IR 05000289/1989012
| ML20246C683 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/16/1989 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20246C680 | List: |
| References | |
| 50-289-89-12, NUDOCS 8908250078 | |
| Download: ML20246C683 (12) | |
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U.S. NUCLEAR. REGULATORY. COMMISSION-
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REGION.I.
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- Docket / Report No. 50-289/89-12 License: DPR-50-Licensee:
GPU Nuclear Corporation P. 0. Box 480
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Middletown, Pennsylvania 17057 Facility:
-Three Mile Island Nuclear Station, Unit 1 Location:
Middletown, Pennsylvania-Dates:
June 10 - July 14,1989 Inspectors:
D. Johnson, Resident Inspector, TMI
.T. Moslak, Resident Inspector, TMI C. Woodard, Reactor Engineer.
F. Young, Senior Resident Inspector,. TMI
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/ 67 Approved by:
.C wgill, Ch'ief Date ctor Projects Section No. 48 ivision of Reactor Projects Inspection Summary:
Areas Reviewed: The~NRC staff' conducted routine safety inspections of power operations activities..The inspectors reviewed plant operations and maintenance / surveillance as they related to safety. Specific items reviewed included a control rod drive indication problem,L the diesel generator fuel l oil quality assurance program, RM-A-2 operation and licensee action on previous inspection findings.
.In' addition, the licensee 10 CFR 50.59 annual report was reviewed for accuracy.
- Results: Plant operations were conducted in a safe manner. Review of the licensee's annual 10 CFR 50.59 report determined' that the licensee had
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accurately characterized changes to the facility. With respect to CRDM position indicating problem, the inspector noted the licensee corrective action was adequate. :The. inspector did note a~ minor weakness in the operator under-standing of the computer program that calculated core flux and tilt.
Licensee corrective ~ actions were initiated to address the weakness.
Licensee action on previous inspection findings was adequate.
8908250078 890816 PDR ADOCK 05000289
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TABLE OF CONTENTS i
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P_ age 11.0 'In'troduction_and' Overview.................
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1.1' Licensee Activities.................
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1.2 NRC Activities'...................
1.3 Persons Contacted...................
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2.0- Plant Operations.....................
2.1-Criteria / Scope of Review (NIP.71707)........
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2.2 ' Control Rod Drive Position Indication Anomaly....
2.3-Operations Summary-.................
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3.0 Equipment; Operability Review - Maintenance / Surveillance
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3.1 ~ Senpe of Review (NIP 62703/61726)
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'3.2 RM-A-2 Operation..................
'3.3 Equipment-Operability Summary
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L 4'.0.J50.59 Report Review (NIP 37700).............
'5.0 Licensee Action on _ Previous-Inspection Findings-(NIP 92703)
-5.1-(Close'd)' Unresolved Item (50-289/85-00-03) Control of-Heavy Loads Design Changes on Lift Lugs and Modifications 6 5 2'(Closed) Unresolved l Item (50-289/83-30-05) Adequacy of
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OJT Records.
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- 5.3 (Closed)' Unresolved Item (50-289/85-25-05)'MSSV-Abnormal Performance.....:...........
6.0. Management Meeting (NIP'30703)..............
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DETAILS-1.0' -Introduction-and' Overview l
f 1.1 Licensee' Activities The -licensee operated the plant at full power during the report
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period. No major plant transients occurred. As of July 14, 1989,-
the TMI reactor was at 100 percent.Fower..
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 personnel, or independent calculation and selective review
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of applicable documents'
Inspections.were accomplished on both
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normal and back shift hours.
NRC staff inspections are generally conducted in accordance with NRC Inspection ProceduresL(NIPS).
These NIPS are noted under the appropriate section in the Table.of Contents to this report.
Back shift inspections were accomplished during the following periods:
Day /Date Time-Wednesday June 14, 1989 7:00 pm - 9:00 pm Saturday June 24, 1989 7:00 am - 1:00 pm Saturday June 24, 1989 9:00 pm - 11:00 pm Thursday July 06,1989 4:00 pm - 5:00 pm Friday July 07,1989 6:00 am - 7:00 am Sunday July 09,1989 9:00 am - 11:00 am During this period, an inspection was performed by the NRC Regional staff to determine the adequacy of the licensee's prograni for the procurement,~ receipt, storage, handling and control of Emergency Diesel Generator fuel oil. The result of this inspection will be documented in a subsequent inspection report.
