IR 05000289/1988029
| ML20235Y912 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/27/1989 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235Y905 | List: |
| References | |
| 50-289-88-29, IEIN-88-074, IEIN-88-74, NUDOCS 8903140679 | |
| Download: ML20235Y912 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket / Report No. 50-289/88-29 License:
DPR-50 Licensee:
GPU Nuclear Corporation P. O. Box 480 Middletown, Pennsylvania 17057 Facility:
Three Mile Island Nuclear Station, Unit 1 Location:
Middletown, Pennsylvania N
Dates:
December 3,1988 - January 14 and 18,1989 Inspectors:
R. Conte, Senior Resident Inspector, TMI D. Johnson, Resident Inspector, TMI (Reporting Inspector)
T. Moslak, Resident Inspector, TMI A. Sidpara, Resident Inspector, TMI Approved by:
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C. Cowgill, Wief, Reactor Projects Section No.1A Date Inspection Summary Areas Reviewed: The NRC staff conducted routine safety inspections of plant outage activities, plant start-up, and power operations. The inspectors reviewed the following functional areas: plant operations, equipment operability (maintenance and surveillance); and engineering / technical support (see Table of Contents).
Results: Plant operations, shutdown, and start-up activities were conducted safely and in a controlled, deliberate manner. The
"D" reactor coolant pump (RCP) seal overhaul was properly managed, with extensive use of contractor assistance for evaluation of alignment problems. Two maintenance items became unresolved items.
One concerned the length of time that job tickets remain open after work completion; the second involved installation of replacement-in-kind components that differ from the original part.
Engineering support was adequate as evidenced by a thorough evaluation of a concern with high pressure injection (HPI) pump operation during the recirculation phase (post-LOCA).
Licensee action on previous inspection findings was adequate.
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' TABLE OF CONTENTS.
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1.0 Introduction and 0verview............................................
I 1.I' Licensee Activities.............................................
I 1.2 NRC-Activities....................................................
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1.3 ' Persons Contacted...............................................
I 2.0 Plant Operations (NIP 71707).........................................
2.1 Criteria / Scope of Review........................................
2.2 Outage for_ "D" Reactor Coolant Pump Seal Repai r.................
2. 3. Plant Operations Summary.........................................
'3 3.0 Equipment. Operability Review - Maintenance / Surveillance (NIP 61726/62703)...................................................
-3 3.1 Criteria / Scope of. Review........................................
3.2
"D" Reactor Coolant Pump Seal-Repair............................
3.3. Emergency Diesel Generator Annual Mai ntenance...................
,3.4-Job. Ticket C1oseout.............................................
3.5 Plant. Materiel Condition........................................
3.6' (Closed)~ Violation NC4 (289/88-13-03): Pressurizer Relief Valve Ta11 pipe Support Loose Bo1ts..................................
3.7 Eq u i pme nt Op e rab i l i ty S umma ry...................................
4.0 En g i n e e ri n g S up po rt..................................................
4.1 NRC Information Notice 88-74....................................
4.2 Li cen see Letter Concerni ng RC-V-28..............................
5.0 Management Meeting (NIP 30703).................................
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ATTACHMENTS Attachment 1 - Activities Reviewed
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DETAILS 1.0 Introduction and Overview 1.1 Licensee Activities In December 1988, the licensee shut down the reactor to troubleshoot and repair tha excessive No. I seal leak-off for the "D" reactor coolant pump (RCP). On December 15, the plant was taken off line.
Repairs were com-pleted.
Power operation resumed on December 29. As of January 14, 1989, the TMI-1 reactor was at 100 percent power.
1.2 NRC Staff Activities The purpose of this inspection was to assess licensee activities for reactor safety, safeguards, and radiation protection. The inspectors made this assessment by reviewing information on a sampling basis through actual observation of licensee activities, interviews with licensee per-sonnel, or independent calculation and selective review of applicable documents.
NRC staff inspections are generally conducted in accordance with NRC In-spection Procedures (NIPS). These NIPS are noted under the appropriate section in the Table of Contents to this report.
