IR 05000289/1986010
| ML20215A672 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/02/1986 |
| From: | Blough A, Conte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215A590 | List: |
| References | |
| 50-289-86-10, SP, NUDOCS 8610060179 | |
| Download: ML20215A672 (25) | |
Text
{{#Wiki_filter:- _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - . . U.S. NUCLEAR REGULATOP.Y COMMISSION
REGION I
Report No.
50-289/86-10 Docket No.
50-289 License No.
DPR-50 Priority -- Category C Licensee: GPU Nuclear Corporation Post Office Box 480 Middletown, Pennsylvania 17057 Facility At: Three Mile Island Nuclear Station, Unit 1 . Inspection At: Middletown, Pennsylvania Inspection Conducted: June 27, 1986 to August 1, 1986
. Inspectors: R. Conte, Senior Resident Inspector (TMI-1) D. Johnson, Resident Inspector (TMI-1) M. Miller, Radiation Specialist J. Rogers, Resident Inspector (TMI-1) , l L. Cheung, Reactor Engineer F. Young, Resident Inspector (TMI-1) Reporting Inspector: M[[M 9.)-#G g{$1R. Contet'Ssnior Resident Inspector (TMI-1) Date Approved By: 8/dd 9-2 -d A. Blough, Chief Date Reactor Projects Section No. 1A Division of Reactor Projects Inspection Summary: Resident and region-based NRC staff conducted routine safety inspections (206 hours) of power operations, focusing on plant and personnel performance.
Specifically, items reviewed in detail in the operation, maintenance, and surveillance areas were: incore detector functional checks; battery surveil-lance; reactor vessel internal vent surveillance; reactor building sump inleakage; lighted annunciators; and inadvertent dropping of make-up purifica-tion filter. Other items included: decay heat valve operability; selected licensee reports submitted to the NRC, including monthly operating reports for 1985, radiological environmental monitoring report, electrical coil /contactor Part 21 report, and licensee event reports; heat exchanger biofouling; instal-lation of engineering safety features ventilation system; and licensee action on previous findings.
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Inspection Results: For the activities sampled, the inspector noted, in general, the licensee properly completed operational evolutions, maintenance, and surveillance activities consistent with regulatory requirements.
This was conducive to the operational readiness of such safety-related systems as the decay heat removal system. The licensee is diligently pursuing the correction of normally lighted main annunciators.
Although reflective of poor individual attention to detail, the missed battery surveillance was considered an isolated case with respect to proper and timely completion of required surveillance. The licensee needs to provide additional information on the acceptability of the alternate method for the reactor vessel internals vent valve surveillance.
Review of submitted reports indicated accurate and timely reporting with - appropriate corrective action planned or taken.
The licensee was very supportive of the on-site review by NRC staff of the engineered safety features ventilation system. Additional information is - needed from the licensee prior to the issuance of the verification inspection.
In general, licensee action on previous inspection findings was acceptable.
Problems continue to be noted on the overall adequacy of environmental quali-fication files. Certain documents continue to reflect errors and lack suffi-ciently documented test data and/or analysis for a knowledgeable individual to independently conclude on the environmental qualification of safety-related components.
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. . DETAILS 1.
Inttpduction and Overview 1.1 NRC Staff Activities The overall purpose of this inspection was to assess licensee activ-ities for the power operation mode as they related to reactor safety.
Within each area, the inspectors documented the specific purpose of the area under review, scope of inspections. The inspector made this assessment by reviewing information on a sampling basis through actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, or independent calcula-tion and selective review of listed applicable documents.
During this period, the senior resident inspector participated in ' Region I's review of the recently completed Performance Appraisal Team (PAT) findings (NRC Inspection Report No. 50-289/86-03) to determine appropriate enforcement action. The results of that . review will be documented in NRC Inspection Report No.
50-289/86-12.
1.2 Licensee Activities During this period the licensee operated the plant at full power.
Routine operations, maintenance, and surveillance were conducted along with the installation of modifications needed for the startup following the next refueling outage (cycle 6).
2.
Plant Operations 2.1 Scope of Review The NRC resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS) in the following areas: review of selected plant parameters for abnormal trends; -- plant status from a maintenance / modification viewpoint; -- control of ongoing and special evolutions, including -- control room personnel awareness of these evolutions; control of documents, including logkeeping practices; -- implementation of radiological controls; -- implementation of the security plan, including access -- control, boundary integrity, and badging practices; and, .-. _- -. .- - - _ _ - -.. --__ ---
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implementation of the fire protection plan, including fire -- barrier integrity, extinguisher checks, and housekeeping.
Because of additional resident office coverage at this facility, more detailed and frequent reviews of operating personnel performance were conducted to determine that: operators are attentive and responsive to plant parameters -- and conditions; -- plant evolutions and testing are planned and properly authorized; procedures are used and followed as required by plant policy; -- equipment status changes are appropriately documented and - -- communicated to appropriate shift personnel; the operating conditions of plant equipment are effectively -- monitored and appropriate corrective action is initiated when - required; ~ backup instrumentation, measurement, and readings are used as -- appropriate when normal instrumentation is found to be defective or out of tolerance; logkeeping is timely, accurate, and adequatel'y reflects plant -- activities and status; operators follow good operating practices in conducting plant -- operations; and, operator actions are consistent with performance-oriented -- training.
Specifically, the inspectors focused attention on the areas listed b.;ow.