1.3 Persons Contacted
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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- T. Hawkins, TMI-1, Start Up and Testing
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H. Hukill, Vice President and Director, TMI-1
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- B. Knight, TMI-1 Licensing
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M. Nelson, Manager, Safety Review
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- S. Otto,, TMI-1-Licensing
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M. Ross, Plant Operations Engineer
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- H. Shipman, TMI-1 Operations i
D. Shov11n, Plant Material Director
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P. Snyder, Manager, Plant Material Assessment
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C. Smyth, Manager, Licensing
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- J. Stacey, TMI Security
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- Denotes attendance at final exit meeting (see Section 7.0)
2.0 Plant Operations i
2.1 Criteria / Scope of Review 1. -
The resident inspectors routinely inspected the facility to determine l.
the licensee's compliance with the general operating requirements of.
Section 6 of Technical Specifications (TS) in the following areas:
review of selected plant parameters for abnormal trends;
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plant status from a maintenance / dification viewpoint, l-
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including plant housekeeping and ire protection measures; control of ongoing and special e olutions, including control ~
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room personnel awareness of the te evolutions;
' control of documents, including log keeping practices;
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implementation of radiological ontrols; and,
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implementation of the security plan, including access control,
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boundary integrity, and badging practices.
Specific findings are addressed below.
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2.2 Control Rod Drive Position Indication Anomaly
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On June 24 at 2:39 am, the operators observed that the absolute
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position indicator for Rod 9 in Group 5 fell to 0%.
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at 100%. The out limit light was still lit and the in limit light remained unlit. The control room operators observed other plant parameters such as Tave, Reactor Coolant System pressure, and neutron
error, but there was no change indicated that would have been evidence of a dropped control rod.
The shift technical advisor reviewed the plant computer display for incore flux tilt for the previous shift and a current reading. No significant changes were noted between 11:20 pm on June 23 and 2:44 am on June 24. However, the data for Displays 17 and 18 showed depressed flux in the location of the rod in question (location K-3).
This was based on a comparison of readings at the times noted above, i
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.The shift personnel notified plant engineering personnel who-responded to review the data.
It was obvious to control room personnel from observed plant paramaters that a rod had not dropped but the information from the plant computer displays was not consistent with these observed plant parameters.
The review of this problem by licensee personnel, revealed that the plant computer factors control rod position into the calculation for l
assembly power. Although there is no self powered neutron detector (SPND) in location K-3 (Group 5-Rod 9) a correction to the power in
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that location was made based upon the fact that the rod position system indicated a dropped rod.
Plant Engineering is evaluating this problem and will consider changes to the appropriate emergency procedures to alert operations to the above computer system characteristics as part of the next revision.
The inspectors reviewed this event with licensee personnel and considered that operators responded properly to this event.
Plant Engineering response to assist operations persornel was timely. No safety concerns were generated as a result of this event.
2.3 Operations Summary The licensee continues to exhibit positive control over the various operational maintenance surveillance and other work related activities that were conducted during this period.
In general, the inspector determined that the licensee, from a housekeeping and fire protection perspective, was maintaining the plant in good condition.
3.0 Maintenance / Surveillance 3.1 Criteria / Scope of Review The inspectors reviewed selected activities to verify proper implementation of the applicable portions of the maintenance 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 RM-A-2 Operation During plant tours, the inspectors have noted that the reactor building radiation monitor (RM-A-2) has been operating close to its full range indicating limits for both the particulate and noble gas channels. The detector also monitors the reactor building atmosphere for iodine.
In the past several months, the RM-A-2 particulate channe'l has experienced several failures due to the filter paper drive unit not operating properly. The inspector met with licensee radiological engineering personnel to discuss the problem and possible solutions, and also discuss the corrective action the licensee had taken to correct the above problems.
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4-The paper drive unit was found bylthe licensee to be excessively o
worn, and was repaired. Additionally, the licensee discovered that-it was not clear to all. rad-con technicians, how to install the rolls
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fof filter / sample paper.