1.3 Persons Contacted
- R. Barley, Manager, Plant Engineering
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G. Broughton, Operations / Maintenance Director
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J. Colitz, Manager, Plant Engineering
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- J. Fornicola, Manager, Quality Assurance
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- H.
Hukill, Vice President and Director, TMI-1
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C. Incorvati, Audit Manager
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M. Nelson, Manager, Safety Review
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- S. Otto, TMI-1 Licensing Engineer
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A. Palmer, Manager, Radiological Field Ope, r t'.ons
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M. Ross, Plant Operations Engineer
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- H. Shipman, TMI-1 Operations
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D. Shovlin, Plant Materiel Director
- P. Snyder, Manager, Plant Materiel Assessment
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- C. Smyth, Manager, Licensing
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- M. Wells, Communications
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- Denotes attendance at final exit meeting (see also Section 5.0).
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2.0 Plant Operations 2.1 Criteria / Scope of Review The resident inspectors routinely 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;
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plant status from a maintenance / modification viewpoint, including plant housekeeping and fire protection measures;
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control of on going and special evolutions, including control room personnel awareness of these evolutions; control of documents, including logkeeping practices;
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implementation of radiological controls; and, implementation of the security plan, including access control,
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boundary integrity, and badging practices.
Specific findings are addressed below.
2.2 Outage for "D" Reactor Coolant Pump Seal Repair The inspectors made several observations of licensee performance during the December 15-29, 1988, plant shutdown, outage, and start-up.
The licensee conducted this outage to correct excessive No. I seal leak-off flow from the "D" RCP.
This situation was similar to the one experienced by the licensee during the September 1988 outage to repair a leak-off problem from the same pump seal. The licensee acted conservatively'and shut down the plant when leak-off flow exceeded a nominal 3-5 gallons per minute (gpm).
Generally, operations were conducted in a safe, controlled manner.
No plant operational problems were experienced during cold shutdown except during one evolution for switching the operating decay heat (DH) removal
train. When operations must transfer to another DH train, the normal
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practice is to first ensure that the idle DH loop is operable, secure the operating pump, and then start the idle DH pump. Operating Procedure (0P) 1104-4, " Decay Heat Removal System Operation," is used for this evolution. On December 23, 1988, the operator accomplished a train switch by first. starting the idle pump. This resulted in more total DH flow. When the water level in the reactor vessel is lowered for main-tenance, as it was at this time, the total flow is limited by OP 1104-4 to prevent vortexing at the DH pump suction.
For several minutes, this flow was exceeded. The operator recognized this error and secured the first DH pump.
Flow was restored to normal. No problem with DH pump
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3-operation was noted. That vortexing did not occur was evidenced by the.
fact that. pump vibration was normal and flow was stable.
Investigation by the licensee revealed that procedural guidance to prevent this situ-_-
ation-was inadequate. A Procedure Change Request (PCR) was issued. The L
- inspector. reviewed the PCR and concluded the additional guidance should prevent recurrence.
K The inspector observed selected portions of RCS fill, plant heat-up, and power ascension activities, and observed that-control-room operators were following procedures and adequately documenting their actions. Quality
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Assurance (QA) personnel observed these activities on normal and.back shift hours.
2.3 Operations Summary-Reactor and plant start-up following the repair to the
"D" RCP, was per-formed in a safe, controlled manner. Operations were conducted in a safe, controlled manner for both the operating mode, transition periods, and during cold shutdown. The pace of activities was well controlled. Cor-rective_ action was assessed as appropriate.
3.0 Equipment Operability Review - Maintenance / Surveillance 3.1 Criteria / Scope of Review The inspectors reviewed selected activities (listed in_ Attachment 1) to verify' proper implementation of the applicable portions of the mainten-ance and surveillance programs.
The inspector used the general criteria listed under the plant operations section of this report.
A more detailed review of equipment operability is addressed below.