General / Operations Control room operations during regular and backshift hours, -- including frequent observation of activities in progress, and periodic review of selected sections of the shift foreman's log and control room operator's log and other control room daily logs Areas outside the control room, including important-to-safety -- buildings detached from main plant buildings Selected licensee planning meetings -- Reactor Building sump in leakage monitoring -- - -. _ __ -- -. -. _ _ _ - _ _ - _ _ _ _ _. . .-- -
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Maintenance Reactor Building Atmospheric Monitor (RM-A2) repairs -- Fire Pump Building repairs -- Control Room annunciator work -- Makeup and Purification Prefilter changeout -- Surveillance Selected core power distribution manual calculations -- Diesel-Driven Fire Pump (FS-P3) surveillances for restoration to -- service after fire pump building damage - Monthly battery checks -- Reactor vessel internals vent valve exercise - -- As a result of this review, the inspectors reviewed specific areas in more detail as described in the sections that follow.
2.2 Findings , 2.2.1 Surveillances Using Plant Computer Data On June 27, 1986, the licensee began to experience data retrieval problems with the Bailey 855 computer. On a once per-shift basis, the Bailey 855 computer is used to perform heat balance calculation to check power range instrument read out, reactor coolant (RCS) leakrate, core imbalance, and quadrant power tilt calculations required by technical specifications. Without the capability of this computer, shift checks of incore instrument data, the heat balance, quadrant power tilt, and RCS leakrate calculation must be performed manually.
The inspector reviewed this problem to ensure that the licensee was meeting and had the ability to meet the requirements of the applicable sections of the technical specifications. The inspector reviewed applicable tech-nical specifications and the following licensee documents: Surveillance Procedure (SP) 1302-1.1, Revision 21, -- dated March 3, 1986, " Power Range Calibration;"
SP 1301-5.3, Revision 8, dated February 24, 1986, --
"Incore Neutron Detector - Monthly Check;" . _. _ -. . _ _ _ _ . -..- -. _ _ _ _ _ - - _. . ,.. .., -
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Operating Procedure (OP) 1103-16, Revision 16, dated -- June 5, 1985, " Heat Balance Calculations;" Abnormal Procedure (AP) 1203-7, Revision 18, dated -- March 20, 1986, " Hand Calculation for Quadrant Power Tilt and Core Power Imbalance;" OP 1105-5, Revision 12, dated July 26, 1985, "Incore -- Monitoring System;" and, SP 1303-1.1, Revision 17, dated February 15, 1986, -- " Reactor Coolant System Leak Rate."
AP 1203-7 requires that the quadrant power tilt and power imbalance hind calculations be performed once every two hours when the computer is unavailable and the plant is at - power.
SP 1302-1.1 requires that if the computer is out of service, a manual heat balance must be performed in accord-ance with this procedure once per shift for indicated neutron power verification. SP 1303-1.1 requires a daily
hand calculation of RCS system leak rate. During this period that the Bailey 855 computer was out of service (June 27 to July 3, 1986), the inspector verified that the above calculations were completed on time per the appropri-ate procedure and properly reviewed.
TS 3.5.4 requires that twenty-three individual incore detectors shall be operable to check core power distribu-tion and periodic calibration of the out-of-core detectors.
TS 4.1.1, Table 4.1-1, Item 34, requires that the incore neutron detectors must be functionally checked once per month.
SP 1302-1.1, Step 5.6, states that if the power range calibration cannot be completed due to the inoper-ability of the incore detectors, power shall be reduced to less than 80 percent.
SP 1301-5.3 requires that the Bailey 855 computer be operable to perform the monthly incore detector check. The licensee was aware that reactor power would have to be reduced to 80% on July 7,1986, due to expiration of the incore instruments surveillance.
The licensee repaired the Bailey 855 computer on July 3, 1986, by replacing a faulty cable with a spare cable. The inspector had no further questions.
2.2.2 Battery Surveillance On July 3, 1986, licensee representatives reported to the inspector that a Technical Specification (TS) required surveillance had been performed past the TS required date.
Specifically SP 1301-5.8, " Station Batteries" monthly check required to be performed by June 27, 1986, had been per-formed on July 2, 1986, five days later than required. The
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apparent cause of the delay was personnel failure to review the computer generated list of surveillances required to be accomplished during the week and to provide' the appropriate maintenance personnel with the required notification and approved surveillance procedure. Normally an assigned individual in the respective maintenance department (electrical, mechanical, I&C) reviews the file containing the " green sheets," which denote the specific surveillances required to be performed for a given week. The individual is responsible to notify the actual personnel who will accomplish the surveillance and provide the appropriate documents to accomplish the surveillance.
In this case, the designated individual was absent from the site for the
period and the designated relief individual failed to accomplish the task.
_ The inspector reviewed the completed surveillance, which was accomplished satisfactorily, although later than required. The inspector determined that the item was not reportable as defined by 10 CFR 50.73 as the surveillance ' was completed satisfactorily. A review of previous sur-veillances indicated no abnormal problems. Weekly battery surveillances were also reviewed and no problems noted.
The licensee's system for scheduling and accomplishing surveillances appears to work well, although the dependency on personnel review has the potential to cause this type of problem. The licensee does not plan any type of program-matic corrective action at this time with respect to the methodology of accomplishing surveillances. Supervisory personnel in the maintenance department oversight in this area will be enhanced to ensure that the problem does not reoccur.
The inspector concluded that this event was an isolated occurrence and that the licensee program to accomplish surveillance is adequate.
The inspector will continue to review completed surveil- , lances on a sampling basis to confirm that they are being accomplished as required by technical specifications. The inspector had no other concerns in this area.