In some instances,'the paper was installed backwards, which. interfered with proper paper advance..These problems were corrected, and subsequent observations by the inspe";
have indicated no additional problems.
The_ elevated. levels on the particulate and gas' channels are.still a concern to the inspectors. The licensee is-evaluating several possible actions to enhance the usefulness of this system. Recently, the alert and alarm set points for the particulate and noble gas channels have been increased from previous settings. This has so far prevented a continuous state of the " alert" alarm being locked in for both the particulate.and gas channels of RM-A-2.
Also, as the particulate channel paper drive is now functioning properly, very few
" alert" alarms have been received. The actual' reactor building particulate and noble gas levels have been consistently below the new setpoints for the " alert" level. This has enhanced the usefulness of-this instrumentation to control room operators in determining potential RCS leak problems.
The inspector's discussion with licensee personnel revealed that additional action'to enhance the operation of RM-A-2 is also'being considered.
Various calibration and sensitivity adjustments to the system can be made to account for the high levelt of background activity in the reactor building.
These modifications or system adjustments will be implemented on "as needed" basis.
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The inspector's review of licensee action on the RM-A-2 problems indicated that a proper level of concern had been evident by licensee personnel. Corrective action to restore RM-A-2 to a more reliable status when these problems occurred was timely and adequate.
The inspectors had no other safety concerns at this time on the operation of RM-A-2.
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3.3 Equipment Operability Summary l
Based on inspector's review, maintenance and surveillance activities continue to be conducted safely.
4.0 50.59 Report Review The licensee submitted their annual report pursuant to the requirement of 10 CFR 50.59 in a letter to the NRC on June 29, 1989. This report contained summaries of the 1988 changes to TMI-1 systems and procedures as L;
described in the Safety Analysis Report.
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The inspector reviewed the report for completeness and accuracy'and.
additionally reviewed several of the changes to the facility procedures and systems to verify that the particular change was properly implemented.
During the course of the year, the inspectors had reviewed 2 of 8 safety reviews' for procedure changes and 7 of 23 safety reviews for modifications
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to plant systems.
Inspector findings on these reviews were documented in previous inspection reports during 1988.
.Two additional reviews were performed for plant modific.itions. These were
.the safety evaluations for BA128975 for modifications to letdown' valve MU-V-3 circuit and BA412539 for a modification to the electric driven fire pump control circuit.
The modification to the MU-V-3 control circuit consisted of elimination of the feature for ' automatic closure on a reactor trip. This was done to prevent thermal cycling of the letdown heat exchangers. The modification also added automatic closure signals for the 1500 PSIG and 500 PSIG ESAS actuation. The auto closure was maintained for the 4 PSIG reactor building pressure signal.
The inspector-reviewed safety evaluation (SE) 128975-001 for this modification. The inspector concluded that it was a comprehensive analysis for this change to plant systems, and adequately analyzed potential safety concerns. No additional problems were noted.
The inspector also reviewed SE 412539-001 for a modification to the electrically driven fire pump FS-P-2.
Previously, this pump started on an ESAS' actuation and was loaded onto the respective emergency diesel generator. This feature was removed to prevent excessive unnecessary diesel loading during ESAS actuations.
The pump can be started mar.ually after resetting-the ESAS signals. Additionally, two diesel driven fire pumps are available during the period when FS-P-2 is locked-out. The inspector concluded, based on a review of this evaluation, that no safety concern existed for this modification. The SE was complete and adequately documented licensee implementation of this system modification.
The inspector review of these modification safety evaluations as well as SE's that were previously reviewed, indicated an adequate ongoing licensee program for evaluating facility changes.
5.0 Licensee Action on previous Inspection Findings The inspector reviewed licensee action on previous inspection findings to ensure that the licensee took appropriate action in response to the findings' or by self-initiative and that the licensee's action was timely.
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5.1 (Closed) Unresolved ' Item (289/85-SC-03) Licensee Letter 84-2147 of 6-28-84 Control of Heavy Loads; Design Changes on Lift Lugs and Modifications The remaining action on this item was to verify completion of modifications to the A-B fue1~ pool gate and the fuel pool cask pit gate which were recommended by NUREG-0612. The inspector reviewed licensee calculation C1101-252-5310-002 which verified that the
' lifting lugs on the' gate that separates the A & B fuel pools were adequate. A safety factor of 8.4 was calculated, which is greater than that recommended by NUREG-0612, No problems were noted.