3.2
"D" Reactor Coolant Pump (RCP) Seal Repair The licensee accomplished a complete seal overhaul for the "D" RCP during the December 1988 outage. This was the second time during Cycle 7 that the seal leak-off from the "D" RCP increased to the point where the seal package required extensive overhaul which necessitated plant shutdown and cooldown. The first was in September 1988 after approximately one month of operation. During that outage, it was discovered that damage had occurred to the 0-ring that provides a seal between the pump shaft and the No. I seal runner assembly. Damage to this 0-ring results in excessive leakage that essentially by passes the normal flow path of seal injection through the No. I seal faces.
The No. I seal normally passes enough flow (3-5 gallons) to provide a pressure drop from the normal injection-pressure of 2250 psig to the approximately 30 psig of the
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make-up tank. When flow increases significantly above 5 gpm (7-8 gpm as was the case for both the September and December 1988 outages), the functioning of either the seal faces or the 0-ring is in question and damage to the seal could occur.
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The licensee concluded during the September.1988 outage that a probable cause for the 0-ring failure was improper installation.
The installation process is difficult.
Cxtra care and precautions were taken to ensure proper 0-ring installation.
During the December 1988 outage, mis-alignment of the pump to motor via the connecting spool piece was suspected. Therefore, the licensee took extra precautions to ensure proper alignment.
Procedure changes were made to Corrective Maintenance Procedure (CMP) 1401-1.17to. accomplish additional measurements and use a different alignment method. 'The lic-ensee concluded, along.with their prime engineering contractor (Westing-house) and an additional engineering contractor (MPR), that the final alignment of the pump to motor was accomplished as carefully as possible.
The inspectors were briefed by the licensee on two occasions; once during the repair process on findings and once after the pump was successfully tested. 'The NRC staff concluded that the licensee made a well thought out effort to effect a proper pump / motor alignment of the "D" RCP. The process was. characterized as involving good planning, adequate contractor
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engineering support, and a deliberate disassembly / reassembly process.
The two contractors _are preparing final reports and will make recommen-dations for additional monitoring of pump parameters d6 ing continuing operations.
The inspector concluded that the entire repair evolution was conducted in an adequate, safe manner. At the conclusion of the inspection period, the pump was running with normal leak-off flow and vibration, albeit with slightly higher vibration than the other RCPs.
3.3 Emeroency Diesel Generator Annual Maintenance The jacket coolant recirculation pump for emergency diesel EG-Y-1A was replaced on December 13, 1988, during the annual surveillance.
Following completion of the required testing, the diesel was declared operable.
However, during retesting of the diesel and prior to performing a similar surveillance on the second diesel, the jacket coolant recirculation pump tripped.
The root cause was determined to be the new motor and the existing thermal overload relay not being compatible.
Installation of a different thermal overload relay corrected the problem.
Inspector follow-up indicated that the new pump motor name plate was not compared by maintenance with the old one prior to installation.
The in-spector noted that the vendor supplied the new pump under the original specification and provided the necessary Certificate of Compliance but did not identify the discrepancy on the name plate. The inspector asked why the entire process from procurement through acceptance testing did not identify this discrepancy.
The licensee Quality Assurance (QA) or-ganization planned to review this, along with other similar events, for
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potential programmatic weakness.
The inspector will assess the QA recom-mendations upon completion of this review.
This item remains unresolved pending the review of the licensee's evaluation (289/88-29-01).
3.4 Job Ticket Closecut The inspector reviewed twelve job tickets that were closed out following completion of the work activity during the early phase of the 7R refuel-ing outage.
It was noted that work package review by the maintenance supervisor / foreman was accomplished promptly, assuring readiness to per-form post-maintenance testing.
However, in many cases, the actual per-formance of post-maintenance testing, as well as the final closeout of the job ticket, took a considerable amount of time.
In some cases, it took as long as five months. The problem was recognized by the mainten-ance management, who planned to implement corrective actions to stream-line the closecut process. The inspector noted that most of the job tickets were categorized as minor maintenance and did not have any sig-nificant operability issue; however, the effectiveness of the corrective actions will be assessed in the future.