2.2.3 Reactor Vessel Internals Vent Surveillance . By license Amendment No. 65, the licensee has received from
the NRC staff an extension allowing the deferral of Tech-nical Specification (TS) 3.1.11 on the operability of the reactor internal vent valves. The deferral is effective , until the completion of the fifth fuel cycle. Surveillance l _-. , _ _ _. _. - m . - _ _ _ -... _ _ _ _ _ _ _, _ _ _ _.. _.. - _ _ _ _ _ _ _ _ _ _. _, - -. _ _ _ _. _ _ _. - _ _. _ _
. o Procedure (SP) 1301-10.1, Revision 8, dated September 17, 1984, " Internal Vent Valve Inspection and Exercise," implements this requirement and was last completed April 21, 1982. The inspector reviewed the procedure. Several concerns dealing with the ability of the procedure to meet its intended purpose were identified by the inspector.
Internals vent valves are installed in the reactor core , support shield to prevent a pressure imbalance which might interfere with core cooling following a postulated inlet pipe rupture. Under all normal operating conditions, the vent valve will be closed.
In the event of the pipe rupture in the cold leg of the reactor loop, the valve will i open to permit steam generated in the core to flow directly to the leak and will permit the core to be rapidly recov-ered and adequately cooled after emergency core coolant has _ been supplied to the reactor vessel.
(Reference: TMI-I ' Operations Plant Manual, Section B-3, page 10) The surveillance procedure is used to demonstrate the . operability of the eight internal reactor vessel vent valves.
The procedure involves the use of a long-handled tool by an operator standing on either the main or aux-iliary fuel handling bridge to manually exercise the valve and allow visual inspection of the valve and its seating surfaces.
Included in the procedures are two methods for performance of the surveillance -- a normal method and an alternate method.
The normal method involves the use of the' polar crane and a scale attached to the long-handled tool. The scale is used to determine the force necessary to hold the valve in the full open position. A limit of 400 pounds of vertically upward-applied force to hold the valve fully open is established by TS 4.16.1.
However, the procedure does not ensure that the serial number of the instrument (scale) be recorded as needed for traceability of the calibration of
the scale used.
Further, the alternate method involves the use of the long-handled tool by personnel standing above the vessel to exercise and inspect the valve as in the normal method.
The alternate method of the procedure contains no provi-sions for the quantitative determination of the force necessary to hold the valve in the full open position.
The licensee's position in this matter is that one man performing the procedure cannot produce enough upward force to exceed the limit. The inspector determined that there was no documentation to support the licensee's position.
Further, the procedure does not require that only one man . ., ,_, -. - - -,, -. , -.. -. - - - -,,, -, -,, ,- - ,-c-.._.,,-w- .,---..-- ,---r -- .g,
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exercise the valves. The licensee committed to review the matter prior to the next performance of SP 1301-10.1.
This is unresolved pending completion of licensee action as stated above and subsequent NRC review (289/86-10-01).
2.2.4 Fire Pump Building Damage On July 18, 1986, the licensee sustained significant damage to the diesel-driven fire pump building. At 12:35 a.m., the fire suppression deluge system associated with the 18 auxiliary transformer actuated. Actuation of this system caused a pressure drop in the station's fire-main pressure causing additional fire pumps to start, including diesel-driven fire pump FS-P3.
Subs m ent licensee investigation revealed that the FS-P3 pump discharge check valve (FV-V27) failed open and the sump associated with FS-P3 was over- - pressurized. Overpressurization of this sump caused the sump's manway cover to be blown off and sent it through the roof of the diesel-driven fire pump building.
. No major damage was sustained to the diesel-driven pump; but support systems, such as electrical conduit and build-ing fire protection deluge piping were damaged when the floor grating was blown through the roof.
Repairs to these . components were completed. Temporary repairs were com-pleted to the building housing the diesel-driven fire pump.
Technical Specifications allowed continued plant operation with the fire pump inoperable, since only two-of-four fire pumps are required to be operable.
In addition to repairs noted above, the licensee issued Plant Incident Report No. 1-86-06, dated July 18, 1986.
Actions were initiated to facilitate repair in order to restore FS-P3 to service.
Long-term issues included: maintenance / engineering to determine the measures to -- prevent recurrence of the failure of FS-V-27; engineering to determine the function of FS-V-18 -- (check valve that permits screenhouse water to feed FS-P3 sump area) and determine if this valve is necessary for system operation; and, an open grating was installed at the suction pit -- access insteid of a solid manway to provide a vent path for the pit.
Engineering will evaluate this to determine if this change is acceptable.
Other information related to the event was: -_ _..-- - ._ _ _ _ ._ _ _ - _ . _ _ - _
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significance of and potential for flooding the Borated -- Water Storage Tank (BWST) tunnel via cable / conduit runs between the tunnel and screenhouse that occurred during the above-noted event; significance of and potential for leakage into aux- -- iliary building from the tunnel area that occurred during the above-noted event; and, qualification of safety-related cable (if any) for -- submersion for the short period that occurred during the event.
Because of licensee continued review / work in this area, the inspector will review related licensee documentation (including job tickets and engineering evaluations) for - completeness and adequacy in a future inspection (289/86-10-02).
2.2.5 RB Sump In-Leakage Increase
On or about June 8,1986, the reactor building (RB) sump level increase rate rose from about 100 gallons per day (gpd) to 375 grd. Since an increase in RB sump rate is one of the indicatuns of reactor coolant system (RCS) leakage, the licensee in'tiated an investigation of the cause.
Licensee personnel made entries into the reactor building during the period to draw RB sump water samples and inspect inside the D-ring for possible leaks.
. The results of the chemical analysis showed no presence of boron in the RB sump indicating that the leak (s) were from the secondary side of the plant. Both visual inspections were inconclusive as far as detecting any noticeable leaks.
The 375 gpd in-leakage rate has remained constant through-out this report period.
The licensee believes that one or more secondary valves have stem packing leakage, but they are continuing their investigation. The inspector will follow further licensee actions in this matter.
