IR 05000289/1993011
| ML20044F803 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/21/1993 |
| From: | Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20044F788 | List: |
| References | |
| 50-289-93-11, 50-320-93-05, 50-320-93-5, NUDOCS 9306010066 | |
| Download: ML20044F803 (16) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION I
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Report Nos.
93-11 93-05 Docket Nos.
50-289
50-320 t
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License Nos.
DPR-50 DPR-73 Licensee:
GPU Nuclear Corporation
P.O. Box 480
Middletown, PA 17057 r
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Facility:
Three Mile Island Station, Units 1 and 2 Location:
Middletown, Pennsylvania
Inspection Period:
March 28,1993 - May 10,1993
Inspectors:
Francis I. Young, Senior Resident Inspector
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Michele G. Evans, Senior Resident Inspector David P. Beaulieu, Resident Inspector
John P. Segala, Intem Engineer
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b.\\bd 8 auk 3 Approved by:
John F. Rogge,' Chief Q
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Reactor Projects Section No. 4B, DRP t
Inspection Summary: The NRC Staff conducted safety inspections of Unit 1 powe-r operations and Unit 2 cleanup activities. The inspectors reviewed plant operations,
maintenance, radiological controls, security, and engineering and technical support activities
as they related to plant safety.
Results: An overview of inspection results is in the executive summary.
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l 9306010066 930521
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PDR ADDCK 05000289 G
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EXECUTIVE SUMMARY Three Mile Island Nuclear Power Station Report Nos. 50-289/93-11 & 50-320/93-05 Plant Operations Overall, the licensee conducted Unit 1 plant operations in a safe manner. The inspector
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found that shift turnovers were comprehensive and accurate, and adequately reflected plant activities and status. Contml room operators effectively monitored plant operating conditions and made necessary adjustments. The licensee appropriately implemented all of the
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applicable steps of the flood emergency procedure when the water level in the Susquehanna River was above the normal flood stage. The inspector reviewed licensee operability determinations associated with reactor building emergency cooling system testing and
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maintenance and found them to be good. The inspector also reviewed an operability i
determination associated with a decay heat removal system check valve whose leak rate appeared to be above the Technical Specification limit. The inspector found the Plant Review
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Group's assessment and operability determination concerning the current leakage acceptable.
The Unit 2 accident generated water (AGW) evaporator continues to operate and approximately 1,873,000 gallons (81% of total) of AGW had been vaporized to the atmosphere at the close of the inspection period.
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Radiological Controls l
t During each Auxiliary Building tour, the inspector paid particular attention to ensure
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radiological surveys were current and that radiological area postings were consistent with
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recent surveys. The inspector noted no discrepancies and concluded that overall radiological controls were good.
i Safety Assessment and Ouality Verification l
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The inspector attended General Office Review Board (GORB) meetings and found that the
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GORB's review of issues was comprehensive and there was a clear focus on plant safety.
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Engineering and Technical Suonort in January 1992, an inadvertent partial actuation of the emergency feedwater system occurred
due to a miswiring of the heat sink protection system (HSPS). The miswiring occurred as a l
result of an inadequate drawing change. The failure to establish an adequate drawing is
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considered a violation (50-289/93-11-01). Since the licensee's evaluation and corrective actions have been completed and were reviewed and found to be acceptable, no response to
the violation is required.
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The licensee has found additional HSPS drawing / wiring errors that currently exist and has committed to perform a walk down of the HSPS wiring during the next refueling outage to correct the discrepancies. The inspector concluded that continued plant operation until HSPS system configuration control is regained in the next refueling outage is acceptable based on
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j satisfactory functional testing of the HSPS system.
The inspector reviewed whether the licensee's removal of the chlorine detection system (CDS) was acceptable, since this system is still referred to in Technical Specifications and the Final Safety Analysis Report. The inspector found that the licensee had adequately evaluated i
the removal of the CDS.
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l Two small pieces of stainless steel debris were found in the makeup system after the pieces j
interfered with the operation of the pressurizer level control valve. Although the licensee has not identified the source of the debris, the inspector found the licensee's makeup system operability determination to be acceptable. The licensee has taken an aggressive and comprehensive approach in their search.
