CR P-97-02582 CR P-97-02872 CR P-97-03246 CR P-97-03283 CR P-98-00113 CR P-98-00670 CR P-98-00715 CR P-98-01285 CR P-98-01319 CR P-98-01398 CR P-98-01501 CR P-98-01654 CR P-98-01770 CR P-98-01776 CR P-98-01782 CR P-98-02078 CR P-98-02099 CR P-98-02516 CR P-98-02746 CR P-98-02925 CR P-98-03254 CR P-98-04080 CR P-98-04458 CR P-98-04615 CR P-98-04743 CR P-98-0504T CR P-98-06037 CR P-98-06492 CR P-98-06496 CR P-98-06703 CR P-98-07147 CR P-98-07576 CR P-98-0757?
CR P-98-07757 CR P-98-07812 CR P-98-08155 CR P-98-08327 CR P-S8-08485 CR P-99-00352 CR P-99-00561 CR P-99-00660 CR P-99-00706 CR P-99-00T41 CR P-99-00792 CR P-99-00973 CR P-99-00995 CR P-99-01440 CR P-99-01711 CR P-99-02455 CR P-99-02466 CR P-99-02868
, CR P-99-02944 CR P-99-03178 CR P-99-0327T CR P-99-03307 CR P-99-03329 CR P-99-03382 CR P-99-03457 CR P-99-03524 CR P-99-03595 CR P-99-03638 CR P-99-03783 CR P-99-03859 CR P-99-038T3 CR P-99-04010 CR P-99-04056 CR P-99-04063 CR P-99-04240 CR P-99-04721 CR P-99-04744 CR P-99-04777 CR P-99-04783 CR P-99-04903 CR P-99-04944 CR P-99-05023 CR P-99-05390 CR P-99-05540 CR P-99-05637 CR P-99-05645 CR P-99-05677 CR P-99-05708 CR P-99-05843 CR P-99-06021 CR P-99-06226 CR P-99-06452 CR P-99-06828 CR P-99-06884 CR P-99-07194 CR P-99-07204 CR P-99-07396 CR P-99-07431 GR P-99-07569 CR P-99-07602 CR P-99-08134 CR P-98-08285 CR P-99-08296 CR P-99-08311 CR P-99-08330 CR P-99-08632 CR P-99-08748 CR P-99-08784 CR P-99-09313 CR P-99-09447 CR P-99-09517 CR P-99-10454 CR P-99-10771 CR P-99-10796 CR P-99-11329 CR P-99-12493 CR P-99-12960 CR P-99-15072 CR P-99-15262 CR P-99-15977 CR P-99-16179 CR P-99-16511 CR P-99-16668 CR P-99-16829 CR P-99-18141 CR P-99-18323 CR P-99-18433 CR P-99-18476 CR P-99-18607 CR P-99-18718 CR P-99-18746 CR P-99-19156 CR P-99-19166 CR P-99-19264 CR P-99-19540 CR P-99-19583 CR P-99-19607 CR P-99-19644 CR P-99-19650 CR P-99-19705 CR P-99-19746 CR P-9S-19773 CR P-99-19786 CR P-99-19833 CR P-99-19850 CR P-99-19890 CR P-99-1S8S6 CR P-99-19936 CR P-99-19945 CR P-99-20014 CR P-99-20026 CR P-99-20045 CR P-99-20046
~Drawin a
OP-1-12002, Main AuxiliaryOne Line Diagram - Bus C/D, Engineered Safety System (Train A),
Rev. 36.
OP-1-12001, Main AuxiliaryOne Line Diagram - Bus NB, Engineered Safety System (Train B),
Rev. 43.
OP-2-12002, Main AuxiliaryOne Line Diagram - Bus C/D, Engineered Safety System (Train A),
Rev. 18.
OP-2-12001, Main AuxiliaryOne Line Diagram - Bus NB, Engineered Safety System (Train B),
Rev. 23.