2.2.6 Lighted Annunciators In an attempt to have all annunciators in the control room de-energized (black board concept) at power operations, the licensee has an active program to correct and/or modify alarm circuitry to achieve this goal. During the period the licensee worked on two alarms, Station Battery Ground and [RCS] Feed and Bleed Auto Terminate (RFBAT) Alar e- . t .
, The inspector reviewed the work that was associated with both of these alarms to ensure that the werk being con-ducted was controlled and performed in accordance with applicable procedures.
For the Station Battery Ground Alarm, the licensee adjusted the alarm setpoint within the allowable band.
For the RFBAT alarm, the licensee attempted to have the light normally off with no RCS feed and bleed operation. Upon switch adjustment in the alarm circuitry, the licensee found a burned out relay. At the close of the inspection period, the licensee had an out-- of-service sticker on the alarm and they were in the process of making repairs. There is minimal impact with this alarm being out of service, since there are other obvious indicators in the control room to signal to the operator the auto termination of the feed and bleed , process.
- The inspector reviewed the applicable job tickets associ-ated with the stated work and determined the work was
performed in accordance with stated requirements. The - review of additional work associated with alarms in the L control room will continue in subsequent inspection reports.
2.2.7 Inadvertent Oropping of a Make-Up Purification Filter During replacement of a filter assembly in the make-up purification (MU) system on Monday, July 28, 1986, the filter was inadvertently dropped. The filter is used to
aid in cleanup of reactor coolant system (RCS) water by removing particulate matter from the water. The contact radiation reading on the filter was approximately 400 rem per hour.
It was being moved using a long-handled tool to minimize personnel radiation exposure. Maintenance per-sonnel were in the process of transferring the filter from the make-up system filter housing unit to a shielded transfer cask when the filter slipped off the tool and fell , onto the floor. No personnel hazard resulted. The filter i remained intact and the licensee successfully picked up the filter, again using a long-handled tool, and placed it in ' the transfer cask the same day.
The inspector reviewed the event, which included discus-sions with key licensee managers, to determine if the event had any immediate adverse effect on power operation or created a radiation health problem. The inspector > determined that the event did not adversely effect plant operations or produce significant radiation problems. At the time of the inspector's review, the licensee was still in the process of evaluating the event.
The final review by the NRC will be conducted in the next inspection period (50-289/86-13).
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2.3 Conclusion The licensee properly implemented the evolutions, maintenance, and surveillance activities noted above.
Event response was appropriate with corrective actions planned or taken. The missed battery sur-veillance was identified by the licensee and appears to be an isolated case.
It does point out the repetitive aspect of poor attention to detail on the part of individuals conducting necessary work activities at this facility. The licensee needs to provide additional information on the acceptability of the alternate method for the reactor vent valve internals surveillance.
3.
Decay Heat Valve Operability 3.1 Scope of Review _ The inspector conducted a review of the licensee's procedures, main-tenance and surveillance testing of the decay heat system (OH) valves that are included in the licensee's 10 CFR 50.55 a(g) inservice testing (IST) program. This program was submitted to the NRC in a
letter dated July 10, 1984, and was supplemented by a letter, from Hukill, GPUN, to Stolz, NRC, dated March 3, 1986. This documentation provided the basis for review of the DH system valves included in the IST program for valves important to safety.
The inspector reviewed the following documents related to testing and maintenance of DH system valves: Surveillance Procedures (SP) 1300-38, Revision 23, dated -- March 11, 1986, "0H System Valve Operability;" SP 1300-3P, Revision 10, dated June 20, 1984, "IST of Check -- Valves During Shutdown;" SP 1300-3R, Revision 14, dated August 13, 1985, "IST of Valves -- During Snutdown and Remote Indication Check;" SP 1300-3T, Revision 9 June 30, 1986, " Pressure Test of -- DH-V22A/B;" SP 1300-3Q, Revision 23, dated May 28,1986, " Quarterly IST of -- Valves During Normal Operations;" 1303-11.16, Revision 17, dated April 1, 1986, " Decay Heat -- System Leakage;" SP 1303-11.54, Revision 4, dated November 20, 1984, " Low -- Pressure Injection."
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Corrective Maintenance (CM) Procedures 1410-V10, Revision 11, -- dated May 16, 1986, " Disassemble, Inspect, Clean, and Repair Gate and Globe Valves;" CM 1410-V13, Revision 6, dated February 15, 1986, " Valve -- Packing Repair / Adjustment;" CM 1410-V14 Revision 9, dated February 14, 1986, " Repack Valves -- in Borated Water Systems;" and, CM 1410-V18, Revision 6, dated April 22, 1986, " Check Valve -- Repair."
The inspector also reviewed drawing C-302-640 for the OH system and various job tickets related to DH system valve maintenance.
- 3.2 Findings The licensee's IST program, as it relates to DH system valves, is properly established and implemented. This was evidenced by com- - pleted test procedures, a review of completed maintenance over the past year, and the results of this testing and maintenance.
Test results reviewed showed that the testing was completed on schedule and that the valves operated with consistency. No major changes in valve stroke times were noted. Check valve leakage was well within tolerance and showed no degradation from previous results.
Extensive preventive maintenance was done on most decay heat system valves prior to restart and during the past twelvt months.
Very few problems were noted with the exception of DH-VSA/B (BWST suction check valves), which are known to have seat leakage but will be repaired during the next refueling outage.
DH system valves appear to be in good material condition and should operate when required.
It should be noted that the licensee has applied for various exemp-tions to the requirements of ASME, Section XI, as it pertains to the testing of various OH system valves. These exemptions have not been granted by NRR to date and the licensee has untti October 3, 1986, to resolve the conflicts.
