IR 05000382/1987004
| ML20207S412 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/10/1987 |
| From: | Constable G, Luehman J, Staker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20207S407 | List: |
| References | |
| 50-382-87-04, 50-382-87-4, IEB-86-001, IEB-86-1, IEIN-86-106, NUDOCS 8703190410 | |
| Download: ML20207S412 (9) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-382/87-04 License:
NPF-38 Docket:
50-382 Licensee:
Louisiana Power & Light Company (LP&L)
142 Delaronde Street New Orleans, Louisiana 70174 Facility Name: Waterford Steam Electric Station, Unit 3 Inspection At: Taft, Louisiana Inspection Conducted:
February 1-28, 1987
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Inspectors:
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Grtuehiiian, Senior Resident Inspector Date ~
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p.R.Staker,_ResidentInspector Date
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7[o!C7 Approved:
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C L. Constable, Chief Dhte'
Reactor Project Section C Inspection Summary Inspection Conducted February 1-28, 1987 (Report 50-382/87-04)
Areas Inspected:
Routine, unannounced inspection of:
(1) Plant Status, (2) Licensee Event Report Followup, (3) Followup of Previously Identified Items, (4) Monthly Maintenance,-(5) Montt}1y Surveillance, (6) Routine Operational Safety Inspection, (7) Plant Startup From Refueling /Startup Testing-Refueling, (8) Potentially Generic Items, (9) Information Notices, (10) IE Bulletins, (11) Allegation Followup, and (12) Cold Weather Preparation.
Results: Within the areas inspected, no violations or deviations were identified.
8703190410 870316 PDR ADOCK 05000302-Q PDR
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DETAILS 1.
Persons Contacted-Principal Licensee Employees J. G. Dewease, Senior Vice President, Nuclear Operations
- R. P. Barkhurst, Vice President, Nuclear Operations
- N. S. Carns, Plant Manager, Nuclear T. F. Gerrets, Corporate QA Manager S. A. Alleman, Assistant Plant Manager, Plant Technical Staff J. R. McGaha, Assistant Plant Manager, Operations and Maintenance J. N. Woods, QC Manager A. S. Lockhart, Site Quality Manager R. F. Burski, Engineering and Nuclear Safety Manager K. L. Brewster, Onsite Licensing Engineer
- G. E. Wuller, Onsite Licensing Coordinator T. H. Smith, Maintenance Superintendent, Nuclear
- Present at exit interviews.
In addition to the-above personnel, the NRC inspectors held discussions with various operations, engineering, technical support, maintenance, and administrative members of the licensee's staff.
2.
Unresolved Items Unresolved items were not identified during this inspection.
An unresolved item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
3.
Plant Status The inspection period began with the reactor in Mode 5.
The reactor entered Mode 4 at 5:37 a.m. on February 1, 1987, after repairs were completed on the 1A reactor coolant pump and was placed in Mode 3 at 3:15 p.m. that same day.
At 12:25 a.m. on Fabruary 3, 1987, the lowest setting main steam line code safety valves lifted in response to increasing steam pressure.
At the time of the event, the steam temperature pressure was being controlled manually on the bypass around the main steam isolation valve because, both the steam bypass control system and the steam generator atmospheric dump
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valves were being worked on by maintenance personnel.
The operators responded too slowly to the temperature which had been increasing over a number of hours.
The NRC inspector verified that the safety valves lifted
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at the correct pressure and properly reset. The licensee has conducted a critique of this event. The reactor was taken critical for the first time in Cycle 2 at 10:19 a.m. on February 4,1987.
At 10:04 a.m.- on February 7,1987, the reactor entered Mode 1 after completing low power physics testing. The unit was placed on the electrical grid at 6:36 p.m. the following day. Upon completion of core physics testing, at the 70 percent power plateau, a further power increase was precluded by a positive moderator coefficient which is prohibited by Technical Specification 3.1.1.3 above the 70 percent power level.