OP-1-12032, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11C, 11D, Engineered Safety System (Train A), Rev. 16.
OP-1-12031, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11C, 11D, Engineered Safety System (Train A), Rev. 23.
OP-1-12033, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11C, 11D, Engineered Safety System (Train A), Rev. 22.
OP-1-12011, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11A, 11B, Engineered Safety System (Train B), Rev. 16.
OP-1-12013, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11A, 11B, Engineered Safety System (Train B), Rev. 16.
OP-1-12012, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 11A, 11B, Engineered Safety System (Train B), Rev. 11.
OP-2-12032, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 21C, 21D, Engineered Safety System (Train A), Rev. 10.
OP-2-12031, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 21C, 11D, Engineered Safety System (Train A), Rev. 10 and Rev. 11.
OP-2-12012, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 21 A, 21 B, Engineered Safety System (Train B), Rev. 10.
OP-2-12011, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 21A, 21B, Engineered Safety System (Train B), Rev. 11.
OP-2-12013, MCC AuxiliaryOne Line Diagram, 600 Vac Bus 21A, 21B, Engineered Safety System (Train B), Rev. 13.
OP-1-98284, Emergency Core Cooling (RHR), Sheet 1, Elementary Diagram, Rev. 39.
OP-1-982841, Emergency Core Cooling (RHR), Sheet 2, Elementary Diagram, Rev. 12.
OP-1-98287, Emergency Core Cooling (RHR), Sheet 2, Elementary Diagram, Rev. 12.
OP-1-982851, Containment Spray System, Sheet 2, Elementary Diagram, Rev. 5 and Rev. 22.
OP-1-98273, Chemical Volume and Control System, Reactor Coolant Charging, Elementary Diagram, Rev. 35.
OP-1-98281, Emergency Core Cooling (Safety Injection), Sheet 1, Elementary Diagram, Rev. 35.
OP-2-98741, AuxiliaryBuilding Ventilation, Sheet 1, Elementary Diagram, Rev. 22 and Rev. 26.
OP-1-98046, 4kV/600 Vac Transformers 11B and 11D, Elementary Diagram, Sheet 1, Rev. 17.
OP-1-98046, 4kV/600 Vac Transformers 11B and 11D, Elementary Diagram, Sheet 2, Rev. 1.
OP-1-98724, Freeze Protection, Condensate and Refueling Water Storage Tanks, Elementary Diagram, Rev. 15.
OP-1-98730, Boric Acid Heat Tracing System Train A, RWST, Elementary Diagram, Rev. 11.
OP-1-98740, Boric Acid Heat Tracing System Train A, RWST, Elementary Diagram, Rev. 11.
OP-1-5128A-40, Flow Diagram - Reactor Coolant Unit 1, Sheet 2, Revision 40.
OP-1-5142-34, Flow Diagram - Emergency Core Cooling (Safety Injection), Revision 34.
OP-1-5144-33, Flow Diagram - Containment Spray Unit 1, Revision 33.
OP-1-5143-53, Flow Diagram - Emergency Core Cooling (RHR) Unit 1, Revision 53.
OP-2-5143-45, Flow Diagram - Emergency Core Cooling (RHR) Unit 2, Revision 45.
OP-1-5143-53, Flow Diagram - Emergency Core Cooling (RHR) Unit 1, dated April26, 1999.
OP-1-5144-33, Flow Diagram - Containment Spray Unit 1, dated April 6, 1999.
OP-1-98284-39, Elementary Diagram Sheet 1 - Residual Heat Removal, dated Api'il 12, 1994.
OP-1-982841-9, Elementary Diagram Sheet 2 - Residual Heat Removal, dated June 14, 1996.
OP-2-982851-1, Elementary Diagram Sheet 1 - Residual Heat Removal.
OP-2 982841-9, Elementary Diagram Sheet 2 - Residual Heat Removal, dated April 26, 1994.