2.3 Conclusion The inspector determined, based on the above review, that the IST program for DH system valves was implemented as documented in the above-referenced letters.
The inspector had no safety-related concerns for the operation of the DH system valves.
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4.
Report Reviews 4.1 Monthly Operating Reports The inspector reviewed the licensee's Monthly Operating Reports (MORs) for the calendar year 1985. The inspector reviewed these reports to verify that operational statistics had been accurately reported and that the narratives summaries of the month's operating experience were contained therein. On a sampling basis, the inspec-tor independently checked the information in the summaries to ensure the information was complete and correct.
The report summarizes the major plant evolutions or activities that occurred during that month.
Each report contains a brief description of plant operating history and major safety-related maintenance performed.
In addition, statistical data about plant performance is - documented in the report.
The inspector determined that the licensee monthly reports complied with the applicable requirements (TS 6.9.1.c) and that required . information was present in each report.
4.2 Report on Radiological Environmental Monitoring Program The inspector reviewed the licensee's Radiological Environmental Monitoring Program annual report for 1985. As required by TS 6.9.4, this report summarizes the results of the sampling and analyses of environmental media to determine the radiological impact of station operations.
These environmental media include air, water, vegeta-tion, and aquatic plants and animals.
In addition, direct radiation is monitored by placement of thermoluminescent dosimeters at various locations around the station.
As a result of this review, the inspector determined that the licensee has generally complied with its Technical Specification (TS) requirements for sampling frequencies, types of measurements, analytical sensitivities, and reporting schedules.
Exceptions to the sampling and analysis program were adequately explained; e.g., loss of sample due to vandalism. The report included summaries of the laboratory quality assurance program and of the land use survey.
The analyses of environmental samples indicated that doses to humans from radionuclides of station origin were negligible.
4.3 Part 21 Report Review In a letter dated June 19, 1986, (Serial No. NS-NRC-86-3142) West-inghouse Electric Corporation notified the Director of the NRC Office of Inspection and Enforcement of a 10 CFR 21 reportable item (appli-cable to TMI-1) concerning the possible failure of coils used in Class IE motor starters and contactors.
The failures were noted by
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non-nuclear customers and the same type coils were used in safety-related (Class IE) applications although no failures were noted by ' the nuclear customers. The high rate of failures was noted upon initial energization of the coils and the failures were traced to coils manufactured between June 1, 1984, and December 31, 1985, from a particular plant.
Failure of the coils could cause a loss of power to any loads supplied by the starters and contactors. The vendor committed to replace these coils with those manufactured subsequent to January 1, 1986.
(These new coils were manufactured using dif-ferent materials and processes and are apparently free of defects.)
The vendor also provided interim advice on the potentially defective coils received by licensees. This advice included subjecting the coils to a five-day energization period before use or, if possible, not using them at all until new coils were received by the licensee.
The inspector reviewed the subject report to ascertain the nature of - the problem (deficiency) as related to TMI-1.
Subsequently, he reviewed licensee corrective actions to ascertain if the licensee received complete and appropriate information from the applicable vendor and if licensee corrective actions were adequate to resolve
the deficiency consistent with vendor recommendations.
The inspector confirmed that the licensee received the subject report and initiated appropriate corrective action.
Licensee review identi-fled that the subject coils were used in new motor control centers (MCCs) being installed for the engineered safety features ventilation system for the fuel handling building. The coils, along with'the MCCs, were in the warehouse and not installed in the plant as yet.
Licensee representatives initiated a material nonconformance report (No. 140-86, dated June 26,1986). Hold tags were placed on the defective material. Quality Control witnessing was established for coil testing and replacement. The licensee intends to replace the potential defective coils with the new coils supplied by the vendor.
However, following the advise of the vendor, the startup and test group may use the old coils for testing purposes.
The inspector was satisfied that the licensee was in sufficient control of this material problem. Quality Control personnel indi-cated that the material nonconformance report (MNCR) would not be closed until new coils were received and they replaced the old coils.
Further, control was exhibited by Quality Assurance (QA) department's established hold / witness points. There also appeared to be adequate communications among various cognizant licensee groups in resolving this problem (licensing, startup and test, on-site technical func-tions, and quality control). Accordingly, this Part 21 report (289/86-PT-02) is considered closed.
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4.4 Licensee Event Reports (LERs) In-Office Review The inspector reviewed the LERs listed below, which were submitted to the NRC Region I office pursuant to 10 CFR 50.73.
Based on resident office review of the LERs, the inspector determined that corrective action discussed in the licensee's report was appropriate, that the information satisfied reporting requirements, and that there were no generic issues.
In addition, the inspector determined that the event is not appropriate either for classification as an Abnormal Occurrence or for Licensing Board, Appeal Board, or Commission notification.
LER 86-008, dated May 21, 1986 - On April 21, 1986, a partial -- loss of the off-site power, due to a ground fault failure of the "A" Auxiliary transformer supply breaker to the 10 4160 Vac electrical bus.
The failure of the closure mechanism of the - subject breaker is still under investigation.
(This event was reviewed in NRC Inspection Report (50-289/86-06.)
LER 86-009, dated May 21, 1986 - As of April 22, 1986, the
-- Environmental Qualification records for reactor building emer-gency cooling fans cable were not available.
This condition was discovered during a sample walkdown of selected components to re-verify environmental qualifications of these components. The subject cable was replaced with properly environmentally quali-fled cable.
(This event was reviewed in NRC Inspection Report 50-289/86-06.)
- LER 86-010, dated May 22, 1986 - On April 23, 1986, the reactor
-- tripped on high reactor coolant system (RCS) pressure resulting from a loss of main feed water while transferring feed pump turbine steam from auxiliary to main supply.