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Emergency Preoaredness
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Licensee performance at the Emergency Operations Facility during the emergency preparedness annual practice exercise was good.
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SUMMARY OF FACILITY ACTIVITIES
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i 1.1 Licensee Activities i
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Unit I remained at 100% power throughout the inspection period.
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t The Unit 2 Accident Generated Water (AGW) evaporator continued to vaporize AGW to the
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atmosphere and at the close of the inspection period approximately 1,873,000 gallons (81%
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of total amount) had been vaporized.
l 1.2 NRC Staff Activities
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This inspection assessed the adequacy of licensee activities for reactor safety, safeguards, and i
radiation protection. The inspectors made this assessment by reviewing information on a
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sampling basis. The inspectors obtained information through actual observation oflicensee activities, interviews with licensee personnel, and documentation reviews.
l The inspectors observed licensee activities during both normal and backshift hours: 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> t
of direct inspection were conducted on backshift and 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> were conducted on deep i
backshift. The times of backshift inspection were adjusted weekly to assure randomness.
2.0 PLANT OPERATIONS (71707)
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2.1 Operational Safety Verincation
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The inspectors observed overall plant operation and verified that the licensee operated the plant safely and in accordance with procedures and regulatory requirements. The inspectors r
i conducted regular tours of the following plant areas-
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- Control Room
-- Auxiliary Buildmg i
- Switch Gear Areas
- Turbine Building i
- Access Control Points
- Intake Structure
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- Protected Area Fence Lind
- Intermediate Building
- Fuel Handling Building
- Diesel Generator Building
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The inspectors observed plant conditions through control room tours to verify proper
alignment of engineered safety features; to verify that operator response to alarm conditions j
was in accordance with plant operating procedures; to verify compliance with Technical Specifications, including implementation of appropriate action statements for equipment out t
of service, and; to review logs and records to determine if entries were accurate and
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identified equipment status or deficiencies. These records included operating logs, turnover j
sheets, and system safety tags.
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The inspector conducted detailed walkdowns of accessible areas to inspect major components and systems for leakage, proper alignment, proper lubrication, proper cooling water supply, and any general condition that might prevent fulfillment of their safety function. The inspector observed plant housekeeping controls including control and storage of flammable material and other potential safety hazards.
The inspector found that shift turnovers were comprehensive and accurate, and adequately reflected plant activities and status. Control room operators effectively monitored plant operating conditions and made necessary adjustments. Housekeeping was commensurate with ongoing work. The inspector concluded that the licensee conducted overall plant operations in a safe and conservative manner.
2.2 Evaporation of Unit 2 Accident Generated Water The inspectors observed overall evaporator operation and verified that the evaporator was operated in accordance with licensee procedures and regulatory requirements. At the close of the inspection period, 1,873,000 gallocs (81%) of the 2.3 million gal'.ons of AGW had been evaporated.
The inspectors identified no conditions that were adverse to safety or contrary to regulatory requirements.
2.3 IIigh River Water I2 vel During the period of March 30 to April 5,1993, the water level of the Susquehanna River in the vicinity of Three Mile Island was above the normal flood stage level. On March 30, the licensee received a forecast that within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> the volume of water passing the intake structure to Three Mile Island Unit I would exceed 350,000 cubic feet per second (cfs). Based on this forecast, the licensee was required to enter Emergency Prccedure (EP)
1202-32, " Flood," and initiate precautionary measures. TMI has a dike that encompasses vulnerable components including the screen house, cooling towers, and vital areas. The earthen dike has an overlay of rock and is built to withstand a design flood of 1,100,000 cfs.
With a river water flow of greater than 350,000 cfs, the licensee was required to inspect the dike on a shiftly basis to ensure that there was no erosion occurring to the dike due to increased water level and volume of water passing down the river. The licensee did not observe any significant degradation of the-earthen dike. On April 2,1993, the river reached a maximum level of 20.5 feet which corresponds to approximately 415,000 cfs.
The inspector reviewed EP 1202-32 to ensure that all applicable steps were being taken. The inspector independently walked down the earthen dike while the river was above flood stage.