Following the submittal of an emergency Technical Specification change, the licensee was permitted by the Office of Nuclear Reactor Regulation to invoke Technical Specification (TS) Test Exception 3.10.2 and proceed to 85 percent power. The moderator temperature coefficient was determined to be negative by a test conducted on February 20th, and the reactor was subsequently taken to full power.
At 6:20 a.m. on February 22nd, a small steam line in the turbine building sheared. The break was located on a 1-inch orifice section of a 4-inch vent line from an extraction' steam line from the moisture separator reheater (MSR) to the high pressure (No.1) heaters.
In order to isolate the line, the high pressure heaters and moisture separator reheaters were
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secured. The reactor was stabilized'at about 77 percent power. The power reduction was made primarily to limit thermal stresses on the low pressure turbine blades. The reactor was subsequently returned to full power.
No violations or deviations were identified.
4.
Licensee Event Report (LER) Followup The following LERs were reviewed and closed. The NRC inspectors verified that reporting requirements had been met, that causes had been identified, that corrective actions appeared appropriate, that generic applicability
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had been considered, and that the LER forms were complete. Additionally, the NRC inspectors confinned that no unreviewed safety questions were involved and that violations of regulations or TS conditions had been identified.
(Closed) LER 382/85-17. " Reactor Trip on Inadvertent Closure of Main Steam Isolation Valve." During the first refueling outage, the licensee performed major modifications to the main steam isolation valve hydraulic system including relocating components that were near the valve yoke.
(Closed) LER 382/85-32, " Automatic Actuation of Reactor Protective System Due to Out-of-Range ASI."
During the first refueling outage, the licensee upgraded core protection calculator / core operating limit supervisory system.
j (Closed)LER 382/85-33, " Automatic Actuation of Reactor Protective System Due to faulty Feed Pump Control."
In addition to the corrective actions described in the report, during the first refueling outage, the licensee
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replaced the feed pump control syste t i'
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(Closed)LER 382/85-44, ~" Reactor Trip Due to Over. Feeding the Steam Generators." 'In addition to the corrective actions described in this report, the licensee has installed a bypass to the high steam generator level reactor trip, which in accordance with the modified TS, can be employed at power levels below 20 percent.
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(Closed) LER 382/86-03,." Inadvertent Control Room Isolations Due to-
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Instrument Anomalies." Subsequent-to the corrective actions described in this rep <srt, the licensee. replaced the chlorine detectors that were originally installed (Capital Controls Company, Advance Chlorine Leak Detector Sensor 1030) with a more reliable system (manufactured by Sensidyne).
(Closed)LER 382/86-23, " Reactor Trip Due to Dropped Control Element Assembly."
(Closed)LER 382/86-24, " Reactor Trip During Startup Due' to Prolonged Low Power Operation."
(Closed)LER 382/86'-28, Revision 0 and 1, " Loose Steam Generator Tube Plugs Due to Inadequate Installation." The NRC inspector has reviewed both of these reports, and the analysis in Revision 1 appears to adequately address all possible concerns.
No violations or deviations were identified.
5.
Followup of Previously Identified Items (Closed) Deviation 382/8633-01, " Failure to Revise OP-1-003 by Date Specified in LER 382/86-15." This item is considered closed as the procedure has now been revised, and no written response to the deviation was required.
(Closed) Open Item 382/8616-01, " Restoration of 3A & 3AB Battery Racks to Design Requirements." The NRC inspector verified that the licensee removed the temporary shims and modified the racks for proper spacing.
(Closed) Unresolved Item 382/8528-03, " Correction of Annunciator Setpoint Errors." The NRC inspector has verified that the annunciator response procedures have been revised to indicate the proper alarm / actuation setpoints.
(Closed) Violation 382/8520-04, " Failure To Conduct a Proper 10 CFR 50.59 Review." This violation, involving operation of the control roon ventilation in the re-circulation mode for extended periods of time, was addressed in NRC Inspection Report 50-382/86-06.
