IR 05000456/1993003
| ML20034H289 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/08/1993 |
| From: | Farber M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20034H288 | List: |
| References | |
| 50-456-93-03, 50-456-93-3, 50-457-93-03, 50-457-93-3, NUDOCS 9303160331 | |
| Download: ML20034H289 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION.
REGION III
i l Reports No. 50-456/93003(DRP); 50-457/93003(DRP) . Docket Hos. 50-456; 50-457 Licenses No. NPF-72; NPF-77 j i
Licensee: Commonwealth Edison Company Opus West Ill + 1400 Opus Place , Downers Grove, IL 60515 j ' facility Name: Braidwood Station, Units 1 and 2 f Inspection At: Braidwood Site, Braidwood, Illinois Inspection Conductedi January 20 through March 1, 1993 Inspectors: S. G. Du Pont
J. R. Roton dkk3 Approved By: M. J. F rbe', Chief . Reacto/ProjectsSectionlA Datt/ / [ Inspection-Summary inspection from Januarv 20 throuah March I. 1993 (Reports No. 50-456/93003(DRP): 50-457/93003(DRP)) Areas Inspected: Routine, unannounced safety inspection by the. resident inspectors of licensee action on'previously identified items; licensee event'
report review; operational safety verification; monthly maintenance observation; monthly surveillance observation; engineering and technical
i support; report review and meetings, Results: No violations were identified.
' The licensee initiated Molar Ratio Chemistry Control on Unit I steam
generators in response to previously identified tube cracking concerns.
, The licensee entered Technical Specification 3.0.3 after discovery that a the Phase "B" Containment Isolation Automatic Actuation Logic had not been tested.
I Engineering and Technical staffs were responsive to non-conservatism
discovered. in the Westinghouse analysis of.the Low Temperature Overpressure Protection / Cold Overpressure Mitigation System setpoint.
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DETAILS 1.
Persons Contacted Commonwealth Edison Company (Ceco) S. Berg, Vice President
- R. Flessner, Executive Assistant
- K. L. Kofron, Station Manager R. Stols, Services Director A. Haeger, Regulatory Assurance Supervisor G. R. Masters, Engineering and Construction Manager
- D. E. Cooper, Operations Manager
- G. E. Groth, Maintenance Superintendent R. Byers, Assistant Superintendent Work Planning D. Miller, Technical Services Superintendent G. Plim1, Quality Verification Manager
- G. Vanderheyden, System Engineering Supervisor S. Roth, Security Supervisor
- K. G. Bartes, Quality Verification Superintendent
- J. Lewand, Regulatory Assurance
- Denotes those attending the exit interview conducted on March 2, 1993.
The inspectors also interviewed several other licensee employees.
2.
Licensee Action on Previous 1v Identified items (92701. 92702) a.
Violations (Closed) 456/90012-01: 457/90015-02: Personnel Errors Associated with the Temporary Lift Program. The inspectors reviewed the licensee's response to the Notice of Violation which cited one Severity Level IV violation. The inspectors found the response to be thorough with appropriate corrective actions taken to preclude recurrence. This violation is closed.
(Closed) 456/92004-01. 457/91024-01 and 456/92007-01: 457/92007-01: Violations Pertaining to Various Personnel Errors.
The licensee initiated several corrective actions as documented in Inspection Reports 456/92011; 457/92011 and 456/92017; 457/92017.
The effectiveness of these corrective actions resulted in the elimination of similar personnel errors during the last three quarters. These violations are considered closed based on the effectiveness of the licensee's corrective actions.
b.
Inspection Followup Item (Closed) 456/92007-02: 457/92007-02: The Containment Chilled Water Isolation Valve Closed During Maintenance. The licensee's review of the event determined that the unaccessible location of
. L ' the terminal points prevented adequate verification prior to
installing the jumpers. The licensee's corrective actions ' included manufacturing of a jumper arrangement with installed switches to allow verification of proper jumper landing prior to closing of the circuit. These actions eliminated recurrence in the last three quarters.
This item is closed.
No violations or deviations were identified.
' 3.