The inspector concluded that no significant erosion of the dike had occurred and that the licensee had appropriately implemented all of the applicable steps of the flood emergency procedur. _ -
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2.4 Reactor Building Emergency Cooling Water System Operability Determinations The inspector evaluated the licensee's operability evaluation associated with maintenance and testing of the reactor building emergency cooling water (RR) system. The RR system consists of two RR pumps that supply a single header and provides river water to three emergency cooling coils in the reactor building. The cooling coils remove heat from the reactor building in the event of a loss of coolant accident. After passing through the cooling coils, the river water flows through a back pressure regulating valve, RR-V-6, and back to the river via the mechanical draft cooling tower. RR-V-6 is a pneumatically operated valve designed to maintain cooling coil pressure at 60 psi, which is above the design reactor building pressure, to prevent leakage out of the containment through a damaged system. RR-V-5, a motor operated valve that bypasses RR-V-6, can be operated from the control room to maintain system back pressure in the event of RR-V-6 failure. On March 17,1993, the licensee performed Special Test Procedure 1-93-004 which fully strokes open RR-V-5, then closes RR-V-5, to verify that the valve motor operator can operate under full flow conditions.
A Shift Supervisor questioned whether the RR system would be rendered inoperable since system back pressure cannot be maintained at 60 psi with RR-V-5 fully open and only one RR pump operating. Technical Specification 3.3 requires that two RR pumps, two cooling coils, and associated engineered safeguards (ES) valves and interlocks be maintained operable.
I The Plant Review Group (PRG) evaluated the Shift Supervisor's operability question and determined that the RR system would not be rendered inoperable with RR-V-5 fully open.
The PRG determined that the System Design Basis Document for the RR system lists the valves under Process Design rather than Licensing Basis Design. The PRG also noted that RR-V-5 and RR-V-6 do not receive an ES signan and therefore the Technical Specification 3.3 requirement for ES valves is not applicable. The PRG concluded that opening RR-V-5 for testing does not render the RR system inoperable per Technical Specification, since the Technical Specification safety function of providing containment cooling does not rely on back pressure control.
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i The inspector reviewed the PRG's operability evaluation and the System Design Basis l
Document for the RR system. The Process Requirement for the 60 psi back pressure is from
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a licensee commitment to the Atomic Energy Commission in Conference Notes #65 dated August 17, 1967. Therefore, the licensee placed the RR system in a non-conforming condition, as defined in Generic Letter 91-18. Generic Letter 91-18 states that the fact that a system is not fully qualified (which includes conforming to licensee commitments) does not, in all cases, render that system unable to perform its specified function if called upon.
Therefore, the inspector agrees with the licensee's operability determination, because the RR system will still be able to perform its design basis function of removing heat from the reactor building in the event of a loss of coolant accident.
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The inspector also questioned the operability of the RR system while the licensee performed maintenance on RR-V-5 in accordance with Corrective Maintenance Procedure (CMP) 1420-LTQ-7, " Dynamic Testing of Motor Operated Valves Using MOVATS ITI Series 3000 Valve Analysis System." CMP 1420-LTQ-7 requires the installation of test equipment such that RR-V-5 is in its mid-position and cannot be operated manually or remotely. The amount of the valve opening on RR-V-6 is limited to 65% so that it will not pass more flow than one pump can provide without exceeding the permissible operating range of the RR pumps (i.e.
pump runout). Pump runout results in pump cavitation and/or excessive vibration. The inspector questioned the licensee if the RR system would be rendered inoperable due to pump runout with RR-V-5 in its mid-position. The licensee performed calculations to demonstrate that even with RR-V-5 fully open, maximum RR system flow would be 7100 gpm. The pump head curve for the RR pumps shows that pump runout with one RR pump in operation is approximately 7400 gpm. Therefore, the maintenance on the RR-V-5 did not affect RR system operability.
The inspector concluded that the Shift Supervisor's questioning attitude concerning RR system operability was excellent and the licensee's operability evaluations associated with RR system maintenance and testing were good.
2.5 Decay Heat Removal System Check Valve Operability Evaluation On April 12, 1993, the inspector observed the performance of Surveillance Procedure (SP)
1303-5.2, " Emergency Loading Sequence and High Pressure Injection Imgic Channel / Component Test." When valve DH-V-4B was opened in accordance with the surveillance, the licensee observed a decrease in core flood tank (CFT) level and pressure.