The licensee has initiated procedure changes to provide more guidance during this portion of plant startup.
(This event was reviewed in NRC Inspection Report 50-289/86-06.)
LER 86-011, dated July 2, 1986 - On June 2, 1986, the reactor -- tripped on a turbine trip anticipatory signal, which was caused by loss of both electro-hydraulic control oil pumps. The loss of both pumps occurred during turbine plant electrical repairs with an abnormal electrical lineup.
(This event was reviewed in , NRC Inspection Report 50-289/86-09.)
5.
Heat Exchanger Biofouling 5.1 Introduction In response to a temporary instruction (No. 2515/77, dated April 16, 1986) issued by the NRC's Office of Inspection and Enforcement, the inspector conducted a review of the licensee's programs and proce-dures for dealing with biological fouling of cooling water heat _ .
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exchangers. This problem has been the subject of NRC Bulletin 81-03 and Information Notices 81-21 and 83-46.
The licensee has also been made aware of this problem via the Institute of Nuclear Power Opera-tions (INPO) Significant Operating Event Report (50ER), specifically 50ER 84-1.
The biofouling problem is normally evidenced by the presence of marine organisms either growing in the heat exchangers or being introduced from the particular water supply being used. The licensee uses the Susquehanna River at TMI-1 for the supply to the decay heat coolers and the nuclear services coolers.
5.2 NRC Review The inspector reviewed the following procedures that pertain to use of cooling water systems.
Operating Procedures (0P) 1104-30, " Operation of the Nuclear - -- River Water System," 1104-32, " Decay Heat River Water," and 1104-38, "The Reactor Butiding Emergency Cooling Water Systems" Surveillance Procedures (SP) 1302-6.7 " Monitoring of Silt -- .
Buildup in the R. W. Screenhouse" and 1301-11.9, "RB Emergency Cooling System" Maintenance Procedure M-144, " Cleaning Heat Exchangers" -- , The inspector also examined system drawings for the applicable systems and discussed the operation of these systems with operations personnel.
5.3 Findings The licensee had responded to Bulletin 81-03 that, at the time, no problems existed at TMI-1 with respect to fouling of heat exchangers due to marine organisms, as had been experienced at other facilities.
The licensee maintains an ecological monitoring program that examines the cooling water supplies for the presence of the subject marine organisms. To date none have been present at this point in the Susquehanna River (York Haven Reservoir). The subject microorganisms have been detected approximately 48 miles south of TMI-1 (downstream) but, no migration upstream has been detected.
The licensee also maintains a dredging program every two years in the area of the main intake structure and the soil from this evolution is examined.
Presently, samples of the sludge at the intake structure are taken and analyzed three times annually, i In the licensee's preventive maintenance program, the heat exchangers are opened periodically and examined for evidence of sludge buildup or microorganism growth.
No problems have been detected at this point in time.
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, Only one system, the reactor building emergency cooling system, employs flow measuring equipment that would detect a flow degradation condition. A surveillance is performed on this system on a refueling outage basis to compare the maximum flow rate in the system with design flow rates.
Previous data was reviewed by the inspector and found to be satisfactory.
The inspector reviewed operating procedures for the subject cooling water systems. These procedures address the various alignments of cooling water systems to increase cooling water effectiveness. The operating procedures or alarm procedures do not address the specific situation that is discussed in this section; i.e, heat exchanger degradation due to marine microorganisms.
Instrumentation is avail-able on all cooling water systems (pressure / temperature) to detect possible loss of cooling ability.
, 5.4 Conclusion The licensee's programs and procedures for dealing with marine , biofouling with the potential for related heat exchanger degrada-tion are adequately addressed at TMI-1. No major problems have been identified to date.
The inspector did not identify any safety-related concerns in this review.
6.
Engineered Safety Featured (ESF) Ventilation System The licensee's design and initial installation of their ESF ventilation system for the Unit 1 fuel handling building was reviewed on July 21, 1986. A technical meeting was held at the licensee's facility to discuss questions identified by the assigned technical reviewers from NRR and to visually inspect components of the system where installation was completed.
Region I also participated in this meeting.
The licensee had coemitted to install an ESF ventilation system in accord-ance with Regulatory Guide 1.52, Revision 2.
Questions concerning the licensee's ESF design addressed primarily the following issues: measure-ment / assurance of maintaining a negative pressure in the fuel handling building, seismic Class 1 conditions, if applicable; source term (s) for fuel rod drop accident; and, identification and required analyses for a new effluent pathway which would exist during fuel movement and accident conditions.
The licensee stated that a written response to these questions would be provided. Proposed technical specifications would also be submitted within three weeks. The inspector stated that a review of test data and a visual inspection of the installed system by the NRC would be conducted prior to any irradiated fuel movement from the reactor building to the THI-1 fuel handling building (289/85-20-03).
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l 7.
Licensee Action on Previous Inspection Findings 7.1 (Closed) Inspector Follow Item 289/84-18-02: Licensee to Revise the Evaporative Loss Term in the Reactor Coolant System (RCS) Leak Rate.
In the previous update to this item (NRC Inspection 50-239/85-30), the licensee committed to revise the evaporative loss term to zero and to make necessary procedural and software changes to reflect this commitment by January 31, 1986. The inspector reviewed Surveillance Procedure (SP) 1303-1.1, Revision 17, dated February 15, 1986, " Reactor Coolant System (RCS) Leak Rate" and the Plant Computer Program "RCLEAK", dated June 17, 1986, used to calculate RCS leak rate to confirm that the evaporative loss term was set to zero by the licensee.
. The SP 1303-1.1 specified computer method (automatic) (page 32, line 33) and the manual method (page 45, line 37) set the evaporative loss term to zero.