Licensee Event Report (LER) Review (92700) LERs were reviewed and closed based on the following criteria: Reportability requirements were met.
- Immediate corrective actions were accomplished.
- Corrective actions to prevent recurrence have been or will be
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initiated per technical specifications.
(Closed) 456/93001: Reactor Trip due to Logic Card failure in Train B Solid State Protection System (SSPS). On January 7, 1993, Unit 1 experienced a reactor trip during the performance of reactor coolant pump underfrequency surveillance testing. The reactor trip was due to the failure of a logic card in the two out of four logic circuit controlling actuation of the trip.
The failed logic card was replaced and the surveillance was completed successfully. The affected logic card was returned to Westinghouse
(vendor) for a failure analysis.
LERs 456/90039 and 456/92013 documented previous failures of logic cards in SSPS.
Because of the previous failures, the quality of these logic cards is considered to be an inspection followup item (50-456/93003-Ol(DRP); 50-457/93003-01(DRP)). The failure analysis of the three documented logic card failures will be evaluated during subsequent inspections to determine the existence of any generic concerns.
No violations or deviations were identified.
4.
Operational Safety Verification (71707) The inspectors verified that the facility was being operated in conformance with the licenses and regulatory requirements and that the licensee's management control system was effectively carrying out its responsibilities for safe operation.
The following activities were observed, evaluated, or reviewed: Unit 1 Steam Generator Molar Ratio Chemistry Control.
Following
the January 7, 1993, reactor trip, the licensee entered a' forced
3 ,
_ . outage to repair a pressurizer isolation drain valve. As part of the outage, the licensee performed a hideout return study to , evaluate the effectiveness of their Molar Ratio Chemistry Control I-for Unit I steam generators.
Initial results showed steam generator crevice chloride concentrations to be higher than anticipated. The equivalence ratio was approximately 1.2; a ratio of 1.0 is the goal. As a result, the target operating ratio was revised upward to allow for more sodium, thereby bringing the equivalence ratio towards it's target of 1.0.
Additionally, steam generator blowdown was returned to a normal configuration. The Braidwood Chemistry Department has now become the lead in developing this methodology for chemistry control.
Entry Into Technical Specification (TS) 3.0.3.
On February 23, 1993, the licensee entered TS 3.0.3 for both units after it was determined the Phase "B" Containment Isolation Automatic Actuation Logic had not been tested in accordance with applicable surveillance requirements. At the time, it was discovered the test lead for Phase B" Actuation Logic was landed on the Containment Spray Actuation Logic terminal. Therefore, when the monthly SSPS Actuation Logic Surveillance was conducted, the Phase "B" Containment Isolation Logic was not tested.
This logic circuit is tested; however, as part of the 18-month time response testing for Engineered Safeguards. Therefore, operability of Phase "B" isolation was docur.rtable. The licensee delayed the Action Requirements for 3.0.3 for 24-hours, in accordance with TS 4.0.3.
The leads (four total) were lifted and re-landed to their proper terminal position (18501-6). The required surveillances were performed satisfactorily and TS 3.0.3 was exited. The licensee's evaluation concluded that the improper wiring had occurred at the factory. During the licensee's discussions with Westinghouse, the licensee was informed that the wiring discrepancy only affected Braidwood and Byron. However, resident inspectors at the McGuire and Catawba Nuclear Power Plants have determined that the discrepancy exists at those facilities. The problem appears to have generic applicability to Westinghouse 4-loop, 7300 series SSPS. The licensee is pursuing a part 21 report from Westinghouse.
No violations or deviations were identified.
5.
Monthly Maintenance Observation (62703) Routinely, station maintenance activities were observed and/or reviewed by the inspectors to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with technical specifications.
The following items were also considered during this review: approvals were obtained prior to initiating the work; functional testing and/or calibrations were performed prior to returning components or systems to
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Y . service; quality control records were maintained; and activities were accomplished by qualified personnel.
The following maintenance activities were observed and reviewed: Nuclear Work Request (NWR) A56128; "ES Drip Leg."
NWR A58627; " Heat Trace Panel."