This item remains unresolved pending review of the results of the actions initiated by the licensee (289/88-29-02).
3.5 plant Materiel Condition During plant walkdowns through the river water pump house, emergency diesel rooms, as well as intermediate building, several minor material deficiencies were identified similar to ones identified in the past.
The deficiencies were communicated to licensee management.
In order to correct the problem, the licensee developed a formal plant walkdown pro-
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cedure involving plant operations and maintenance personnel.
Inspector review of this procedure concluded that it was adequate. The procedure is to be fully implemented in January 1989.
The inspectors will assess its effectiveness in subsequent inspections.
3.6 (Closed) Violation (289/88-18-03): Pressurizer Relief Valve Tailpipe Support Loose Bolts The violation involved torquing the pressurizer relief valve tailpipe support bolts under a completed job ticket. The change in scope of work was not reviewed and approved prior to implementation as required by Technical Specification Section 6.8.2, as well as the licensee's Correc-tive Maintenance Procedure (CMP) 1407-1, Revision 32. The licensee's response, C311-88-2155, dated November 14, 1988, adequately addressed the corrective actions: the licensee reviewed the job ticket procedure and counselled the Plant Materiel personnel who generate job tickets.
Following review of the licensee actions, the inspector determined that the corrective actions implemented were satisfactory and timely. This item is closed.
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3.7 Equipment Operability Summary Generally, maintenance and surveillance were accomplished in a safe man-ner. Two unresolved items were identified: job ticket closeout delays (88-29-02) and replacement-in-kind parts (88-29-01).
4.0 Engineering Support 4.1 NRC Information Notice 88-74 On September 14, 1988, the NRC staff issued Information Notice (IN) 88-74 to address potential inadequate performance of Emergency Core Cooling System (ECCS) pumps during the recirculation phase following a loss of coolant accident (LOCA). The IN 88-74 problem is that, during the re-circulation phase, the low pressure injection (LPI) pumps that provide flow to the high pressure injection (HPI) pumps may not establish the required HPI net positive suction head (NPSH) due to flow restrictions in the.line. Additionally, the potential existed for partially closed valves in the LPI discharge.
The licensee evaluated the concerns in IN 88-74 as not applicable to-TMI-1 for the following reasons. The line connecting the LPI discharge
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to the HPI suction has low flow resistance (approximately 10 psig).
Suction pressure at the HPI pumps would be approximately 100 psig at 550 gpm and the NPSH required is approximately 12.psig. Therefore, the NPSH margin is very large.
It also was determined during.this evaluation that strainers installed in each line could potentially clog and result in reduced flow'in these lines. The licensee further concluded that the reactor building sump strainer, a one eighth inch mesh, would provide adequate water cleaning to ensure continued operation of the HPI pumps in the recirculation mode. The licensee is presently evaluating removal of these strainers as it' appears that they are not required.
The licensee also concluded that no power-operated valves in the line would be' lost due to single failure of a power supply. The normally shut DH-V-7A/B motor-operated valves are opened in the recirculation mode of operation and are powered by separate class 1E power supplies. Also, no power-operated, cross-connect valves in the LPI discharge are subject to the single failure criteria discussed in the IN 88-74 situation. At TMI-1, flow is controlled by manually throttling two valves, DH-V-19A/B, in the discharge of each LPI pump. Additionally, the LPI pumps do not provide flow to the suction of the building spray pumps; therefore, the concern for flow in this situation does not exist at TMI-1.
The inspector reviewed the licensee evaluhtion and system schematics, and concluded that the licensee was correct.
The licensee's evaluation was thorough and adequately addressed all concerns identified in IN 88-74.
Discovery that the previously discussed strainers in the system may not be required and could possibly be removed was considered a positive action on the licensee's part. The inspector had no safety concerns.