Ir this report, an observation was made on the need for training of plant personnel on i
10 CFR Part 50.59 reviews. The licensee has implemented a " Safety
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Evaluation Training Course" on 10 CFR Part 50.59 reviews.
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Operations Review Coneittee nembers and alternates, plant nanagement, and various plant engineering and technical support staff have since attended this course.
No violations or deviations were identified.
6.
Monthly Maintenance Station maintenance activities affecting safety-related systems and components were observed / reviewed to ascertain that the activities were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with TS.
The NRC inspectors observed portions of the work associated with the following Condition Identification Work Authorizations (CIWA).
CIWA 028521 - Safety System Statur Panel Switch No. 10 CIWA 031493 - Steam Generator 1 indicated wide range level difference CIWA 030540 - Packing replacement on Valve SI-346 CIWA 030087 - Packing replacenent on Safety Injection Pump Recirculation Line Isolation Valve SI-120A No violations-or deviations were identified.
7.
Monthly Surveillance The NRC inspectors observed / reviewed TS required testing and verified that testing r s performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation (LCO) were met, and that any deficiencies identified were properly reviewed and resolved.
The NRC inspectors reviewed the results of the TS required functional testing performed on snubbers during the first refueling cutage.
In a letter dated August 14, 1986, the licensee informed NRC Region IV, pursuant to TS 4.7.8.e, that the 10 percent sample plan (TS 4.7.8.e(1))
was chosen for snubber sampling and functional testing for the first refueling outage.
Because of a high failure rate among the small nechanical snubbers (PSA-1/4, 1/2), the NRC inspectors' review concentrated on this group.
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The high failure rates also eventually forced the licensee to test 100 percent of the small mechanical snubbers in order to meet the TS requirements for additional testing following the discovery of failures.
Of 137 small mechanical snubbers examined, 37 were classified as failures.
The licensee attributed the cause of 21 of the 37 failures to installation
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error.
A typical analysis of one of these installation failures stated,
"... the snubber had twisted guiderods, caused by improper rotation of the snubber paddle with respect to the housing during installation."
The NRC inspectors discussed the large number of snubber installation errors with licensee management.
The NRC inspectors were told that apparently these' snubbers were reinstalled improperly following testing prior to initial plant startup.
The licensee intends to revise MM-12-001, " Pipe Hanger Support Installation, Fabrication, and Removal,"
to caution mechanics about proper snubber orientation.
Inspector followup of the change to MM-12-001 is identified as Open Item 382/8704-01.
A summary of the licensee's testing of small mechanical snubbers was provided to the Vendor Program Branch of the NRC Office of Inspection and Enforcement for review and possible incorporation into a future inspection of the snubber vendor.
On February 20, 1987, the NRC inspector observed portions of NE-2-002,
" Variable Tavg Test," which was performed to gather data for a calculation of the existing moderator temperature coefficient (MTC).
As discussed in paragraph 3 of this report, the plant was limited to 85 percent power until MTC was determined to be less than or equal to 0.0x10-4 delta K/K/ degrees Fahrenheit.
No violations or deviations were identified.
8.
Routine Operational Safety Inspection By observation during the inspection period, the NRC inspectors verified that the control room manning requirements were being met.
In addition, the NRC inspectors observed shift turnover to verify that continuity of system status was maintained.
The NRC inspectors periodically questioned shift personnel relative to their awareness of the plant conditions.
Through log review and plant tours, the NRC inspectors verified compliance with selected TS and limiting conditions for operations.
During the course of the inspection, observations relative to protected and vital area security were made including access controls, boundary
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integrity, search, escort, and badging.
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On a regular basis, radiation work permits (RWP) were reviewed and the specific work activity was monitored to assure the activities were being conducted per the RWPs.
Selected radiation protection instruments were
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periodically checked and equipment operability and calibration frequency were verified.