NWR A55642; "MSR Heating Steam Piping."
NWR A44885; " Door D551."
NWR A57360/A57613; "2A Safety Injection Train Cleansweep."
NWR A58771; "2CWOO20, 2C Circwater Box Inlet Valve Close."
NWR A58747; "1DC06E, Bus 112, +108 VDC Ground AAR."
NWR A58716-93-001; "U1 Turbine Building Crane."
NWR A54586; " Condenser Waterbox Galleries."
NWR A00001-92-Ol; " Fuel Receipt, A2R03."
NWR A54052-002; "0C CW Makeup Pump."
NWR A57005; " Low Pressure Heater 23C Level Control Valve."
NWR A58649; "lFI-AF015A, 1C SG Aux Feedwater, Train A, Flow Indicator Pegged Low."
NWR A5B182; "1VA01SB, IB SX Cubicle Cooler Fan."
No violations or deviations were identified.
6.
Monthly Surveillance Observation (61726) The inspectors observed several of the surveillance testing required by technical specifications during the inspection period and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that results conformed with technical specifications and procedure requirements and were reviewed, and that.
any deficiencies identified during the testing were properly resolved.
The following surveillance activities were observed and reviewed: BwHS 4002-037; "414 and 401 Elevation 8-Hour Battery Operated Emergency Light 18-Month Surveillance."
JBwVS 0.5-3.0C.1-1; "ASME Surveillance Requirements for Component Cooling Pump ICC0lPA."
BwVS 0.5-3.SX.1; "ASME Surveillance for Essential Service Water Pumps."
BwHS 4002-010; "RSH 48V Station Battery Partial 18-Month Surveillance to Clean Battery."
OBw0S FP-M7; " Outdoor Diesel Oil Storage Tank Foam System Valve Position Monthly Surveillance."
B815 4.9.3.1-001; "Ar alog Operational Test of Cold Overpressure Protection PORV 455A."
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, F ' ' ' . , 18w0S 8.1.1.2.A-1, "1A Diesc1 Generator Operability Monthly Section-5."
OBw0S FP-M2; " Wheeled Portable Fire Extinguisher Inspection Monthly Surveillance."
1BwVS 3.2.1-1; " Bus 141 Undervoltage Protection Monthly Surveillance."
1BwVS 3.2.1-2, " Bus 142 Undervoltage Protection Monthly , Surveillance."
1BwVS 3.1.1-7; " Reactor Coolant Pump Undervoltage Monthly Surveillance."
r No violations or deviations were identified.
7.
Enaineerina and Technical Support (37828) The inspectors evaluated the extent to which engineering principles and i evaluations were integrated into daily activities. During this ' inspection period, the inspectors reviewed and observed numerous activities which demonstrated the high degree of involvement and technical expertise of the licensee's engineering staff. These activities included: Operability Determination for Low Temperature Overtemperature = Protection / Cold Overpressure Mitigation System (LTOP/COMS).
, Operability Determination of the Oas Decay Tanks.
.
! Operability Determination for LTOP/COMS.
, Problem: As a result of non-conservatism in the Westinghouse analysis ! of the LTOP/COMS setpoints, past operation within the limits-of the Technical Specification 3.4.9.1 Heatup/Cooldown Curves may have exceeded.
the Appendix G limits. Additionally, under certain conditions, the Cold Overpressure Protection System Power Operated Relief Valves would not-
have satisfactorily mitigated the mass input and heat input transients since the Appendix G limits would have been exceeded.
Resolution: In the operability determination, the licensee identified , three items of concern. These items were.
_ ) a.
Reactor Vessel I b.
Heatup/Cooldown Curves ! J c.
LTOP/COMS System
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' i - . . Their operability justification for each of these items is discussed below: a.
Reactor Vessel: As a result of the non-conservatism in the Westinghouse analysis, I past operation of the reactor vessel may have caused the Appendix G limits to be exceeded by 50 psig maximum (with four , RCPs running) during a design basis transient as identified in , FSAR 5.2.2.11 (Mass and Heat Injection), and/or during heatup/cooldown operations.
