IR 05000315/1998009
| ML17334B770 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/07/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17334B769 | List: |
| References | |
| 50-315-98-09, 50-315-98-9, 50-316-98-09, 50-316-98-9, NUDOCS 9805190012 | |
| Download: ML17334B770 (36) | |
Text
U.S. NUCLEAR REGULATORYCOMMISSION
REGION III
Docket Nos:
License Nos:
50-315; 50-316 DPR-58; DPR-74 Report Nos:
50-315/98009(DRS); 50-316/98009(DRS)
Licensee:'ndiana Michigan Power Company Facility:
Donald C. Cook Nuclear Generating Plant Location:
1 Cook Place Bridgman, Ml 49106 Dates:
April 14 through 15, 1998.
Inspector:
D. Butler, Reactor Engineer Approved by:
R. N. Gardner, Chief Engineering Specialists Branch 2 Division of Reactor Safety 9805190012 980507 PDR ADOCK 05000315 G
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EXECUTIVESUMMARY D. C. Cook, Units 1 and 2 NRC Inspection Report Nos. 50-315/98009(DRS); 50-316/98009(DRS)
I The purpose of this inspection was to determine the safety significance and regulatory impact of 34 concerns identifie during the 1997 Architectural and Engineering (AE) inspection (50-315/97201; 50-316/97201).
The following observations were made:
D in tr An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to verify or check the adequacy of Engineering Control Procedure (ECP)
calculation Nos. 1-RCP-09 and 2-RCP-09, "RWST Level." SpeciTically, the suction pipe entrance head losses and Bemoulli velocity head losses were not included in the uncertainty analysis.
(Section E8.1)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to verify or check the adequacy of ECP Nos. 1-CG-39 and 2-CG-39,
"Refueling Water Storage Tank Level." Specifically, vortexing (air entrainment) was not addressed when the RWST low-low level setpoint was developed.
(Section E8.3)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identiTied pertaining to the failure to verify or check the adequacy of ECP Nos. 1-2-N3-01, "CNTMTSump Water Level Indication," 1-RPC-14 and 2-RPC-14, "Containment/Containment Sump Level." Specifically, post-accident containment environment effects were not incorporated in the uncertainty analysis.
(Section E8.5)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III, was identified pertaining to the failure to correctly translate containment water inventory requirements into specifications, drawings, procedures, and instructions.
Specifically, it was not demonstrated that sufficient water could be recovered during a design basis accident to prevent pump vortexing. (Section E8.6)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate '/~ inch contairiment sump particulate retention requirements into specifications, drawings, procedures, and instructions.
Specifically, the containment sump screen sections contained /~ inch gaps and the ~/~ inch sump roof vent holes were not covered with screening material.
(Section E8.8)
~
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identifie pertaining to the failure to correctly translate CCW heat exchanger design flow into specifications, drawings, procedures, and instructions.
Specifically, the cooldown analysis and operating procedures used a CCW flow that exceeded the UFSAR design value.
(Section E8.14)
/n'I
An apparent violation of 10 CFR 50, Appendix B, Criterion III, was identified pertaining to the failure to correctly translate RWST Appendix R inventory requirements into specifications, drawings, procedures, and instructions.
Specifically, calculation No.
TH-90-02, "RCS Volume Make-up Required After Appendix R Fire," RWST volume requirements were not incorporated into procedure No. PMP-4100, "Plant Shutdown Safety and Risk Management."
(Section E8.18)
An apparent violation of 10 CFR 50, Appendix B, Criterion III,was identified pertaining to the failure to correctly translate the % inch recirculation sump roof vent hole design into specifications, drawings, procedures, and instructions.
Specifically, the vent holes were plugged without verifying their design basis.
(Section E8.31)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not fullyanalyzing unit operation above UFSAR Tables 6.3-2 and 9.5-3 ESW 76'F ultimate heat sink (lake) design. temperature.
Specifically, the units were operated in 1988 for 22 continuous days at an average lake temperature of 81'F.
(Section E8.28)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not considering the loss of spent fuel pool (SFP) cooling during a design basis accident.
Specifically, the safety evaluations for the Unit 2 dual train CCW/ESW outage did not address the reduction in SFP time-to-boil ifthe Unit 1 CCW flow isolated due to a Unit 1 design basis accident.
(Section E8.29)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for creating a single failure vulnerability in a procedure. revision to ES-1.3,
"Transfer to Cold Leg Recirculation." Specifically, Revision 2 to ES-1.3 piggy-backed all high head injection pumps onto one residual heat removal pump. (Section E8.30)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not performing a safety evaluation for unit operation with CCW temperatures in excess of the 95'F UFSAR Table 9.5-3 design value.