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4 4.2 Licensee Letter Concerning RC-V-28 Licensee letter dated October 25, 1988 responded to NRC Inspection Report
.(IR) 50-289/88-05 concerning RC-V-28 environmental qualification. IR
'50-289/88-05 questioned the time required for licensee engineering per-sonnel to respond to a concern that RC-V-28 was not environmentally.
i qualified.
s-The licensee and NRC site staff met on October 21, 1988, to discuss the method by which the licensee corporate engineering staff processes safety concerns. The inspector's concern was that, since a seven-day time clock existed per Technical Specification (TS) 3.1.13, if a positive deter-mination of RC-V-28 operability could not be made in seven days, then the action statement (de-energization of the valve operator) should have been taken.
In this case, a communication lapse occurred between the site, corporate engineering, and the inspector. The licensee had deter-mined that RC-V-28 was operable, but they were delayed in providing a written justification to the inspector. As a result of the Detnaer 21, 1988 meeting, the inspector concluded that the licensee had addressed the RC-V-28 equipment qualification (EQ) issue in a timely manner. The inspector had no other safety concerns.
5.0 Management Meetings The inspectors discussed the inspection scope and findings with licensee man-agement weekly and at a final meeting on January 18, 1989. Those personnel marked by an asterisk in paragraph 1.3 were present at the final management meeting.
The inspection results, as discussed at the meeting, are summarized in the x -
cover page of the inspection report.
Licensee representatives did not indi-cate that any of the subjects discussed contained proprietary or safeguards information.
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d ATTACHMENT 1 NRC INSPECTION REPORT NO.- 50-289/88-29 ACTIVITIES REVIEWED Plant Operations Control room operations during regular and back shift hours, including fre-
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quent observation of activities in progress and periodic review of selected-sections of the shift foreman's log and control room operator's log and selected sections of other control room daily logs.
Areas outside the control room.
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Selected licensee planning meetings.
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Start-up and shutdown related to RC-P-ID outage.
During this inspection period, the inspectors conducted direct inspections during
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Day /Date Time Sunday, December 18, 1988 12:00 p.m. - 2:30 a.m.
Saturday, December 24, 1988 9:30 a.m. - 11:30 a.m.
Monday, December 26, 1988 7:30 a.m. - 9:30 a.m.
Monday, January 2, 1989 10:00 a.m. - 12:00 Noon Saturday, January 7, 1989 7:30 a.m. - 10:30 a.m.
Maintenance / Surveillance
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Reactor coolant pump ID seal repair.
Emergency Diesel Generator EG-Y-1A/B annual maintenance.
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Reactor Coolant System (RCS) Leak Rate The inspector selectively reviewed RCS leak rate data for the past inspection period. The inspector independently calculated certain RCS leak rate data reviewed using licensee input data and a generic NRC " BASIC" computer program "RCSLK9" as specified in NUREG 1107.
Licensee (L) and NRC (N) data are tabulated below.
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TABLE RCS LEAK RATE DATA All Values GPM DATE/ TIME TNUREG1107)
CORRECTED DURATION L
N N
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'O g
g g
U 01/01/89.
0.1307 0.13 0.01 0.11 0.1083 7:53 a.m.
2 Hours 01/01/89 0.1006 0.10-0.01 0.09 0.0876 3:49 p.m.
2 Hours-01/04/89 0.0604 0.06-0.06 0.04 0.0443 12:36 a.m.
2 Hours 01/04/89-0.0351-0.04-0.16-0.06-0.0450 4:07 p.m.
2 Hours 01/08/89 0.0530 0.06-0.03 0.07 0.0666 12:57 a.m.
2 Hours 01/08/89 0.1637 0.16 0.07 0.17 0.1696 8:30 a.m.
2 Hours 01/10/89 0.0904 0.09-0.09 0.01 0.0180 5:15 p.m.
2 Hours G = Identified gross leakage U = Unidentified leakage L = Licensee calculated N - NRC calculated Columns 2 and 3, 5 and 6 correlate + 0.2 gpm in accordance with NUREG 1107. N u is corrected by adding 0.1044 gpm to the NUREG 1107 N due to total purge flow u
through the No. 3 seal from RCP's..
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