In a discussion between the Nuclear Engineering Department and Westinghouse on January 28, 1993, Westinghouse indicated that a recent analysis had determined that past operation of a reactor vessel with an overshoot of the Appendix G limits of at least 100 psig would have had no adverse
effects on the integrity of the reactor vessel. This result was i to be published in the Westinghouse Potential Issue Letter in mid-February 1993.
To date, Westinghouse has not issued this document.
Based on the above, the licensee considered the reactor vessel to , be operable.
b.
Heatup/Cooldown Curves: . By operating the reactor vessel exactly at the heatup or cooldown Technical specification (TS) limit (on the line), the TS limit could be nominally exceeded by 50 psig with four reactor coolant
pumps (RCP) running, or by 7 psig with one RCP running.
Westinghouse will document by analysis that an overshoot of the .
heatup/cooldown curve limits during past operation did not compromise the integrity of the vessel.
c.
LTOP/COMS System:
The LTOP/COMS system protects the reactor vessel from exceeding , Appendix G limits. Transient analysis was performed by , Westinghouse, and described in FSAR 5.2.2.11, to determine the.
, postulated worst case mass input and heat input events and the pressure of these events. A discussion of each transient, and the l effects of these transients as a result of the problem identified
' above, follows.
As stated in Item a. above, Westinghouse will document by analysis that a 100 psig overshoot of the Appendix G values during vessel operation does not compromise the integrity.
, of the reactor vessel.
(1) Mass Input Transient For the worst case mass input event, a pressure.of 550 psig ! at 70 F is indicated in FSAR 5.2.2.11 as being the maximum - pressure that will be reached. After discussions with Westinghouse as a result of the non-conservative analysis,
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. . . this valve has been determined to be suspect. The pressure ' resulting from this transient is a function of the PORV settings. As a result, and assuming single failure analysis criteria, the actual pressure resulting from this transient would be the sum of the highest setpoint PORV plus an overshoot indicated by Westinghouse to be approximately 30 psig. This is also based on a maximum PORV opening time of two seconds which was assumed in the Westinghouse analysis.
At 120 F, the non-conservative pressure differential of 50 psig would not result in the appendix G limit from being exceeded for either unit since the effective transient values would be 620 and 590 psig.
For one RCP running, the margin would be even greater. This margin also increases as temperature increases.
(2) Heat Inout Transient The maximum pressure for this transient would be reached at a RCS temperature of 250 F (SG temperature of 300 F).
Allowing for PORV overshoot and a non-conservatism of 50 psig, the transient pressure would be less than the limiting pressure for this temperature which is 800 psig, i.e. at 800 psig the PORV piping limit is more limiting than the Appendix G limits. This transient therefore has no adverse impact from the non-conservatism in the Westinghouse analysis.
d.
Braidwood Unit 1 Pendina Technical Specification (TS) Chance The current Braidwood Unit 1 TS heatup and cooldown curves have an effective applicability of only 4.5 Effective Full Power Years due to the recent issue of Regulatory Guide 1.99, Revision 2.
Commonwealth Nuclear Engineering Department and Westinghouse prepared revised heatup and cooldown curves for 32 EFPY as well as - new PORV setpoints/ Appendix G limits and submitted these to Braidwood Station for incorporation into the next applicable revision of the TS. Compliance with Regulatory Guide 1.99, Revision 2, will necessitate revising PORV setpoints/ Appendix G limits. Assuming a non-conservative pressure differential of any magnitude, the Appendix G limits would be exceeded at 70 F by that value. At 120of, sufficient margin exists such that a maximum non-conservative value of 50 psig would have no adverse impact.
Sufficient margin exists at 250 F such that the heat input transient has no adverse impact from the non-conservatism in the Westinghouse analysis.
The licensee has established compensatory actions to ensure that the Appendix G limits are not exceeded.
Additionally, the following actions to support the operability assessment remain open:
s
Y - r . Followun Action #1: Review Westinghouse potential issue for ' this concern to assure no additional compensatory actions are required.