The inspector also reviewed Safety' Evaluation (SE) 128057-001 for a modification to the cask pit gate lifting device. This modification provided two drilled holes in the gate for attachment of the appropriate shackles,- which are controlled by procedures.
This modification also provided a safety factor greater than recommended by NUREG-0612. 'No discrepancies were noted in the above documenta-tion. The above licensee action closes this item.
5.2 (Closed) Unresolved Item (50-289/85-30-05), NRC to Review Adequacy of Records for OJT for the Non-Licensed Area This item concerned the adequacy of licensee administrative systems for tracking the completion of On-The-Job-Training (0JT) for personnel in the maintenance department.
The licensee issued a new standing order.#17 in 1986, but did not implement this system due to the impending transfer to an INPO approved OJT tracking system. This system was implemented in 1987/1988 and has resulted in the completion of a major portion of licensee maintenance personnel being decertified and new OJT records being generated.
The inspector reviewed the DJT records for various licensee personnel and considered the system to be an adequate and comprehensive method for tracking OJT. This' system is now fully implemented at TMI 1.
The inspector had no other questions on this item.
5.3 (Closed) Unresolved Item (50-289/85-25-05): Main Stum Safety Valve (MSSV) Performance This item was opened to document the resolution of problems associated with MSSV performance following a reactor trip and was updated in NRC Inspection Report Nos. 50-289/85-30, 86-13, 87-10 and l
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87-11. This item remained open pending the licensee's evaluation of 1: ?
specific recommendations provided in the B & W Owners Group (B&WOG)
Safety and Performance Improvement Program (SPIP).
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1 Each of the SPIP recommendations was evaluated by the licensee and the actions taken by the licensee were subsequently reviewed by the
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NRC staff.
Following are the results of that review.
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SPIP Recommendation (TR-023 MSS): Inspection of.MSSV Release Nut Cotter Pins
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Based on a review of BWOG TAP Report 12-1151244-0, Davis-Besse(D-B)-
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Reactor _ Trip of 3/2/84,Section II.E.1, the licensee inspected'
release nut cotter pins of.all (eighteen) TMI-l'MSSVs on 5/23/84.
l The release nut cotter pins on' fourteen MSSVs were changed out by the
_ valve manufacturers' (Dresser) field service representative following
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.MSSV tests. 1The cotter pins on the remaining four MSSVs were checked
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If for proper installation at that time and found acceptable.
Subsequently, Surveillance Procedure (SP) 1303-11.3, " Main Steam Safety Valve Testing", was revised (Revision 14) to require installation of new stainless steel cotter pins following all future MSSV tests. This procedure change added a check off to the data sheet _.to' provide verification that the stainless steel cotter pin'was properly installed _ during the valve reassembly. These actions.were taken'to ensure that the cotter pin, which was installed each time a valve was tested, was new and consisted of non-magnetic material.
Based on a review of licensee records and applicable procedures, the.
' inspector concluded that the licensee has implemented the subject
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SPIP Recommendation (TR-024-MSS): Evaluate Anomalous Post-Trip MSSV-
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Performance Based on a review of B&WOG TAP Report #0C1-85-01, Oconee I' reactor trip on 12/2/84,Section III.B.I, the licensee conducted an evalua-
. tion of past MSSV performance to assure reliable' operation.
During cycle 6R, fifteen (15) of eighteen (18) TMI-1 MSSVs were overhauled. One MSSV was replaced and the remaining two valves were tested and their set points verified prior to shutdown for the start of 6R. During startup, following 6R, the fifteen refurbished MSSVs and the one replacement MSSV were tested and set to the required setpoints.
During two reactor trips experienced in operating cycle 7, licensee review of the post trip performance of the MSSVs indicated that during both events...
--The MSSVs lifted at the correct setpoint
--Excessive blowdown resulting in low secondary system pressure did not occur
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--All-MSSVs completely reseated af'ter lifting-
--No..relifting of MSSVs occurred In general, plant response to both trips was normal and.the MSSVs demonstrated. proper post trip control of secondary system pressure.