DH-V-4B is a normally closed containment isolation valve on the discharge of the decay heat removal (DHR) pump (DH-V-1B). Downstream of DH-V-4B is a check valve, DH-V-22B.
Downstream of DH-V-22B, the DHR system discharges between the two check valves that connect the CFT to the reactor vessel. When DH-V-4B was opened, check valve DH-V-22B leaked, causing the CFT (at 600 psig) to flow into the DHR system. Technical Specification 3.1.6.10 has a maximum allowable leakage of 5.0 gpm for DH-V-22B with normal reactor coolant system pressure (2155 psig) on the downstream side. With only CFT pressure (600 psig) on the downstream side, the maximum allowable leakage was calculated to be 2.45 gpm. The design pressure for the DHR system is 505 psig, the hydrostatic test pressure is 637 psig, the relief valve on the DHR pump discharge is set at 520 psig, and DH-V-4B is designed to close at 400 psig. The data recorded during the performance of SP 1303-5.2 was not sufficient to determine the DH-V-22B leak rate. The next day the licensee opened DH-V-4B to record additional data. The data on April 13, 1993, indicated that DH-V-22B leak rate was 2.53 gpm which is above the Technical Specification limit (2.45 gpm).
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1.cak rate testing of DH-V-22B is normally performed in accordance with SP 1300-3T,
" Pressure Isolation Test of CF-V-4A/B, SA/B, and DH-V-22A/B." SP 1300-3T measures DH-V-22B leakage by collecting the leakage in a drum that is connected to a DHR system vent. The Plant Review Group (PRG) determined that CFT level decrease method of determining leak rate is not representative of the conditions which are established during the performance of 1300-3T (the qualified test), and should not be compared quantitatively with such a test. The PRG based this decision on data from SP 1303-5.2 that was performed in September 1991, which showed a DH-V-22B leak rate of 3.96 gpm based on CFT level decrease. SP 1300-3T, the qualified test, was performed shortly after and DH-V-22B leak rate was 0.53 gpm. The licensee has scheduled SP 1300-3T, during the upcoming refueling outage to qualify the DH-V-22B leak rate. The licensee has contingency plans to inspect / repair DH-V-22B based on these test results.
The inspector discussed with the licensee whether the CFT level instrumentation is accurate enough to quantify a leak rate. CFT level is normally indicated to 0.01 feet. Due to the relatively Jarge diameter of the CFT,0.01 feet corresponds to 5.08 gallons. Due to the short duration of the test, a leak rate to within 0.01 gpm cannot be precisely attained using CFT level decrease method. The inspector agrees with the PRG that even though the measurement of CFT level decrease is not a qualified test, it still serves as an indicator of valve operability. The CFT lesel decrease can be used as a gross indicator that the leak rate may exceed Technical Specification requirements. Actions can then be taken to seat the valve, as the licedsee did in August 1991, or perform a qualified leak rate test to quantify DH-V-22B leakage.
The inspector concluded that the PRG's assessment and operability determination concerning the current leakage through DH-V-22B was acceptable.
t 3.0 RADIOLOGICAL CONTROLS (71707)
During entry into and exit from radiologically controlled areas, the inspectors verified that proper warning signs were posted, personnel entering were wearing proper dosimetry, personnel and material leaving were properly monitored for radioactive contamination, and monitoring instruments were functional and in calibration. The inspectors also reviewed extended Radiation Work Permits (RWPs) and survey status boards to verify that they were
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current and accurate. The inspectors observed activities in radiologically controlled areas and verified that personnel were complying with the requirements of applicable RWPs and that workers were aware of the radiological conditions in the area.
During each Auxiliary Building tour, the inspector paid particular attention to ensure radiological surveys were current and that radiological area postings were consistent with recent surveys. The inspector noted no discrepancies and concluded that overall radiological controls were good.