Followuo Action #2: Determine if individual Commonwealth Edison actions should be taken for long term resolution of this concern and/or if a Westinghouse Owners Group will be formed with the purpose of addressing a'long term resolution.
Followuo Action #3: Revise FSAR Section 5.2.2.11 to reflect current analysis.
Operability Determination of the Gas Decay Yanks Problem: During routine plant tours, the inspectors noted a number of drums containing contaminated materials (i.e. gloves, hoses, etc.) had been stored inside the Gas Decay Tanks Rooms inside the Auxiliary Building. The Gas Decay Tanks hold radioactive waste gases for decay . ' prior to release to the atmosphere. The inspectors felt the drums represented a hazard to the structural integrity of the Gas Decay Tanks under a seismic event.
The inspectors asked if operability determination of the Gas Decay Tanks in relation to the numerous drums stored in the area had been performed. An operability determination had not been performed, but was subsequently completed. The following details the results of the licensee's determination.
Resolution: Site Engineering and Sargent & Lundy (Architectural Engineers) performed a walkdown of the Gas Decay Tank rooms. All drums located within these rooms are of the same size, 30" high with a ' diameter of 24", and are single stacked over a plastic sheet covering the floor surface in front of the Gas Decay Tanks.
The Gas Decay Tanks are safety-relatad, ASME Class C components. Their function is to prevent the release of radioactive gases to the atmosphere and limit the exposure to the public within allowable limits.
The plant location where the tanks are located is a seismic Category I Area.
Three areas of concern were identified: ! Impact on the integrity of the Gas Decay Tanks under a seismic a.
event.
, b.
Impact on the flood level in the area and, if affected, its impact on the structural wall (Flood Zone S2-3).
i Impact on fire protection loading in the area (Fire Zone 11.2-0).
c.
Seismic Evaluation
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~ i ' ! Analysis performed by Sargent & Lundy indicated the storage of ' ' drums inside the Gas Decay Tank rooms was acceptable with-
adherence to specific criteria involving maximum number of drums (68), storage configuration, and location within the Gas Decay . ' Tank room. This will minimize potential seismic interactions between drums and will maximize their stability.
Flood Analysis A review was performed to determine the impact on the Braidwood i Flood Analysis results for the Gas Decay Tank area. The
additional impact of the 68 drums stored in the room due to their I water displacement has been evaluated in relation to the adequacy of the surrounding structure.
Results indicated that the , ! resulting flood level does not degrade the integrity of the interfacing structures.
l The floor covering under the drums was removed to minimize the f potential for floor drain blockage. The flood level is dependent ! on the number of drums in the area. The Flood Analysis Evaluation. . accounted for the 68 drums currently located in the rooms (the
maximum number allowed in the rooms).
Fire Protection The added drums do not limit the operability of the Fire Protection System in the Gas Decay Tank Rooms, which consists of a f portable fire extinguisher located outside the entrance door to the area. The potential for a fire in the area involving the materials stored inside the drums is limited as each drum is a . [ - sealed container.
All recommendations provided in the operability analysis were included i as compensatory actions. A followup action to verify the analyses has - been completed; the results confirmed the conclusion that the Gas Decay Tanks and the structural components in the Gas Decay Tanks area are not impacted by the 68 drums.
, No violations or deviations were identified.
8.
Report Review During the inspection period, the inspector reviewed the licensee's ' Monthly Performance Report for January 1993. The inspector confirmed that the informatic" provided met the requirements of Technical l Specification 6.9.1.8 and Regulatory Guide 1.16.
l i The inspector also reviewed the licensee's Monthly Plant Status Reports for December 1992 and January 1993.
No violations or deviations were identified.
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l' - . % ' 9.
Inspection Followup Items-Inspection Followup Items are matters which have been discussed with the licensee, which will be reviewed by the inspector and which involve some action on the part of the NRC or licensee or both. An Inspection Followup Item disclosed during the inspection is discussed in Paragraph 3.
10.
Exit Interview (30703) The inspectors met with the licensee representatives denoted in Paragraph I during the inspection period and at the conclusion of the inspection on March 2, 1993. The inspectors summarized the scope and-results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in nature.
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