OP-1-12003-22, 250 Vdc Main One Line Diagram Engineered Safety System (Train A/B/N).
OP-1-98055-18, 250 Vdc Battery AB Distribution Schematic Diagram, Sheet 1.
OP-2-98055-17, 250 Vdc Battery AB Distribution Schematic Diagram, Sheet 1.
OP-1-98057-13, 250 Vdc Battery CD Distribution Schematic Diagram, Sheet 1.
OP-2-98057-13, 250 Vdc Battery CD Distribution Schematic Diagram, Sheet 1.
2-RH-14, RHR Pump Discharge Piping - AuxiliaryBuilding, Sheet 1, Rev. 14.
2-RH-15, RHR Pump Discharge Piping - AuxiliaryBuilding, Sheet 1, Rev. 13.
2-RH-16, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 16.
2-RH-17, RHR Pump Discharge Piping - AuxiliaryBuilding, Sheet 1, Rev. 12.
2-RH-18, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 13.
2-RH-19, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 13.
2-RH-20, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 17.
2-RH-25, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 12.
2-RH-26, RHR Pump Discharge Piping - AuxiliaryBuilding, Rev. 12.
2-SI-7, Residual Heat - AuxiliaryBuilding, Sheet 1, Rev. 2.
2-SI-8, Residual Heat - AuxiliaryBuilding, Rev. 23.
2-SI-9, Residual Heat - AuxiliaryBuilding, Rev. 19.
U dated Final Safet Anal sis Re ort UFSAR Section 5.4, Containment Isolation Systems UFSAR Section 6.0, Engineered Safety Features UFSAR Section 6.2, Emergency Core Cooling System UFSAR Section 6.1.1, Engineered Safety Features Criteria UFSAR Section 7.8, Post-Accident Monitoring System UFSAR Section 8.3.4, 250 Vdc System UFSAR Section 8.3.5, 250 Vdc N-Train Battery System UFSAR Section 8.5, Design Evaluation UFSAR Section 8.6, Tests and Inspection UFSAR Section 9.3, Residual Heat Removal System UFSAR Section 14.3.1, Large Break LOCA Analysis Miscellaneous Documents D.C. Cook Restart Plan - Revision 5.
D.C. Cook Units 1 and 2 Technical Specifications.
Valve Inservice Test Program - Third Ten Year Interval, Rev. 1.
DB-12-ECCS, Rev. 0, Design Basis Document for the Emergency Core Cooling System.
DB-12-RHRS, Rev. 0, Design Basis Document for the Residual Heat Removal System.
DB-12-250Vdc, Rev. 0, Design Basis Document for the 250 Vdc System.
Licensing Basis Review Project, Residual Heat Removal System, dated May 21, 1999.
Licensing Basis Review Project, 250 Vdc System, dated May 5, 1999.
NEMAPublication WC 5-1968, Thermoplastic-Insulated Wire and Cable for the Transmission and Distribution of Electrical Energy.
NEMAPublication WC 3-1969, Rubber-Insulated Wire and Cable for the Transmission and Distribution of Electrical Energy.
Engineering Guide EG-IC-004, Rev. 3, Instrument Setpoint/Uncertainty.
Engineering Guide EG-IC-005, Rev. 0, Instrument Uncertainty Procedure Review.
Engineering Guide EG-IC-008, Rev. 0, Instrument Out-of-Tolerance Evaluation.
227340-STG-6300-01, Rev. 3, Engineering Control Packages.
227340-STG-6300-02, Rev. 2, Control of Instrumentation and Control Information.
227340-STG-6300-03, Rev. 2, Control of Instrument Configuration Documents.
800000 LTG-6300-01, Rev. 2, Instrumentation and Controls Information Program.
IEEE 379-1994, Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems.
IEEE 308-1971, Criteria for Class IE Electric Systems for Nuclear Power Generating Stations.