(Section E8.32)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not performing a safety evaluation for unit operation with reactor coolant pump thermal barrier flow less than the 35 gpm UFSAR Table 9.5-2 design value.
(Section E8.33)
An apparent violation of 10 CFR 50.59, "Changes, Tests, and Experiments," was identified for not performing a safety evaluation for residual heat removal operation without automatic overpressure protection as described in UFSAR Section 9.3,
"Residual Heat Removal System."
(Section E8.34)
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An apparent violation of 10 CFR 50, Appendix B, Criterion XVI,was identified pertaining to not promptly correcting an identified condition adverse to quality. Specifically, calculation No. DCCHV12CR11N, "Control Room Temperature Evaluation," identified in 1990 that control room equipment/component life could be reduced to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ifthe ESW temperature reached 87.5'F. Adequate documentation to demonstrate control room equipment shutdown capability at elevated temperatures could not be located.
(Section E8.12)
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E8 Miscellaneous Engineering Issues E8.1 n
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- Refueling water storage tank (RWST) level setpoint error due to flow-induced effects.
The Architectural and Engineering (AE) team noted that the RWST level instruments pressure taps were located on the RWST suction pipe for the emergency core cooling system (ECCS) pumps and containment spray (CTS) pumps.
The maximum flow rate expected during a design basis accident was about 17,800 gpm. At this flow rate, the suction pipe entrance head loss and Bernoulli velocity head loss would cause the indicated RWST tank level to be lower than the actual tank level. Engineering Control Procedure (ECP) Instrument uncertainty calculation No. 1-RCP-09, dated November 1, 1994, "RWST Level," did not include these head losses.
As a consequence, an indicated lower tank level could affect ECCS and CTS pump suction transfers from the RWST to the containment recirculation sump during a design basis accident.
This could lead to a premature transfer to the sump causing ECCS and CTS pump loss due to vortexing (air entrainment) and/or loss of net positive suction head (NPSH) from insufficient sump water level. These errors also affected ECP No. 2-RCP-09.
The failure to verify or check the adequacy of ECP Nos. 1-RCP-09 and 2-RCP-09 is considered an apparent violation (EEI 50-315/98009-01; EEI 50-316/98009-01) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
E8.2
- RWST level instrument loop uncertainties were not accounted for in the Technical Specification (TS) volume verification surveillance.
The AE team was concerned that procedure Nos. 01(02)-OHP 4030.STP.030, "Daily and Shift Surveillance Checks," Revision 25(23), verified the RWST volume to be greater than the TS required 350,000 gallons (89%) without accounting for instrument loop uncertainties.
This item is considered an unresolved item (URI 50-315/98009-02; URI 50-316/98009-02) pending further NRC review.
E8.3 I
- RWST low-low level residual heat removal (RHR) and CTS pump automatic trip did not include all instrument loop uncertainties.
The AE team was concerned that vortexing (air entrainment) could take place when the ECCS and CTS pumps were aligned to the RWST. The low-low level setpoint as described in ECP No. 1-CG-39, dated October 24,1994, "Refueling Water Storage Tank Level," was set at 9 inches above the RWST discharge pipe. However, the ECP did not address the potential for vortexing. The licensee determined that vortexing could occur 12 inches above the discharge pipe. As a consequence, the potential existed to damage the ECCS and CTS pumps due to vortexing. These errors also affected ECP
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No. 2-CG-39. The failure to verify or check the adequacy of ECP Nos. 1-CG-39 and 2-CG-39 is considered an apparent violation (EEI 50-315/98009-03; EEI 50-316/98009-03) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
- Potential for vortexing (air entrainment) was not adequately addressed in a timely manner.
The RWST level biases due to velocity effects were identified during the NRC Systems Based Instrument and Control Inspection (50-315/93012; 50-316/93012).
The licensee determined that from an RWST inventory point of view the bias effects were conservative since the indicated level was lower than the actual level. However, the AE team believed that the overall vortexing effects could have been identified earlier. This is considered an unresolved item (URI 50-315/98009-04; URI 50-316/98009-04)
pending further NRC review.
- Containment sump level instrument loops did not include post-accident uncertainties.
Calculation ECP Nos. 1-2-N3-01, dated March 16, 1994, "Redundant CNTMTWater and CNTMTSump Water Level Indication," 1-RPC-14, dated May 17, 1994,
"Containment/Containment Sump Level," 2-RPC-14, dated May 17, 1994,
"Containment/Containment Sump Level," did not account for the loop uncertainty impact on post-accident containment levels, did not include considerations for RHR and CTS pumps NPSH requirements, and did not account for pump vortexing (air entrainment).
