IR 05000289/1989025
| ML18009A990 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/18/1990 |
| From: | Collins E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18009A989 | List: |
| References | |
| 50-289-89-25, NUDOCS 9002010372 | |
| Download: ML18009A990 (15) | |
Text
U ~
S.
NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.
License No.
Licensee:
50"289/89"25 DPR"50 GPU Nuclear Cor oration P. 0.
Box 480 Middletown Penns lvania 17057 Facility:
Location:
Dates:
Inspectors:
Three Mile Island Nuclear Station Unit 1 Middletown Penns 1vania November
1989 December
1989 R. Brady, Resident Inspector, TMI D. Johnson, Resident Inspector, TMI T. Moslak, Resident Insp=-ctor, TMI F. Young, Senior Resident Inspector, TMI E.
E. Collins, Acting Chief Reactor Projects Section No.
4B Division of Reactor Projects Ins ection Summar
Ins ection Re ort No. 50-289/89-25 for November
1989 December
1989 Areas Reviewed:
The NRC staff conducted routine safety inspections of power operations activities.
The inspectors reviewed plant operations, maintenance/
surveillance and engineering/technical support activities as they related to plant safety.
Specific items reviewed included:
reactor trip of November 29, cold weather preparations, engineered safety features (ESF) walkdown of the emergency feedwater system, troubleshooting on the Electro-Hydraulic Control (EHC) system, and licensee action on previous inspection findings.
Results:
No unresolved items were identified.
Operators response to the reactor trip was good and the transient was well controlled.
Good maintenance and engineering support were noted during the shutdown period.
Plant operations were conducted in a safe manne TABLE OF CONTENTS 1.0 Introduction and Overview
.
1.1 Licensee Activities.
1.2 NRC Activities
.
1.3 Persons Contacted.
2.0 Plant Operations.
2.1 Facility Inspection (NIP 71707).
2.2 Reactor Trip, November 29, 1989.
2.3 Power Coastdown 2.4 Cold Weather Preparation (NIP 71714)
2.5 Operations Summary
.
3.0 Equipment Operability
.
3.1 Survei 1 1 a;;ce Observations (NIP 61726).
3.2 Maintenance Observations (NIP 62703)
3.3 ESF Walkdown (Emergency Feedwater)
(NIP 71710)
3.4 Equipment Operability Summary.
4.0 Electro-Hydraulic Control System Troubleshooting.
5.0 Licensee Action on Previous Inspection Findings (NIP 92703).
5. 1 (Cl osed)
Unresolved Item (50-289/88-06-01)
5. 2 (Cl osed)
Unreso1 ved Item (50-289/88-13-01)
5.3 (Closed)
Unresolved Item (50-289/88-13-02)
5.4 (Closed)
Unresolved Item (50-289/88-17-04)
5. 5 (Cl osed)
Unreso1 ved Item (50-289/88-18-01)
6.0 Management Meeting (NIP 30703).
~Pa e
1
2
2
4
4,
5
6
7
8
9
DETAILS 1.0 Introduction and Overview 1.1 Licensee Activities The unit was at 100% power for a majority of this report period.
At 8:06 a.m.
on November 29, 1989, the plant tripped from 100%.
The unit returned to full power operations on December 1,
1989 and remained there for the remainder of the inspection period.
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 of applicable documents.
In'spections were accomplished on both normal and back shift hours.
NRC staff inspections are generally conducted in accordance with NRC Inspection Procedures (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:
~Da /Date Wednesday, 11/22/89 Thursday, ll/23/89 Friday, 12/1/89 Monday, 12/18/89 Time 6:00 pm 12:00 mid 12:00 mid 2:00 am 12:00 mid 6:30 am 3:00 am
-
6:00 am 1.3 Persons Contacted R. Bensel, Engineer, Plant Engineering
- G. Broughton, Operations/Maintenance Director
- H. Hukill, Vice President and Director, TMI-1 B. Knight, TMI-1 Licensing M. Nelson, Manager, Safety Review M. Ross, Plant Operations Engineer M. Schaeffer, Engineer, Plant Engineering H. Shipman, TMI-1 Operations
- D. Shovlin, Plant Material Director C.