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4.0 MALNTENANCE AND SURVEILLANCE (61726, 62703, 71707)
4.1 Maintenance Observations The inspector reviewed selected maintenance activities to assure that: the activity did not violate Technical Specification Limiting Conditions for Operation and that redundant components were operable; required approvals and releases had been obtained prior to commencing work; procedures used for the task were adequate and work was within the skills
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of the trade; maintenance technicians were properly qualified; radiological and fire prevention controls were adequate; and, equipment was properly tested and returned to service.
The inspector observed the performance of Corrective Maintenance Procedure (CMP) 1420-LTQ-7, " Dynamic Testing of Motor Operated Valves Using MOVATS ITI Series 3000 Valve Analysis System," on RR-V-5. Overall, the inspector found that individuals performing the maintenance were knowledgeable, maintenance procedure quality was good, and proper QA l
documentation existed for replacement parts. The inspector also reviewed maintenance associated with the RR system as discussed in Section 2.4 of this report.
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Four days following the removal of debris from MU-V-17 (see Section 7.3), the licensee noted that the valve was operating sluggishly. The licensee found that the instrument air valve to the MU-V-17 positioner was partially closed. The licensee conducted an investigation, but was unable to confirm how the valve was placed in the partially closed position. As described in Inspection Report 50-289/93-09, there was a recent instance where two instrument air valves associated with the emergency feedwater system were mispositioned
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by Instrument and Controls technicians following maintenance. Although it is unknown if the partial closure of the instrument air valve to the MU-V-17 positioner was also the result of a technician error, the licensee needs to be sensitive in ensuring instrument air valves are properly positioned.
4.2 Surveillance Observations The inspectors reviewed and inspected selected surveillance tests to determine whether
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properly approved procedures were in use, details were adequate, test instrumentation was
properly calibrated and used, Technical Specifications were satisfied, testing was performed l
by qualified personnel and test results satisfied acceptance criteria or were properly dispositioned.
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Surveillance Procedure 1300-3T, " Pressure Isolation Test of CF-V-4A/B, SA/B, and
DH-V-22A/B"
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Surveillance Procedure 1303-5.2, " Emergency Loading Sequence and High Pressure
Injection Logic Channel / Component Test" i
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Surveillance Procedure 1300-3H, " Inservice Test of MU Pumps and Valves"
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Surveillance Procedure 1303-11.8, "High Pressure Injection" t
The reviews of the these surveillances are discussed in detail in Sections 2.5 and 7.3 of this report.
5.0 SECURITY (71707)
The inspectors verified the implementation of the Physical Security Plan by verifying:
Protected Area and Vital Area barriers were well maintained and not compromised; isolation zones were clear; personnel and vehicles entering and packages being delivered to the Protected Area were properly searched and access control was in accordance with approved licensee procedures; persons granted access to the site were badged to indicate whether they
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have unescorted access; security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorized; security posts were adequately staffed and equipped; and, adequate illumination was maintained.
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I On April 28,1993, the inspector walked down the protected area fence and found it was properly maintained. The inspectors also observed the search by security personnel of several
vehicles prior to entry into the protected area. The inspectors concluded that the Security Plan was being properly implemented.
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6.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (40500)
6.1 General Office Review Board Meeting
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On April 21 and 22,1993, the inspector attended General Office Review Board (GORB)
meetings. The inspector observed discussions of plant status and significant events, Independent Onsite Safety Review Group (IOSRG) activities, Licensee Event Reports and a Chemistry Program overview. The inspector found that GORB's review of the issues discussed was comprehensive and there was a clear focus on plant safety.