As a consequence, this could impact ECCS and CTS pumps during operator manual transfer from the RWST to the containment sump when implementing emergency operating procedure (EOP) Nos. 01(02)-OHP 4023. ES-1.3, "Transfer to Cold Leg Recirculation." The failure to verify or check the adequacy of ECP Nos. 1-2-N3-01, 1-RPC-14, and 2-RPC-14 is considered an apparent violation (EEI 50-315/98009-05; EEI 50-316/98009-05) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
- Containment recirculation sump design basis water volume requirement was not maintained.
As of September 12, 1997, engineering reviews evaluating design basis accident flow diversions into the inactive portions of the containment sump could not be located.
As a consequence, it was not known ifsufficient water could be recovered during a design basis accident to prevent ECCS and CTS pump vortexing (air entrainment).
This could jeopardize long term pump operation.
The failure to correctly translate containment sump water inventory requirements into speciTications, drawings, procedures, and instructions is considered an apparent violation (EEI 50-315/98009-06; EEI 50-316/98009-06),of 10 CFR 50, Appendix B, Criterion III, "Design Control."
1-; The licensee's definition for a "single active failure" was not consistent with the AE team's interpretatio I
The licensee indicated that their failure modes and effects analyses only had to postulate a "single active failure" as a failure-to-start.
However, the AE team concluded that failure-to-run effects should also be analyzed.
In response, the licensee contacted Westinghouse (5) and were informed that 5 also considered failure-to-run scenarios in their failure analyses.
This is considered an unresolved item (URI 50-31 5/98009-07; URI 50-316/98009-07) pending NRC review of the licensee's failure modes and effects analyses.
E8.8 Iv It
- Recirculation sump screen edge gaps exceeded the containment '/. inch particulate retention requirement.
The recirculation sump particulate retention requirement limited particle sizes to less than N inch to prevent plugging of the% inch containment spray nozzles.
Request for Change (RFC) No. DC-12-2361, completed July 9, 1979, "Modification to the Recirculation Sump," removed the sump horizontal perforated plate in the recirculation sump and installed a fine particulate screen behind the vertical grating at the sump entrance.
However, the fine screen installation was deficient in that the individual screen section edges contained l~ inch gaps.
In addition, ~/. inch containment sump roof vent holes had been installed without screens.
As a consequence, a common mode failure of redundant CTS trains could occur. The failure to correctly translate containment sump particulate retention requirements into specifications, drawings, procedures, and instructions is considered an apparent violation (EEI 50-315/98009-08; EEI 50-316/98009-08) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
E8.9
Following a postulated loss of coolant accident (LOCA), with the ECCS operating in the recirculation mode, the ECCS piping located outside containment could provide an unmonitored and unfiltered leakage path back through the RWST.'he RWST was vented to the atmosphere.
Four of six valves were not leak tested.
This is considered an unresolved item (URI 50-315/98009-09; URI 50-316/98009-09) pending NRC verification that the total back leakage was less than 10 gpm.
E8.10 the plant with one CCW train.
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-1: Capability to cooldown TS 3.0.3 stated, in part, that when a limiting condition for operation was not met, the plant was to be brought to a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The AE team determined that this requirement should be achievable with one CCW train.
In response, the licensee provided the AE team calculation No. NEMP 960519AF, dated June 25, 1996, "CCW LOCA/Cooldown Analysis for the Unit 2 Uprating Program."
However, the calculation used an emergency service water (ESW) temperature of 87.5'F and a CCW temperature of 120'F.
Both temperature values exceeded UFSAR design values, 76'F and 95'F, respectively.
This is considered an unresolved item (URI 50-315/98009-10; URI 50-316/98009-10) pending further NRC revie E8.11 r
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type incorrectly modeled in the cooldown analysis.
- CCW heat exchanger The licensee identified that the CCW heat exchanger was modeled as a counter flow type rather that a single pass heat exchanger.
This is considered an unresolved item (URI 50-315/98009-11; URI 50-316/98009-11) pending NRC verification that sufficient margin exists in the cooldown analysis to accommodate the heat exchanger modeling error.
E8.12 n
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- Identified decreases in control room equipment life due to high temperatures were not promptly corrected.
Calculation No. DCCHV12CR11N, dated June 22, 1990, "Control Room Temperature Evaluation," determined that the minimum control room equipment/component life at an ESW temperature of 87.5'F was 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
However, adequate documentation to demonstrate control room equipment shutdown capability at elevated temperatures could not be located.