Smyth, Manager, Licensing
"D. Hassler, TMI-1 Licensing
" Denotes attendance at final exit meeting (see Section 7.0)
2.0 Plant 0 erations 2.1 Facilit 'ns ection The resident inspectors routinely inspected the facility to determine 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; plant status from a maintenance/modification viewpoint, including plant housekeeping and fire protection measures;
control of ongoing and special evolutions, including control room personnel awareness of these evolutions; control of documents, including log keeping practices; implementation of radiological controls; and, implementation of the security plan, including access control, boundary integrity, and badging practices.
In general, the inspector determined that the licensee, from 'a housekeeping and fire protection perspective, was maintaining the plant in good condition.
Overall, management attention toward plant safety continued to be noted.
2.2
~R*
2 At 8:06 am on November 29, 1989, the reactor protection system (RPS)
provided an automatic shutdown (trip) of the reactor.
Plant conditions at the time of the trip were as follows:
-Unit was at 100K power-Integrated Control System (ICS) was in automatic
-"B" Emergency Diesel was out of service (OOS) for maintenance-RPS monthly surveillance was in progress-T(AVG) reduced to 578 from a Nominal 579 degrees F
The licensee determined the cause to be a failure in the speed error circuit of the electro-hydraulic control (EHC) System.
This caused fast closure of the turbine intermediate valves that supply steam to the turbine resulting in an inadvertent load rejection.
In response to the load rejection, primary plant temperature and pressure increased.
The reactor tripped on high pressur iC 4 ~
~ ~
The inspector observed the post-trip recovery which was performed in a calm, deliberate fashion.
Proper use of the abnormal operating procedures and good control of the evolution and control room environment were observed.
Plant response was normal during recovery with the exception of the following:
Main Steam Safety Valve MS-V-21A failed to reseat properly.
The operating crew lowered steam generator pressure and the valve reseated.
NI-2 source range nuclear instrumentation had spurious readings, During the shutdown period the licensee cleaned the detector connections and NI-2 appeared to be functioning normally.
The inspector also reviewed the shift techni.cal advisor's shutdown margin (SDM) determination and verified that SDM was adequate as required by technical specifications.
The licensee conducted a post-trip review in accordance with AP-1063
"Reactor Trip Review Process".
The purpose of this review was to conduct technical reviews or analysis of plant performance.
The inspector attended this meeting and identified no unacceptable conditions.
The inspector observed the reactor startup on December 1,
1989.
The inspector reviewed the following:
-Operating Procedure (OP)
1102-2 "Plant Startup"
-OP 1103-8 "Approach to Criticality"
-Estimated critical rod position calculation-Critical boron calculation The inspector verified completion of the licensee's pre-critical checkoffs.
A nuclear engineer was present during the startup, and observed the withdrawal of the Axial Power Shaping Rods (APSR).
APSR withdrawal was completed after criticality was achieved, and was conducted IAW approved procedures.
Criticality was achieved at 2:55 am and the turbine was on line at 6:00 am.
The unit returned to 100%
power at 6:30 pm on December 1,
1989.
A member of the licensee's gA organization also audited the startup.
2.3 Power Coastdown During this inspection period, the fuel was depleted to the point where the core could not support full power output.
In order to extend the full power operation, the licensee performed two abnormal reactivity addition maneuvers; an average core temperature (Tave)
reduction of five degrees, and withdrawal of the axial power shaping rods (APSR).
The combined reactivity. addition of these maneuvers extended 100% power operations approximately one wee ~;
s
~
During the startup from the reactor trip of November 29, the licensee withdrew the APSRs.
Due to potential effects on core power distribution and fuel reliability, the licensee performed a safety evaluation and prepared a temporary change to OP 1102-4 "Power Operations".
The inspector reviewed the safety evaluation and procedure change.
Based on core performance exhibited during this cycle, previous transients and a calculation of the axial xenon stability index done by Babcock 5 Wilcox, the licensee concluded the core would not experience diverge0t axial xenon osci llations.
The inspector had no safety concerns with licensee action in this area.
2.4 Cold Weather Pre arations The inspector reviewed the licensee's maintenance and surveillance activities to ensure the licensee has implemented protective measures for extreme cold weather.