Line 562 of the RCLEAK program sets the evaporative
loss term to zero.
Prior to January 17, 1986, the value used was i
0.27 gpm.
i i The inspector also selectively reviewed RCS leak rate data for the month of January, April, and July 1986. Data up to January 17, 1986, i reflected the use of the non-zero evaporative loss term.
Data l obtained subsequent to January 17, 1986, reflects a zero evaporative ! loss term.
i The inspector also independently calculated applicable RCS leak rate data using licensee input data and a generic NRC " basic" computer , program "RCSLK9" as specified in NUREG 1107.
Licensee and NRC data I are tabulated below. Although plant specific parameters are used, , l the program is somewhat generic so RCSLK9 does not consider the l reactor coolant pump's No. 3 seal combined leakoff (0.1044 gpm).
l The correlation of licensee and NRC data is as follows.
NUNIDENT = NGLR IDENT-N CORRECTED NUNIDENT " GLR + 0.1044 - NIDENT =NUNIDENT + 0.1044 LUNIDLK = Licensee Unidentified Leakrate NUNIDENT = NRC Calculated Unidentified Leakrate N = NRC Calculated Gross Leak Rate based on RCS mass GLR balance (which should correlate with L ~ RCSLPL Licensee Calculated Leakage Plus Losses Term)
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N = NRC Calculated Identified Leak Rate (essentially IDENT reactor coolant drain tank mass change) 0.1044 = Total leakage due to No.3 RCP seal leakoff - TA8LE 1 RCS LEAK RATE DATA (All Values GpM) DATE/ TIME N UNIDENT CORRECTED DURATION > L RCSLPL N GLR (NUREG 1107) N UNIDENT.
L UNIOLK 1/14/86 7:13:05-0.0088-0.01 .21 .38* -0.3734 ' 2 Hours 1/14/86 20:10:26 0.1228 0.12-0.13-0.40* -0.2918 , 2 Hours 4/21/86 23:10:51 0.2268 0.23 0.08 0.18 0.1825 2 Hours 4/23/86 7:33:55 0.5729 0.59 0.44 0.54 0.5319 2 Hours . +6/29/86 7:49:00 0.243 0.24-0.13 0.03-0.021 2 Hours +6/30/86 18:49:00 0.1247 0.12-0.09 0.01 0.0134 2 Hours +7/1/86 08:44:00 0.0832 0.08-0.13-0.03-0.0124 2 Hours +7/2/86 01:42:00 -0.0187 0.10-0.08 0.02-0.0895 2 Hours 7/20/86 08:07:24 0.4641 0.46-0.02 0.08 0.0883 2 Hours 7/21/86 01:58:54 0.3282 0.33-0.04 0.06 0.0697 2 Hours 7/29/86 07:46:04 0.4758 0.48-0.01 0.09 0.0906 2 Hours Columns 2 and 3; 5 and 6 correlate + 0.2 gpm in accordance with NUREG 1107
- Evaporator term (0.27 gpm) was appTied for this time period
+ License manual calculation
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The inspector concluded that the licensee has implemented his commi,tments regarding evaporative loss term.
Further, licensee leak rate determinations were in good agreement with those calculated by the NRC staff program.
7.2 (Closed) Violation 289/85-30-02: Failure to Verify Fire Door to be Functional.
In a response letter, dated March 14, 1986, the licensee acknowl-edged the finding and proposed appropriate corrective actions and measures to prevent recurrence.
The root cause was due to the poorly coordinated issuance of procedure revisions which trans-ferred functional check of the door (0-108) from one procedure to another. SP 1303-12.20, Revision 4, December 20, 1985, " Fire Door Inspection - Control Building and Diesel," incorporated the check of D-108. Selected completed data to this SP for January and - July 1986 indicated proper completion of the surveillance procedure.
The inspector also reviewed an internal memorandum, dated March 3, 1986, frcm the Review Program Coordinator for TMI-1.
This memorandum' emphasized the need for procedure preparers / owners / responsible offices to clearly define or identify procedure revisions that need to be processed and/or issued simultaneously. The coordinator will be specifically tracking these special case procedure revisions once properly identified by applicable departments.
The inspector concurs that the licensee achieved compliance with TS 4.18.7.2.C when Revision 4 to SP 1303-12.20 was issued. The inspector adds that full compliance will continue to be achieved assuming continued proper implementation of the SP and attention to detail on the part of personnel when they pass through fire doors such as D-108. An existing ventilation problem sometimes prevents door D-108 from completely closing and latching unless personnel are careful to assure that the door is closed.
7.3 (0 pen) Unresolved Item 289/86-06-07: Environmental Qualification of Electric Cables 7.3.1 Scope of Review The Kerite Cable (TI-162) files and BIW SR-insulated and SR-jacketed cable (TI-163) files were reviewed to ascertain whether the files contained sufficient evidence that these cables were qualified for the environmental conditions in which they must operate and that the qualification documents in the files were auditable.