As a consequence, the decrease in control room equipment qualified life from 15,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at an ESW temperature of 76'F to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> at 87.5'F could impact plant shutdown with a loss of the nonsafety control room chillers during a design basis accident.
The failure to promptly correct an identified condition adverse to quality is considered an apparent violation (EEI 50-315/98009-12; EEI 50-316/98009-12)
of 10 CFR 50, Appendix B, Criterion XVI,"Corrective Action."
E8.13 n
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- Previously unanalyzed failure modes in the instrument air system.
The AE team evaluated safety related control room air-operated valve (AOV)failure modes.
The valves were supplied from a common nonsafety air header and had been evaluated for a loss of control air. However, the failure analysis did not evaluate the potential for the control air regulators to fail high. This affected the 20 psig, 50 psig, and 85 psig air headers.
This is considered an unresolved item (URI 50-315/98009-13; URI 50-316/98009-13) pending further NRC review.
Subsequently, the licensee modified the nonsafety air header in both units by installing relief valves downstream of each header control air regulator.
E8.14 nr
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- 4: CCWheatexchanger flow in excess of the UFSAR design value.
The UFSAR specified coo!down flow rate was about 8000 gpm (based on UFSAR Table 9.5-3 flow rate of 4.0 x 10'bs/hour).
Calculation No. SAE/FSE-C-AEP.AMP-0088, dated August 20, 1997, "D. C. Cook Units 1 & 2 RHR Cooldown Analysis for a JPO,"
used a CCW flow rate of 4.35 x 10'bs/hour which equates to about 8700 gpm through the shell side of the CCW heat exchanger.
In addition, procedure Nos. 01(02)-OHP 4021.016.003, dated December 3, 1996, "Operation of the Component Cooling Water System During Reactor Startup and Normal Operation," permitted CCW flows up to 9000 gpm. An engineering evaluation had not been performed to demonstrate the
0
acceptability of this condition. As a consequence, the potential existed for the CCW heat exchangers to have increased vibration and/or to structurally fail. The failure to correctly translate the CCW heat exchanger design flow into specifications, drawings, procedures, and instructions is considered an apparent violation (EEI 50-315/98009-14; EEI 50-316/98009-14) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
Since the AE inspection, the licensee performed design change notice (DCN) No. 6944, dated January 28, 1998, "Safety Evaluation for UFSAR Changes to Add Maximum Flow Rates for Component Cooling Water Heat Exchanger and Letdown Heat Exchanger."
The DCN indicated that the CCW heat exchanger could withstand an inlet and outlet bundle velocity up to 8 ft/sec. The-maximum expected bundle velocity was 4.14 ft/sec at 9000 gpm.
In addition, Tubular Exchangers Manufacturers Association Standards recommended that the heat exchanger tubes be supported at least every 52 inches.
The CCW heat exchanger tubes were supported every 30 inches. 'This documentation supports that the CCW heat exchangers could withstand a 9000 gpm flow rate without a significant increase in flow induced vibration.
'8.15 t
- Generic letter (GL)
No. 89-13, "Service Water System Problems Affecting Safety-Related Equipment,"
testing of the CCW/ESW heat exchangers.
The licensee used a maximum fouling factor acceptance criteria of 0.00169 or less.
This value was the maximum allowable fouling acceptable to remove the design heat load. This met the GL intent for an operating cycle, however, the AE team was concerned that this left no margin iffouling were to occur during the operating cycle.
In addition, the AE team determined that the fouling factor acceptance criteria did not include instrument uncertainties.
This is considered an unresolved item (URI 50-31 5/98009-1 5; URI 50-316/98009-1 5) pending further NRC review.
E8.16
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-1: Generic letter No.
89-13, "Service Water System Problems Affecting Safety-Related Equipment," testing of the emergency diesel generator heat exchangers.
Heat exchanger performance trending included the EDG jacket water, lube oil and aftercoolers.
ESW outlet temperatures were recorded and trends were charted over several tests.
The trends indicated that the temperature profiles were relatively constant over the testing period. However, the AE team identified that the heat exchanger outlet temperature was automatically regulated by a flow control valve. Therefore, the trending data only indicated that the flow control valves were operating correctly. This is.
considered an unresolved item (URI 50-315/98009-1 6; URI 50-316/98009-1 6) pending further NRC review.
E8.17 I
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- '1 to demonstrate Unit 2, 250 Vdc, CD battery operability.