- As a follow-up to this. SPIP recommendation,- the B&WOG has completed the.following actions:
(1) The B&WOG Availability Committee has completed a study.and published the results and recommendations in the report, " Main Steam Pressure Control Review, dated December 1986.,
(2) A Valve Task Force was formed by the B&WOG to evaluate performance and promulgate recommendations to ensure MSSVs perform reliability.
The. licensee will continue to evaluate the applicability of future
' recommendations provided by these organizations.
Based on a review of licensee records and discussions with licensee representatives, the inspector concluded that the licensee has
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l implemented SPIP Recommendation No. TR-024-MSS.
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SPIP Recommendation (TR-108 MSS):
Evaluate Using The
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- Maximum' Allowable Set Pressure For the Lowest Set MSSV-
- The licensee evaluated this recommendation and concluded that since the present' set point (1040 psi) for the' smallest safeties is only 10 psi lower than the maximum allowable set point (1050 psi), the 10. psi change is small in comparison to the variations in the MSSV setpoints resulting-from normal calibration tolerance and set point drift.
Additionally, operator response to MSSV relift has been to take manual control of header pressure by lowering the Turbine Bypass Valve set point. Such manual intervention can be performed quickly and' easily.- Accordingly, the licensee concluded that changing MSSV set point was not warranted.
Licensee experience gained during turbine trips which occurred on September 16, 1988 and on October 30, 1988, further substantiated the conclusion that early and/or H
repetitive lifting of the MSSVs can be averted with preventative methods.
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Through interviews with licensee representatives, and review of licensee procedures / records, the inspector determined that the licensee's evaluation of this recommendation was adequate.
SPIP Recommendation (TR-174 MSS): Evaluate Response of Turbine Bypass
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y The purpose of this recommendation was'to. assure a' rapid. Turbine Bypass Valve (TBV)! response. The: recommendation called for improving the response'of the TBV's by. assuring a strokeLtime of 3 seconds orf
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-less (with A +' O.5 second tolerance permitted) and establishing-surveillanceLand maintenance criteria to maintain a rapid response time. To implement this recommendation,. stroke tests were performed on the TBV during a' plant shutdown period between'10/17/88 and 10/19/88.
Results of these tests: indicate that all TBVs met the acceptance criteria..In keeping with the.B&WOG recommendations, the licensee established a preventative maintenance; testing schedule 1for the TBV. in which three of six' valves would be stroke tested every refueling outage with all valves being tested within a.three year
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The ac' ions'and. evaluations completed by the licensee are acceptable.
t The inspector had no further questions regarding the implementation
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SPIP Recommendation (TR-223-MSS): B&WOG' Member Utilities Should U' se-
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- the MSSV Maintenance. Pro:edure~ Comparison Matrix Developed By the B&W Valve Task Force as a Guideline To Evaluate Their. MSSV Maintenance
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~ Procedures and/or Program The licensee' reviewed and compared General Maintenar,ce Procedure.
1404-4, " Repair of Main Steam Safety Relief Valves", against the comparison matrix. Subsequently, this procedure was-revised to include more specific guidance on valve reassembly-based on this review. LIn addition to using the revised procedure when performing
. repairs and maintenance to MSSV, the licensee will continue to use guidance provided by the valve technical manual and an on-site-representative of the valve's manufacturer.
The inspector reviewed licensees actions regarding this item and determined that these actions were satisfactory.
SPIP Recommendation (TR-224-MSS):
Utilize Generic MSSV Setpoint
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Testing Guidelines
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The licensee reviewed and compared SP 1303-11.3 to generic guidelines-for setpoint testing of the MSSV's to evaluate the completeness of this procedure. Based on this review, the procedure was subsequently revised to improve its level of detail. This revision (No. 21)
incorporated many of the generic testing guidelines to standardize the'particular methods used for testing to reduce the potential for error.
The inspector determined that these changes significantly improved the procedure clarity.
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" 6.0iManaganent Meeting
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The inspectors: discussed the" inspection scope.'and findings with licensee-management weekly and 'atJa fina1 meeting' on July 17,.l1989. _ Those -
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personnel; marked by an' asterisk' in paragraph 1. 3 were present at the final-
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management meeting.
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