Documents reviewed in this determination included:
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Kerite FR Cables System Component Calculation Worksheet (SCEW) Sheet -- TI-770-044, Revision 6, dated April 18, 1986; Franklin Institute Research Laboratories, Final -- Report No. FC4158, " Tests of Electrical Cables Under Simultaneous Exposure to Gamma Radiation, Steam, and Chemical Spray While Electrically Energized," dated June 1975; GPUN Calculation No. EQ-TI-162-02, Revision 0, -- " Analysis of Temperature / Pressure Test for Kerite FR Cable," dated April 17, 1986; EBASCO Document for " Qualification Documentation for -- Kerite FR/FR Control Cables," dated June 10, 1977; - EDS Nuclear Report 02-0370-1058, Revision 2, dated -- 6/81 (stea:n profile); . GPUN TOR 282, Revision 4, dated July 25, 1984 -- (chemical spray, radiation, normal / ambient temperature, and plant conditions); Division of Operating Reactors (DOR) Guidelines -- Qualification Report Review Checklist for Kerite FR Cables; Design Change Notice No. 048110, "To include Kerite -- FR Cables into EQ Master List," dated July 8,1986; and, BIW SR Insulated and SR Jacketed Cables SCEV Sheet TI-770-045, Revision 1, dated April 21, -- 1986; 00R Guidelines Qualification Report Review Checklist -- for BIW SR Insulated Cables; 00R Guidelines Qualification Assessment Report; -- BIW Cable Test Report No. B924, dated December 1983; -- Franklin Institute Research Laboratory Test Report -- F-C2935, " Test of Electrical cables Under Simulated Post-Accident Reactor Containment Service for Continental Wire and Cable Company", datea October 1970; Arrhenius Calculation; -- I ..
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' GPU Cable Specification No. SP 1101-31-206, Revision -- 0, dated October 23, 1980; and, Design Change Notice No. 042865, "To Include BIW -- Cables into EQ Master List," dated April 22, 1986.
7.3.2 Findings 7.3.2.1 The test conducted for BIW cables did not include Loss of Coolant Accident (LOCA) tests. The System Component Evaluation Work (SCEW) sheets indicated that these cables were used both inside and outside the reactor containment and that tney were subject to severe temperature / pressure and steam environment. The D0R Guidelines, paragraph 5.1, states that qualification should be based on type test.
- The licensee did include in their EQ file the LOCA test data for Continental Wire and Cable Company SR insulated cable, which the licensee used as a basis for justifica-tion of qualification.
However, the configurations of
these two types of cables were different.
The BIW cables, as installed at TMI-1, sere silicone rubber insulated and silicone rubber jacketed, while the Continental cable, as tested in 1970, was silicone rubber insulated, glass braid over silicone, glass finished, Mylar binder tape, overall asbestos braid, asbestos finished. The EQ file, as reviewed by the inspector, did not contain justification for the differences between these two types of cables.
Further, in the DOR Guideline Qualification Report Review Checklist, Items 2, 16, 22,23, and Note No. 4 (on page 8 of 9) did not indicate that the qualification was not based on type test and that insulation resistance (IR) measurements were taken from Continental Cables, not from BIW cables.
According to the licensee, the cables located inside the reactor containment were required to function after main steam line break (MSLB) accidents. However, the MSLB profile was not shown in Figure 1 (page 9 of 9), which showed LOCA profile versus " qualified" profile.
.On July 18, 1986, the licensee provided a document EQ-86-821, dated July 17, 1986, stating that the asbestos braid, asbester finish were more porous than the SR jacket. However, the glass braid and glass finish were never addressed. The synergistic effect of these multicoating test cables was not compared to that of the BIW cables.
Insulation resistances and leakage currents of the BIW cables during the post-LOCA conditions were not predicted and justified.
Variation in manufacturing
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process and the insulation thickness tolerances between the BIW cables and the Continental cables were not discussed.
A conference call was made between the licensee and NRC RI personnel on July 31, 1986. The licensee stated that they could provide additional data and analysis to support the qualification of the BIW cables.
NRC Region I will review the additional information to be provided by the licensee. Also, this aspect of the unre-solved item continues to be unresolved pending a decision by NRC Office of Inspection and Enforcement on whether analysis is acceptable in lieu of type testing for the qualification of BIW cables.
- 7.3.2.2 The inspector identified the following deficiencies in the EQ file for Kerite FR Cables.
- SCEW Sheet TI-770-044 indicated that the cables were
-- qualified for a post-DBA pressure of 84.4 psia.
However, the required pressure block was left blank.
The licensee stated that the omission of the required pressure was due to oversight and that this would be corrected.
The SCEW sheet indicated that the qualification was -- for both power and control cables, while the EQ file indicated that the qualification was for control cable only. This inconsistency was brought to the attention of the licensee. The licensee stated that the EQ file was correct and the SCEW sheet would be revised.
Subsequent to completion of the inspection, the licensee submitted to NRC the following documents.
Design Change Notice No. 048111, dated July 9, 1986, __ which requires SCEW sheet TI-770-043 be revised to include the " required pressure: and to delete the power cables" from the qualified item.
The required pressure of 50.6 psia was added to SCEW -- Sheet TI-770-043, dated July 9, 1986, and the " power cables" portion was deleted. All these changes were properly initialled and dated.
Thus, the discrepancies for the Kerite cable have been correcte r m
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This aspect of the unresolved item is considere cable was adequately established by the licensee.
Installation of ESE (0 pen) Unresolved Item (289/85-20-03):See " Details" paragraph 6.
7.4 yentilation System System.
8.
Exit Interview The inspectors discussed the inspection scope and findings with the 31, 1986.
licensee management at a final exit interview conduct following: J. Colitz, Plant Engineering Director, TMI-1 . R. Harper, Corrective Maintenance Manager , C. Incorvati, Audit Supervisor, TMI-1 T. O'Connor, Plant Engineering, TMI-1 - S. Otto, TMI-1 Licensing Engineer J. Randazzo, TMI-1 Licensing Engineer D. Tuttle, Rad Con Services The inspection results, as discussed at the meeting, are summarized Licensee representatives the cover page of the inspection report. indicated that none of the , safeguard information.
Unresolved Items are matters about which information is required in to ascertain whether they are acceptable items, violations, or deviations.
documented in paragraphs 2.2.3, 7.1, and 7.3. }}