- Inadequate justiTication Cell No. 34 was discovered reading less that the TS required 2.13 volts. Temporary modification No. 2-IHP-5021.EMP.009, dated June 19, 1997, installed an individual ceil
%(
charger on cell No. 34. Subsequently, the cell voltage increased to 2.214 volts. The licensee performed an operability determination and concluded that the battery was operable.
However, the cell voltage reading was taken with the individual charger installed. The AE team was concerned that the reading did not represent the true cell state-of-charge.
The cell remained on an individual cell charger for 51 days. The cell was replaced on August 11, 1997. This is considered an unresolved item (URI 50-315/98009-17; URI 50-316/98009-17) pending further NRC review.
E8.18
- The licensee did not always meet RWST, Appendix R, alternate borated water supply requirements.
The fire protection quality assurance program was described in letter No.
AEP:NRC:0847AE, dated August 1, 1997, "Quality Assurance Program Description (QAPD) Proposed Revision." Section 1.7.19.1, Fire Protection QA Program-Introduction, stated, in part, that the Fire Protection QA Program encompasses design, procurement, fabrication, construction, surveillance, inspection, operation, maintenance, modification, and audits.
In addition, Section 1'.7.19.3, "Design Control and Procurement Document Control," stated, in part, that design changes, including field changes and deviations, are controlled by procedures.
As of September 12,1997, calculation No. TH-90-02, dated February 20, 1990, "RCS Volume Make-up Required After Appendix R Fire," determined that the borated water volume from the RWST should be increased from 30,629 to 87,000 gallons.
However, procedure No. PMPQ100, dated February 20, 1996, "Plant Shutdown Safety and Risk Management," was not revised from 30,629 to 87,000 gallons. As a consequence, there were times when the RWST water volume was less than 87,000 gallons in the shutdown Unit RWST which did not meet the Appendix R alternate borated water supply requirement for the operating unit. The failure to correctly translate the RWST Appendix R inventory design basis'into specifications, drawings, procedures, and instructions is considered an apparent violation (EEI 50-315/98009-18; EEI 50-316/98009-18) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
E8.,19 Iv Itm-1 loop uncertainties.
2 -: CCW outlet temperature The licensee could not provide a CCW heat exchanger outlet temperature loop uncertainty calculation. Only the high and low temperature alarm values were accounted for by ECP No. 1-2-C4-02. This is considered an unresolved item (URI 50-315/98009-19; URI 50-316/98009-19) pending further NRC review of the uncertainty measurement process.
E8.20 Iv It temperature loop uncertainties.
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- ESW intake The licensee performed a calculation and identified that the loop uncertainty error was a 3.52'F. This is considered an unresolved item (URI 50-315/98009-20; URI 50-316/98009-20) pending further NRC review of the calculation.
e
E821 I
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-1-1 temperature loop uncertainties.
- Control room The licensee performed a calculation and identified that the loop uncertainty error was t 5.35'F.
This is considered an unresolved item (URI 50-315/98009-21; URI 50-316/98009-21) pending further NRC review of the calculation.
E8.22 program weaknesses.
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2 -: Setpoint control The AE team determined that the RHR and CCW instrumentation and control systems were adequately designed and installed.
However, other setpoint control aspects need to be reviewed for potential impact on other systems.
This is considered an unresolved item (URI 50-315/98009-22; URI 50-316/98009-22) pending further NRC review.
E8.23
- Licensee considered changes to procedures in the conservative direction to be non-intent changes.
The non-intent procedure change process was permitted by TS 6.5.3.1a, "Technical Review and Control." The AE team identified several procedures where process parameters were changed and implemented as non-intent procedure changes without performing a 10 CFR 50.59 screening.
The licensee considered changes to procedure process parameters that were in the conservative direction to be non-intent changes.
Therefore, the change could be implemented with only two signatures prior to receiving a formal review. The formal review was not required for 14 days. This is considered an unresolved item (URI 50-315/98009-23; URI 50-316/98009-23) pending further NRC review of the procedure change process.
E8.24
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72 1-4: Plant drawings and
'design specifications did not conform to American Standard Code for Pressure Piping requirements (ANSI B31.1, 1967 edition).
The AE team identified the following Code deviations:
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CCW piping inside containment conflict with the piping specification and classification requirements.
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CCW system overpressure protection deviated from B31.1 requirements.
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RHR low pressure protection interlock protection was defeated during Mode 4 operation.
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CCW heat exchanger lacked overpressure protection.
This is considered an unresolved item (URI 50-315/98009-24; URI 50-316/98009-24)
pending further NRC review.
E8.25 I s nr Iv it m-1
-2 1-1: Plant equipment abandoned in place without proper reviews and controls.