The inspector verified the licensee has inspected systems susceptible to freezing, by verifying the operability of heat tracing systems, use of space heaters, and the proper setting of thermostats.
The inspector reviewed Operations Surveillance Procedure OPS-285
"OPS Winterization Checks" for scope and to verify its completion.
Procedure OPS-285 is an extensive annual'heck of out-building heating/ventilating systems operations, including thermostat settings, heat trace operability and insulation conditions, and anti-freeze for various diesel engine cooling systems and sprinkler systems.'uring review of OPS-285, the inspector noted that licensee personnel had identified a number of corrective actions required concerning heat tracing and heaters and had identified several items that needed to be added to the cold weather preparations list.
Based on review of the completed procedure, the inspector concluded that the licensee's overall efforts in freeze protection were adequately implemented and controlled.
The inspector had no other questions in this area.
2.5 0 erations Summar 3.0 E ui Operations continue to be conducted in a safe manner.
Operator response to the reactor trip was prompt, and proper control of the plant recovery operations was noted.
The reactor startup was performed in a safe manner.
ment 0 erabilit 3.1 Surveillance Observations On a sampling basis, the inspector selected a surveillance and maintenance activity to ensure that specific programmatic elements described below were being met.
Details of thi s review are documented in the following section The inspector observed performance of the following surveillance tests to determine that: the tests conformed to technical specific-ation (TS) requirements; administrative approvals and tagouts were obtained before initiating the surveillance; testing was accomplished by qualified personnel in accordance with an approved procedure; test instrumentation was calibrated; limiting conditions for operations were met; test data was accurate and complete; removal and restoration of the affected components were properly accomplished; test results met TS and procedural requirements; deficiencies noted were reviewed and appropriately resolved; and the surveillance was completed at the required frequency.
This observation included:
-SP 1301-4. 1 Reactor Protection System-SP 1302-5.30 Diesel Generator Protective Relaying-SP 1303-4.15 Radiation Monitoring System Monthly (Area Channels)
3.2 Maintenance Observations The inspector observed portions of selected maintenance activities to determine that the work was conducted in accordance with approved procedures, regulatory guides, technical specifications, and industry codes or standards.
The following items were considered during this review: limiting conditions for operation were met while components or systems were removed from service; required administrative approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and gC hold points were established where required; functional testing was performed prior to declaring the particular component(s)
operable; activities were accomplished by qualified personnel; radiological controls were implemented; fire protection controls were implemented; and the equipment was verified to be properly returned to service.
These observations included:
SP 1301-8.2 Diesel Generator Annual Inspection Trouble-shooting of Electro-Hydraulic Control system 3.3 En ineered Safet Features Walk Down Emer enc Feedwater S stem The inspector performed a comprehensive verification of the Emergency Feedwater (EF) system operability.
The inspector reviewed the systems design and operational requirements as specified in the Final Safety Analysis Report and Technical Specifications.
The inspector reviewed the following licensee documents:
~
~
O
~
T ~ ~
TT Tl
~ 4 TT ~
'IT'
~
~
T
~ T
Operating Procedures (OP 1106-6)
DWG. D302-082 (Rev.
13)
Emergency Feedwater The inspector verified that enclosure I (Startup Valve Checklist) of OP 1106-6 reflected the as built drawing and that the checklist was comprehensive.
The inspector performed a field walkdown of all accessible portions of the EF system.
The inspector verified that EF system valve positions were in accordance with the licensee's valve lineup procedure and as-built drawing.
The inspector also reviewed the licensee surveillance procedures to ensure requirements outlined in TS were being properly implemented and at the required frequency.
3.4 E ui ment 0 erabi lit Summar Main.enance activities were carried out in a safe manner.
System walkdown of the EF system determined that the system was properly maintained and could perform its designated function if required.
4.0 Electro-H draulic Control S stem EHC Troubleshootin In response to an action item from the post trip review, plant engineering was assigned to investigate the root cause of the reactor trip of November 29, 1989.
During review of the transient output, it was noted that there was a prompt decrease in generated Megawatts.
This load shed could not have been caused by an external control source, such as the Integrated Control System (ICS), It was concluded that the load shed was caused by fast closure of the turbine contr'ol or intercept valves, therefore troubleshooting efforts concentrated on those circuits within the EHC system that caused fast closure of these valves.