ANSI N195-1976, Fuel Oil Systems for Standby Diesel Generators.
ANSI/ANS 59.51-1989, Fuel Oil Systems for Emergency Diesel Generators.
Regulatory Guide 1.137, Fuel-Oil Systems for Standby Diesel Generators, dated January 1978.
AEP Nuclear Generation Group Directive 800000-DIR-2300-04, Rev. 0, Application and Use of Design Bases, Single Failure Criteria, Engineering Design Bases, And Current Licensing Basis.
Ingersoll-Rand Certified RHR Pump Curves, N-315, -316, -317, and -318, dated June 18, 1971.
AEP Letter AEP:NRC:1246, Response to Generic Letter 96-01, Testing of Safety-Related Logic Circuits, dated April 17, 1996.
AEP Letter AEP:NRC:1246A, Response to Generic Letter 96-01, Testing of Safety-Related Logic Circuits, dated October 28, 1997.
Engineering Action Plan Progress Log Number ECCS-99-134, ECCS Documentation, dated May 25, 1999.
ES-PIPE-1002-QCN, Rev. 0, Operability Screening Guideline for Pipe Support Conditions and Discrepancies Found by ln-Service Inspections, dated November 25, 1998.
Westinghouse Electric Corporation Letter AEP-97-073, Approval of Unit 2 15 Percent RHR and High Head Safety Injection Pump Degradation, SECL-97-050, dated April22, 1997, including Westinghouse Safety Evaluation Checklist 97-050, Rev. 0.
Letter AEP:NRC:1274, Technical Specification Amendment - Safety Injection Pump Runout Flow Limits, dated September 14, 1998.
Letter AEP:NRC:1238C1
~ NRC Inspection Report 50-315/316-96006(DRP),
dated November 15, 1996.
Westinghouse Electric Corporation Letter AEP-97-004, Large Break LOCA Evaluation for 3-Minute Safety Injection Interruption, dated January 17, 1997.
Westinghouse Document E-677125, AuxiliaryPumps, Rev. 1.
Seismic Qualification Screening Evaluation Worksheet, 1-MCAB, Rev. 0, AuxiliaryBuilding Battery Equipment Area.
Westinghouse Electric Corporation Letter ET-NRC-92-3699, Results of Technical Evaluation of Containment Initial Temperature Assumptions for Large Break LOCAAnalysis, dated June 1, 1992.
Unit 1 Preoperational Test Data 1-PO-050-513A, Safety Injection System, dated September 17, 1973.
Unit 2 Preoperational Test Data 2-PO-050-513A, Safety Injection System, dated May 20, 1977.
RST-1999-005-NED, Independent Assessment of Instrument Uncertainty Program, dated June 30, 1999.
DC Distribution System Technical Notebook.
Impel Corporation Electrical Protection Coordination Study, dated August, 1988.
Short Circuit Test Results of AC-Rated RK5 Fuses Applied in Direct Current Systems, Ronald J.
Roman, AEP.
AEP:NRC:9182, Analysis of DCCNP 250 Vdc System Compared to UL198L Test Results, IE Notice 84-65, dated February 5, 1985.
Fuse Protection of DC Systems, Cynthia Cline, Gould Shawmut.
AEP DIT-B-00051-00, Operability Evaluation Input for Battery Chargers Y~ BC-AB1, /~ BC-AB2,
/~ BC-CD1, /~ BC-CD2, dated July 9, 1999.
C8D Technologies, Inc., Battery Charger Final Test Report.
NRC Information Notice 84-65, Underrated Fuses Which May Adversely Affect the Operation of Essential Electrical Equipment, dated August 16, 1984.
Operability Evaluation 91-18-ODE-060, CR P-99-2455 and CR P-99-2466.
Operability Evaluation 91-18-ODE-355, Supplement 1 to ODE-060, CR P-99-2455 and CR P-99-2466.
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