Licensee policy No.'227000-POL-5400-02, dated July 14, 1995, "Treatment of Abandoned in Place Items," stated, in part, that items abandoned in place should typically be removed as part of the design change process.
However, the AE team identified eight (8) pieces of plant equipment that were abandoned in place without following the policy statement.
This is considered an unresolved item (URI, 50-315/98009-25; URI 50-316/98009-25) pending further NRC review.
E8.26 documentation discrepancies were identified.
The AE team identified the following discrepancies:
1-: Several design
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Design calculations were not revised to account for higher ultimate heat sink (lake) temperatures.
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RHR calculation contained several non-conservative design inputs
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RWST level instrument uncertainty calculation contained an elevation error and a pump flowerror.'
Several drawing errors were identified.
This is considered an unresolved item (URI 50-315/98009-26; URI 50-316/98009-26)
pending NRC review of these discrepancies.
E8.27 flashing due to a procedure deficiency.
- Potential for CCW The AE team was concerned that calculation No. NEMP 960519AF, "CCW LOCA/Cooldown Analysis for U2 Uprating Program," assumption had not been incorporated into operation procedure No. 01(02)-OHP 4021.001.004, "Plant Cooldown from Hot Standby to Cold Shutdown." The potential existed for a water hammer and/or other damaging type transient to occur during cooldown. This is considered an unresolved item (URI 50-315/98009-27; URI 50-316/98009-27) pending further NRC review.
E8.28 I
It stated ultimate heat sink (lake) temperatures.
1-: Exceeding UFSAR The AE team identified as of September 10, 1997, that the units had been operated above the ESW 76'F temperature limitspecified in UFSAR Tables 6.3-2 and 9.5-3. A 10 CFR 50.59 safety evaluation that fullyanalyzed plant operation above 76'F was not provided. Specifically, during July and August of 1988, the units were operated in an unanalyzed condition for 22 continuous days at an average ultimate heat sink
(l tl, II
temperature of 81'F. The failure to fullyanalyze plant operation above 76'F is considered an apparent violation (EEI 50-315/98009-28; EEI 50-316/98009-28) of 10 CFR 50.59, "Changes, Tests, and Experiments," requirements.
E829 I
and ESW outage.
1-
- Unit 2 dual train CCW During the Unit 2 full core off-load outage in 1996 and with Unit 1 at 100% power, both Unit 2 CCW and ESW trains were taken out-of-service on August 7 through 8, 1996, leaving one Unit 1 CCW train available to supply spent fuel pool (SFP) cooling.
Specifically, the March 11 and March 20, 1996, 50.59 safety evaluations performed for the core off-load did not recognize that the Unit 1 CCW system could not perform its safety function under the design basis assumptions described in the UFSAR.
SpeciTically, a single CCW train operating at 95 F could not maintain the SFP bulk water temperature less than specified (160'F) in UFSAR Section 9.4, "Spent Fuel Pool Cooling System."
In addition, with a single Unit 1 CCW train providing SFP cooling, a Unit 1 design basis accident would isolate CCW causing the loss of SFP cooling. As a consequence, the SFP time-to-boil margin would be reduced.
The safety evaluations failed to consider SFP cooling loss during a design basis accident on Unit 1 and the resulting reduction in time-to-boil margin during the Unit 2 dual train CCW and ESW outage.
This is considered an apparent violation (EEI 50-315/98009-29; EEI 50-316/98009-29) of 10 CFR 50.59, "Changes, Tests, and Experiments, " requirements.
E8.30
- Emergency operating procedure Nos. 01(02)-OHP 4023.ES-1.3, Revision 2, "Transfer to Cold Leg Recirculation," procedure change created a single failure vulnerability.
Procedure ES-1.3 was revised (Revision 2) in June 1992 to piggy-back both centrifugal charging and safety injection trains onto the west RHR pump.
However, the 10 CFR 50.59 safety evaluation for this procedure revision did not identity that UFSAR Section 6.2, "Emergency Core Cooling Systems," in use in 1992, required that the operator first transfer one ECCS train to recirculation and then transfer the other ECCS train. As a consequence, a failure of the west RHR pump would cause the loss of all high head emergency core cooling.
In addition, procedure ES-1.3, Revision Nos. 3 and 4, and their corresponding safety evaluations did not identify the single failure vulnerability and that ES-1.3 incorrectly implemented the UFSAR described transfer sequence from injection to recirculation. The safety evaluations were inadequate by failing to identify the single failure vulnerability. This is considered an apparent violation (EEI 50-315/98009-30; EEI 50-316/98009-30) of 10 CFR 50.59, "Changes, Tests, and Experiments," requirements.