A consultant from GE was utilized to aid in the troublshooting and evaluation process.
During the troubleshooting work, a loose shield wire on the input from the primary speed sensor was observed by the licensee.
To evaluate the potential for the loose wire to cause the event, an 1800 RPM signal was simulated and the speed signal shield wire was moved.
This caused a fast closure signal to be generated, indicating that the loose shield wire could have been the root cause of the turbine trip.
The EHC cabinet was checked for additional loose connections and none were identified.
The results of the troublshooting effort were documented in a report issued by plant engineering on December 1,
1989.
The report included scope of troubleshooting, additional maintenance performed on the EHC system, and long term corrective actions, including evaluating the
.
feasibility of upgrading the current EHC syste The inspector discussed the results with the licensee and reviewed the report.
The inspector concluded the troubleshooting was comprehensive.
The inspector also noted timely response by plant engineering, in providing support to plant operations.
5.0 Licensee Action on Previous Ins ection Findin s
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.
5.1 Closed Unresolved Item 50-289/88-06-01 MSSV Rin Settin Evaluation This item was opened during a review of licensee response to NRC's Information Notice 86-05 concerning Main Steam Safety Valves (MSSV),
where improper ring settings could result in insufficient lift capacity.
The licensee's MSSV ring settings for the Dresser 3707-RA valves vary from -3 to -9 on the lower ring and from +104 to +248 on the upper ring.
The numbers denote the number of notches from the valve seat plane reference point.
Dresser recommends ring settings of -8 (lower ring) and +160 (upper ring) for model 3707-RA valves.
Based on, the vendor recomendations, the licensee considered their ring settings acceptable.
The inspector expressed concern that the acceptance of the current ring settings was somewhat arbitrary since there was no available test data to correlate with actual valve settings.
This item was unresolved pending the licensee obtaining test results or other data that would justify their MSSV ring settings.
The licensee obtained full flow test data for the Dresser model 3707-RA valves conducted at the Wyle Laboratories for the Davis-Besse plant;-'his test determined the full rated lift pressure and the blowdown capacity.
The lower ring settings varied from -36 to -7 notches; the upper ring settings were +113 to +420.
These valves all exhibited proper lift pressures and met the 3-9% blowdown capacity specification.
The MSSV ring settings at TMI-1 are within these ranges.
Based on review of this data and discussion with licensee engineers, this item is closed.
5.2 Closed Unresolved Item 50-289/88-13-01 Procedure Non-Adherence on Tem orar Loss of Deca Heat Removal-Licensee Voluntar Re ort This item concerned a temporary loss'f Decay heat removal (DHR) on June 27, 1989.
The licensee committed to provide the NRC with a voluntary report addressing this even The inspector reviewed the voluntary report dated October 11, 1989.
The licensee had concluded that this event was not reportable.
This conclusion was based on the fact that decay removal capability was maintained.
The Borated Water Storage Tank (BWST) and both DHR pumps were available for injection IAW Emeregency Procedure 1202-35 "Loss of Decay Heat Removal System".
The plant conditions met requirements of Technical Specification 3.4 "Decay Heat Removal Capability".
The event was caused by personnel error during the performance of surveillance procedure SP-1302-5.8
"High and Low Pressure Injection Analog Channels".
The technician performed steps out of sequence which caused DH-V-2 (isolation valve from."B" loop hot leg to decay heat pump suction) to close.
To prevent reoccurrence of this event, the licensee revised SP 1302-5.8.
This change removes power to decay heat suction valves DH-V-1 and DH-V-2 while performing the surveillance test.
The inspector reviewed the procedure and had no further questions.
Based on the above licensee action, this item is closed.
5.3 Closed Unresolved Item 50-289/88-13-02 A
roximatel 300 allons of Reactor Coolant S stem RCS Water S illed in Containment Due to Poor Work Control On June 26, 1988, during the 7R refueling outage, approximately 300 gallons leaked from the RCS into the reactor building.
The spill occurred when operations personnel pumped water from the pressurizer to the "C" reactor coolant bleed tank with partially disassembled valves in the flow path.