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7.0 ENGINEERING AND TECHNICAL SUPPORT (37828,40500)
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7.1 (Closed) Unresolved Item (URI,50-289/91-30-01) Heat Sink Protection System Miswiring (VIO, 50-289/93-11-01)
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On January 22,1992, at 1:17 p.m., an inadvertent emergency feedwater (EFW) system partial actuation occurred as a result of a miswiring of the heat sink protection system l
(HSPS). The miswiring occurred during a modification to HSPS, performed during the SR
refueling outage in January through March 1990. HSPS is a two out of four channel system that is applied to two independent trains. Each train controls the respective 'A' or 'B'
portion of the emergency feedwater system. Each channel is independently powered from a
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vital bus. Train 'A' and 'B' logic cabinets are powered from vital bus 'A' and 'B'
respectively, with an auctioneered non-vital power supply that will maintain the logic circuits
energized if the vital power supply is lost. The non-vital power is provided to prevent problems associated with re-energizing a dead train and is not used to maintain system operability. During the 8R refueling outage, additional logic nests were added to both trains of HSPS. Due to the miswiring, the new train 'B' logic nests were powered from the channel II power supply which is not backed up with the non-vital power supply. When the
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train 'B' vital power supply, which also supplies channel II, was deenergized to perform inverter maintenance, the new logic nests did not receive the non-vital power supply and were
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deenergized. This resulted in a partial actuation of EFW. The partial deenergization of HSPS train 'B' resulted in an unusual response of the EFW system. One of the steam admission valves to the steam driven EFW pump, MS-V-13A, and two EFW control valves,
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EF-V-30B/D, opened which initiated flow to the once-through-steam-generators (OTSGs.)
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Plant operators identified the unusual EFW actuation, and after verifying no valid EFW start signal was present, the operators closed EF-V-30B/D to terminate flow to the OTSGs. The
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operators then restored vital power to train 'B' and channel II and restored the EFW system to its normal alignment. Approximately 200 gallons of water from the condensate storage
tanks were injected into each steam generator.
On February 2,1992, the licensee corrected the power supply miswiring. The inspector
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reviewed the licensee's operability determination and the training of operators while the miswiring remained uncorrected. The inspector found the licensee's operability determination and operator training to be acceptable as discussed in Inspection Report 50-289/91-30-01.
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In February and June 1992, the licensee submitted Licensee Event Reports 92-001-00 and 92-001-01, respectively, that listed three root causes for the miswiring: 1) The 8R construction drawings were not clear and could be easily misinterpreted; 2) The Startup and Test group, who is responsible for major plant modifications, was not involved during the initial phase of
construction to provide technical guidance to construction workers, and; 3) There was
inadequate implementation of modification test program requirements which allowed the j
power supply wiring error to go undetected. Functional testing should have been conducted i
to verify power separation between HSPS channels and trains.
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associated with this incident. The inspector found the licensee's description that the 8R drawings were not clear and could be easily misinterpreted to be acceptable but over-simplified. The inspector reviewed Drawing No. 5391-839-005, which shows that the
licensee initially intended to supply power to the new nests by daisy chaining the power supply from an existing nest in the train 'B' cabinet. The licensee then wrote a Field Change
Request No. 070909 to supply power to the new nests by running separate wiring to the j
power supply terminal block in the bottom of the train 'B' cabinet. The Field Change Request did not update the drawing but only provided a word description of the desired change. The word description indicated which power supply terminal block connection points to use but did not indicate which cabinet the power supply terminal block was located. The
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technicians modifying train 'A' correctly ran the power supply from terminal block in the i
train 'A' cabinet but the technicians modifying train 'B' ran the power supply from the terminal block in the channel Il cabinet. Since tiie Field Change Request did not specify which cabinet should supply power, it is unknown how each set of technicians (train 'A' and
'B') chose the power supply. In addition, for each HSPS train, a foreman and QC inspector were assigned full time to oversee the modification. The QC inspector and a Startup and Test person performed a 100% visual point to point wiring verification for the modified wiring. Since the power supply was not specified, it is unknown how the power supply was verified by QC and Startup and Test personnel.
i The inspector reviewed the licensee's procedural controls associated with drawing changes.
Changes to drawing are controlled by EMP-015, " Field Questionnaires, Change Notices, and Change Requests." EMP-015,Section III, states that approved Field Change Requests shall contain sufficient engineering detail so as to not require interpretation. EMP-015 does not specifically prohibit the use of a word description to change a drawing. The inspector questioned the licensee concerning this practice since it may have contributed to this incident.
The licensee indicated that it is not a normal practice to change a drawing using a word description. However, the licensee indicated that the use of a word oescription does not constitute a deficiency in the dM n control process and that they want the flexibility to be
able to use word descriptions iv <.hange drawings when needed. The inspector found this approach to be acceptable.