E8.31 I
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2 vent hole design basis not understood.
- Recirculation sump roof The NRC resident staff questioned the licensee as to the purpose for the% inch recirculation sump roof vent holes.
The design basis for the vent holes could not be determined since they did not appear on flow diagrams and were not discussed in the
't
UFSAR. The vent holes were subsequently plugged in 1996 for Unit 2 and 1997 for Unit 1. The design control process was not utilized since the licensee believed the repair was returning the sump roof to its original design.
However, the holes were described in letter No. AEP:NRC:00110, dated December 29, 1978, committing to install the vent holes. As a consequence, the design control process was bypassed and the UFSAR
.was not updated with the containment sump roof vent hole design basis.
The failure to correctly translate the% inch containment recirculation sump roof vent hole design basis into specifications, drawings, procedures, and instructions is considered an apparent violation (EEI 50-315/98009-31; EEI 50-316/98009-31) of 10 CFR 50, Appendix B, Criterion III, "Design Control."
E8.32
The AE team identified as of September 10, 1997,,that both units had been operated with CCW temperatures (120'F) above UFSAR Table 9.5-3 specified design value of 95'F. As a consequence, the potential existed for the reactor coolant pump (RCP) seals to fail. A 50.59 safety evaluation had not been performed to review this unanalyzed condition. The failure to perform a safety evaluation for exceeding a UFSAR stated design value is'considered an apparent violation (EEI 50-315/98009-32; EEI 50-316/98009-32) of 10 CFR 50.59, "Changes, Tests, and Experiments," requirements.
E8.33 thermal barrier CCW flow less than the UFSAR value.
- Unit operation with RCP The AE team identified as of September 10, 1997, that both units had been operated with RCP thermal barrier CCW fiows between 25 and 35 gpm. However, UFSAR Table 9.5-2 stated, in part, that the minimum flowwas 35 gpm to each thermal barrier. A 50.59 safety evaluation had not been performed to review this unanalyzed condition, specifically, for flows less than 28 gpm. The failure to perform a safety evaluation for exceeding a UFSAR stated design value is considered an apparent violation (EEI 50-315/98009-33; EEI 50-316/98009-33) of 10 CFR 50.59, "Changes, Tests, and Experiments," requirements.
E8.34 I
r Iv t-1
system without overpressure protection.
- Operation of the RHR The licensee identified on September 11, 1997, that both units had been operated with RHR overpressure protection that did not meet UFSAR requirements.
Specifically, this operating practice did not meet the assumptions identified in UFSAR Section 9.3,
"Residual Heat Removal System." The RHR overpressure interlock associated with RHR hot leg inlet isolation valve Nos. ICM-129 and IMO-128 was defeated without performing a 50.59 safety evaluation for an operating practice that differed from the UFSAR. The purpose of the interlock was to prevent the operators from opening the valves when the reactor coolant system (RCS) pressure was above 400 psig and to provide automatic valve closure when RCS pressure exceeded 600 psig. However, when operating in Mode 4, power was removed from these valves to prevent spurious
fl f
I
closure during shutdown cooling operation.
This defeated the automatic closure feature as described in the UFSAR. The failure to perform a safety evaluation for an operating practice that differed from the UFSAR is considered an apparent violation (EEI 50-315/98009-34; EEI 50-316/98009-34) of 10 CFR 50.59, "Changes, Tests, and Experiments," requirements.
E8.35 TS RWST volume inconsistencies.
UFSAR Section 6.2.2 stated, in part, that the RWST was maintained with a minimum volume of 350,000 gallons of borated water above the bottom of the RWST discharge pipe. TS 3/4.5.5 stated, in part, that the maintained minimum volume of RWST borated water was 350,000 gallons. This is considered an information followup item (IFI 50-315/98009-35; IFI 50-316/98009-35) pending further NRC review.
E8.36 I
It instrumentation and equipment allowed out-of-service times.
- ECCS level The licensee relies on RWST and containment sump level instrumentation during ECCS pump suction transfer from the RWST to the containment recirculation sump.
The out-of-service time for ECCS equipment was defined in TS as 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
However, the out-of-service time for the level instrumentation was defined in TS as 30 days. The AE team was concerned that the instrumentation out-of-service time was not commensurate with the ECCS out-of-service time. This is considered an information followup item (IFI 50-315/98009-36; IFI 50-316/98009-02) pending further NRC review.
X1 Exit Meeting Summary On April 15, 1998, the inspectors presented the inspection results to licensee management.
The licensee acknowledged the findings presented.