The licensee concluded the event was caused by poor communications between the maintenance, operations, and radiological controls personnel and poor knowledge and control of plant status by on-shift operations personnel.
The licensee made the following programmatic changes to address the problem:
Revised AP 1070 "TMI-1 Mai'ntenance Plan" to require all work, except for emergencies, be scheduled; AP-1070 also requires that the performer of the work to notify the shift supervisor/shift foreman of the work to be done.
This notification shall be done at the beginning of each shift unless the task is continued from the previous shif To support the 8R outage work plan and schedule, the position of shift outage advisor (SOA) will be established.
The SOA monitors implementation of the outage integrated schedule, provides guidance for the prevention and resolution of conflicts and delays, and advises the shift superviso The operations staff has developed a shift supervisor (SS) turnover outage attachment check list.
This checklist is to be completed and reviewed by the SS during shift turnover.
The checklist insures operability of reactor coolant system (RCS) fill and drain paths and requires the integrity of these paths be verified; provides guidance on RCS and OTSG level control; and checks anticipated RCS/OTSG inventory changes.
Based on the above actions this item is closed.
5.4 Closed Unresolved Item 50-289/88-17-04 Reactor Bui ldin S ra Pum B Rela Calibration-Lebanon Rela Grou Work Schedulin During the 7R Refueling Outage, the NRC outage inspection team noted that relay calibration activities performed by the Lebanon relay organization, which is part of the GPU Nuclear/Metropolitan Edison organization, were not coordinated with the TMI-1 Division scheduling group.
The reason this work and lack of schedule coordination was of concern is that QA/QC notification is required for the subject calibrations by the preventive maintenance procedure.
The lack of scheduling information and lack of notification to the Quality Assurance Department that this work was in progress was also identified by the GPU Nuclear Quality Assurance Department in Quality Deficiency Report No. HRH-039-88.
On August 26, 1988, the TMI-1 Planning and Scheduling Manager issued a memorandum which specified that protective relay calibration will be added to the electrical maintenance work schedule during normal plant operation and the relay department work tasks could be shown as a separate section on the weekly work schedule during outages.
This action is intended to provide the Quality Assurance organization appropriate notification of which relay calibration work is scheduled on a shift-by-shift basis.
Additional internal GPU Nuclear memoranda notified appropriate individuals of the importance of showing work by Lebanon Relay on the weekly work schedule.
Quality Deficiency Report No. HRH-039-88 was closed out by GPU Nuclear on September 23, 1988.
During the annual preventative maintenance outage of the "A" emergency diesel generator in December 1989, the NRC staff verified that relay calibration efforts by the Lebanon relay group were appropriately scheduled on the weekly work schedule under the electrical maintenance work schedule.
Based on the above action, this item is closed.
5.5 Closed Unresolved Item 50-289/88-18-01 Licensee to Define Event Review Process and U
rade Administrative Procedures This itgm concerns the licensee's criteria that is to be used by shift supervision to determine when abnormal plant conditions merit a formal multi-disciplinary review.
Sufficient guidance did not
exist to ensure low threshold events (events which fall below other reporting requirements)
received proper independent evaluation and review.
The inspector reviewed Revision 30 to AP 1012 "Shift Relief and Log Entries" dated January 4,
1989.
The procedure requires the shift foreman to initiate a multi-discipline review, per requirements of AP 1029 for. significant plant abnormalities which fall below the criteria for other reporting requirements.
The administrative procedure also provides extra guidance, in the form of examples, to define significant plant abnormalities which require multi-discipline review.
The inspector reviewed the following "Multi-discipline review of Level II Log Entries" generated IAW AP 1012 and AP 1029:
-89-01 Ladder wired to RPS "C" conduit-89-02 - Inadvertent PORV lifting caused by technician terror-89-03 -
RPS channel
"8" trip due to failed component and smart Automatic selector switch (SASS) mismatch Each received review by the plant review group (PRG)
and results were documented in PRG minutes.
The inspector had no questions, on the licensee's program and implementation.
Based on this review, the item is closed.
6.0 Mana ement Meetin The inspectors discussed the inspection scope and findings with licensee management weekly and at a final meeting on December 19, 1989.
Those personnel marked by an asterisk in paragraph 1.3 were present at the final management meeting.