The inspector evaluated the licensee's post-modification testing of HSPS. The licensee did j
not specifically verify the proper power supply to the new logic modules because the power supply to HSPS train 'B' was not changed by the modification and therefore was considered outside the scope of the modification. If the unanticipated wiring error that placed them outside the scope of the modification did not occur, the testing performed would have been acceptable. Therefore, the inspector determined that the functional testing performed by the licensee was not unreasonable, but should have been broader in scope.
While reviewing this specific miswiring incident, the inspector learned that the licensee has found several other places where the HSPS field wiring does not reflect HSPS drawings.
There are other HSPS power supply wiring / drawing errors in three HSPS cabinets that j
currently exist that the licensee cannot trace out and correct while the plant is at power. In LER 92-001, the licensee mentioned the additional HSPS power supply wiring errors and committed to trace out power supply wiring during the 10R refueling outage. However, the inspector learned of additional HSPS drawing errors that were identified while performing maintenance that are not associated with the power supply These wiring / drawing errors stem from the initial installation of HSPS during the 6R refueling outage. The licensee used three outside contractors for the installation. The licensee and contractors were not effective in ensuring the drawings for all three contractors reflected all of the numerous field modifications. The licensee was aware of the drawing deficiencies and hired a contractor to l
consolidate the drawings from the 6R outage and then modify these updated drawings to perform the SR HSPS modifications. Due to scheduling pressures, the contractor did not l
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complete the drawing upgrades prior to drafting the 8R design modifications. Consequently,
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the current HSPS wiring diagrams cannot be relied upon to determine existing system
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configuration. The licensee has three change notices to HSPS draw' gs that are not m
associated with the HSPS power supply. The inspector reviewed all of the drawing changes associated with the three change notices and agrees with the licensee that the drawing
deficiencies have no affect on system operation. However, the wiring drawings cannot be
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relied upon when performing HSPS maintenance. When the licensee performs HSPS maintenance, they must trace down wiring to the extent possible to ensure that maintenance will not have an unanticipated adverse affect. The inspector discussed his concerns associated with the current HSPS system configuration with the licensee and they committed to expand the scope of HSPS walkdowns beyond that specified in LER 92-001. The licensee committed to perform a walk down of HSPS wiring during the 10R refueling outage and make the necessary changes to ensure HSPS wiring diagrams reflect actual system configuration.
The inspector evaluated whether it was acceptable for the licensee to continue to operate with an unknown HSPS system configuration. The inspector independently verified HSPS surveillance testing results and agrees with the licensee that the HSPS system will function as i
designed if called upon. Therefore, continued plant operation until HSPS system configuration control is regained in the 10R refueling outage is acceptable.
The inspector concluded that LERs 92-001-00 and 01 were acceptable. Plant operator response to the unusual actuation of EFW was excellent. The inspector determined that the -
current HSPS wiring / drawing deficiencies are a significant weakness. The inspector determined that the Field Change Request No. 070909 was inadequate, because the word description of the desired change did not state where the new logic nests were to be powered l
from. This incident is safety significant, because the incorrect installation was verified by
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both QC and Startup and Test personnel. In addition, the miswiring resulted in an inadvertent actuation of the EFW system that unnecessarily challenged plant operators. The failure to establish an adequate drawing for the power supply to the new HSPS train 'B' logic nests is a violation of 10 CFR 50, Appendix B, Section V, which states that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures or drawings. Since the licensee's evaluation and corrective actions associated with the HSPS miswiring have been completed and were reviewed and found acceptable during this inspection, no response to this violation is required. This violation j
(50-289/93-11-01) is closed.
7.2 Chlorine Detection System Removal During the inspection period, the licensee began removing the chlorine detection system (CDS). The inspector evaluated whether the removal of the CDS was acceptable since this system is still referred to in Technical Specifications and Firal Safety Analysis Report (FSAR). The CDS was designed to interface with the control building ventilation system (CBVS) to isolate the control room in the event of a postulated onsite chlorine gas release
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which could occur during a gas delivery or during the operation of the chlorination system.