The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
I l
iJ,
PARTIALLIST OF PERSONS CONTACTED
~L'QQ K. Baker, Production Engineering A. Blind, Nuclear Engineering Vice President D. Mafer, Plant Engineering J. Kingseed, Nuclear Safety and Analysis T. Postlewait, Design Engineering J. Sampson, Site Vice President J. Gavula, Chief, Engineering Specialists Branch 1, Rill J. Maynen, Cook Resident Inspector IP 92903 INSPECTION PROCEDURE USED Followup - Engineering
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ITEMS OPENED, CLOSED, AND DISCUSSED All identified items apply to Docket Nos. 50-315; 50-316 URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI URI IFI IFI 97201-01 97201-02 97201-03 97201-04 97201-05 97201-06 97201-07 97201-08 97201-09 97201-10 97201-11 97201-12 97201-13 97201-14
'7201-15 97201-16 97201-17 97201-18 97201-19 97201-20 97201-21 97201-22 97201-23 97201-24 97201-25 97201-26 97201-27 97201-28 97201-29 97201-30 97201-31 97201-32 97201-33 97201-34 97201-01 97201-02 EEI 98009-01 URI 98009-02 EEI 98009-03 URI 98009-04 EEI 98009-05 EEI 98009-06 URI 98009-07 EEI 98009-08 URI 98009-09 URI 98009-10 URI 98009-11 EEI 98009-12 URI 98009-13 EEI 98009-14 URI 98009-15 URI 98009-16 URI 98009-17 EEI 98009-18 URI 98009-19 URI 98009-20 URI 98009-21 URI 98009-22 URI 98009-23 URI 98009-24 URI 98009-25 URI 98009-26 URI 98009-27 EEI 98009-28 EEI 98009-29 EEI 98009-30 EEI 98009-31 EEI 98009-32 EEI 98009-33 EEI 98009-34 IFI 98009-35 IFI 98009-36 RWST Level Instrumentation Bias Errors RWST Level Instrument Uncertainties RWST Low-Low Level Setpoint Vortexing Issue Corrective Actions Containment Sump Level Uncertainties Containment Sump Design Basis Single Active Failure Definition Sump '/~ inch Particulate Retention Design Basis Valve Leak Testing One Train CCW Cooldown CCW Heat Exchanger Modeling Error Control Room Temperature Evaluation Instrument AirSystem Failure Modes CCW Heat Exchanger Flow Exceeds UFSAR Value CCW/ESW Heat Exchanger Testing EDG Heat Exchanger Testing 250 Vdc Battery Single Cell Charging Appendix R Alternate Borated Water Supply CCW Outlet Temperature Loop Uncertainties ESW Intake Temperature Loop Uncertainties Control Room Temperature Loop Uncertainties Setpoint Control Program Non-intent Procedure Changes Piping Code Deviations Abandoned Plant Equipment Design Document Discrepancies Potential For CCW Flashing Exceeding Ultimate Heat Sink Temperatures Unit 2 Dual Train CCW/ESW Outage EOP ES-1.3 Single Failure Vulnerability Sump Roof Vent Hole Design Basis Exceeding CCW Design Temperature RCP Thermal Barrier CCW Flow Low RHR Overpressure Protection UFSAR And TS RWST Volume Inconsistencies RWST Level Instrument Out-Of-Service Times
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LIST OF ACRONYMS USED AE AEP ATWS CAL CCP CCW CFR CREVS CR CTS DBA DCN DCP DRS ECCS ECP EDG EEI EOP ESW oF gpm IFI LER LOCA MWt NPSH PMI PMP pslg QA RCP RCS RFC RHR RWST SFP Sl TS UFSAR URI Architectural and Engineering American Electric Power Anticipated Transient Without Scram Confirmatory Action Letter Centrifugal Charging Pump Component Cooling Water Code of Federal Regulations Control Room Emergency Ventilation System Condition Report Containment Spray System Design Basis Accident Design Change Notice Design Change Package Division of Reactor Safety Emergency Core Cooling System Engineering Control Procedure Emergency Diesel Generator Escalated Enforcement Item Emergency Operating Procedure Essential Service Water Degree Fahrenheit gallons per minute Information Follow up Item Licensee Event Report Loss-Of-Coolant-Accident Mega-Watt thermal Net Positive Suction Head Plant Manager Instruction Plant Manager Procedure Pounds Per Square Inch Gauge Quality Assurance Reactor Coolant Pump Reactor Coolant System Request For Change Residual Heat Removal Refueling Water Storage Tank Spent Fuel Pool Safety Injection Technical Specification Updated Final Safety Analysis Report Unresolved Item Westinghouse
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