The licensee is removing the CDS because the circulating water and river water gaseous l
chlorination systems have been replaced with non-chlorine based systems.
l Technical Specification 3.5.6 requires the CDS to be operable only when chlorine containers exceeding 150 lbs. are located onsite. The licensee administratively prohibits chlorine cylinders greater than 150 lbs. from site per Warehouse Instruction File 90-10 and TMI l
Security Directive 13 such that Technical Specification 3.5.6 is never applicable. The only place onsite that uses chlorine is the sewage treatment facility. The inspector verified that
there were no chlorine containers located at the sewage treatment facility that were greater
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than 150 lbs.. Since the size of the chlorine containers are controlled, the licensee
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determined that the Technical Specification Change Request was not a prerequisite for the CDS removal. The licensee plans to submit a Technical Specification Change Request to delete any reference to the CDS.
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The FSAR notes that the CDS will be removed in the 10R refueling outage. The licensee plans to change the FSAR after the CDS has been removed to indicate that there are
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l administrative controls to prohibit the purchase, ordering and the delivery onsite of chlorine containers greater than 150 lbs..
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I The inspector reviewed the Safety Evaluation associated with the removal and found it to be
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thorough. The Safety Evaluation noted that the TMI probabilistic risk assessment confirmed that the removal of the chlorine detection system with the 150 lbs. cylinder restriction is not a
major contributor to core damage frequency. The licensee also determined that the concentration of chlorine from an offsite release would be insignificant and would not affect control room habitability.
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The inspector concluded that the licensee has adequately evaluated the removal of the CDS.
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7.3 Debris in Makeup System
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On March 12, 1993, following the reactor trip, the licensee noted that MU-V-17 was hanging up at various positions and not traveling full stroke. MU-V-17 is a safety related valve on
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the discharge of the makeup pumps and is used to control pressurizer level. This flow path is
isolated on an ES actuation signal. The licensee disassembled the valve and adjusted the
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actuator spring tension, but this did not correct the problem. The valve was disassembled again and valve inlet piping was inspected with a fiber optic camera. The licensee found two small pieces of debris. One piece was a cylinder that was approximately 0.5 inches long and l
0.192 inches in diameter made of 316 stainless steel and the other piece was a small curly
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piece of 304 stainless steel. Debris in the makeup system is of concern because it may be indicative of failure of a component, it can result in the failure of a downstream component, and it is possible to introduce the debris into the reactor coolant system.
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The licensee performed an operability evaluation based on a review of the most recent inservice testing of makeup system components. The licensee also reviewed the most recent makeup system testing per Surveillance Procedure (SP) 1300-3H, " Inservice Test of MU Pumps and Valves," and SP 1303-11.8, "High Pressure Injection." Based on these reviews and the absence of MU system component problems during normal operation, the licensee determined that no immediate equipment operability concern existed.
Plant Engineering attempted to locate the source of the debris by evaluating probable flows and tracing back to a component, such as a filter, where the debris could not pass. The licensee then made a list of all components within the boundaries to evaluate if any of the components contained a piece that would resemble the debris. The licensee has not been able to identify the source of the debris. Plant Engineering has tasked Technical Functions with performing another review to either find the source of the debris or assure themselves that debris did not come from a MU system component.
The inspector reviewed the most recent MU system surveillances and agrees with the licensee's operability decision. The inspector concluded that although the source of the debris has not been identified, the licensee has been aggressive and has taken a comprehensive approach in their search.
8.0 EMERGENCY PREPAREDNESS (71707)
On April 28,1993, the inspector observed the performance of the emergency preparedness annual practice exercise from the Emergency Operations Facility (EOF). The inspector found that staff augmentation was prompt and facility activation was timely. Radiological dose assessment and plant evaluations by the technical support group were effectively performed.
Event classifications were appropriate and notifications to off-site authorities were timely.
Emergency Support Director briefings of EOF personnel were comprehensive and frequent.
The inspector concluded that overall licensee performance at the EOF was good.
9.0 NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES At periodic intervals during this inspection, meetings were held with senior plant management to discuss licensee activities and areas of concern to the inspectors. A.t the conclusion of the reporting period, the resident inspector staff conducted an exit meeting with licensee management summarizing inspection activity and findings for this report period. Licensee comments concerning the issues in this report were docui *ented in the applicable report section. No proprietary information was identified as be: ag included in the report.