The NRC inspectors kept themselves informed on a daily basis of overall status of plant and of any significant safety matter related to plant operations.
Discussions were held with plant management and various t
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members of the operations staff on a regular basis.
Selected portions of operating logs'and data sheets were reviewed daily.
The NRC inspectors conducted various plant tours and made frequent visits of the control room.
Observations included:
witnessing work activities in progress; verifying the status of. operating and standby safety systems and equipment; confirming valve positions, instrument and recorder readings, and annunciator alarms; and housekeeping.
No violations or deviations were identified.
9.
Plant Startup From Refueling /Startup Testing-Refueling The NRC inspectors witnessed portions of NE-02-020, Revision 0, "CEA Insertion Time Measurement," which was performed on February 2, 1987.
After all the prerequisites and initial conditions were satisfied, licensee personnel were initially unable to perform the test.
At first, it appeared that there was a problem with the test software that was inserted into the control element assembly calculator (CEAC).
However, after consulting Combustion Engineering (CE), it was determined that the problem was procedural.
Instead of having both CEACs operable for the test, the one not used for the test software was placed in an inoperable condition.
Without both CEACs operable, the software would not run the test.
On February 4, the NRC inspectors observed the approach to criticality which.was performed in accordance with NE-2-030, Revision 0, " Initial Criticality." Subsequently, the NRC inspectors observed the performance of NE-2-060, " Isothermal Temperature Coefficient Measurement," for the essentially all-rods-out configuration and NE-2-040, Revision 0, "CEA Group Worth and CEA Coupling Checks."
No violations or deviations were identified.
10.
Potentially Generic Items The NRC inspector provided the licensee with copies of the following 10 CFR Part 21 reports:
End of Life susceptibility of E-Line and H-Line instruments
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manufactured by Foxboro Company. (87-03)
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Armature binding in General Electric HFA relays. (87-11)
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I-Reactor coolant pump anti-reverse rotation device pin wear.
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(87-05)
Main steam isolation valve thrust bearings. (87-06)
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Infomation on Valcor solenoid valve spring failure. (86-09)
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Corrected list of model numbers from Validyne. (87-10)
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Incomplete flux material on weld rods manufactured by Airco.
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(87-04)
The NRC inspector also infomed the licensee about the emergency diesel generator failure at Braidwood NPS due to inadequate clearance between the cross head and cross head guide.
No violations or deviations were identified.
11.
Information Notices
. In response to IE Information Notice No.86-106, "Feedwater_ Line Break,"
the licensee performed UT examinations for wall thickness reductions, in addition to those performed per Plant Procedure PE-5-034, " Surveillance Procedure for Steam Lines Subject to Moisture Induced Erosion / Corrosion."
All pipe wall thickness measurements were within their criteria. The licensee also has a new Procedure PE-5-036, " Surveillance Procedure for Secondary and Auxiliary Water Systems Subject to Flashing, Erosion, and Corrosion," currently in the review process.
The NRC inspectors verified the licensee received IE Information Notice 86-101, " Loss of Fluid Levels in the Reactor Coolant Syst.em."
In part due to the events described in LER 382/86-15, "Simultanecusly Using Two Methods of Draining Reactor Coolant System Results in Loss of Shutdown Cooling," the licensee has already upgraded procedures relating to decay heat removal capability.
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IE Bulletins (Closed)
IEB 86-01, " Minimum Flow Logic Problems That Could Disable RHR Pumps." This is not applicable to Waterford 3.
No violations or deviations were identified.
13. Allegation Followup The NRC inspector performed inspections to detemine the validity and, if verified, the safety significance of the following concern:
(Closed) Concern 4-87-A-003 It was alleged that certain individuals were directed to install a valve (potentially contaminated) without use of the protective measures specified on the radiation work permit (RWP) for the job.
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Finding It was found that surveys had been performed, and the valve was released from radiological controls. A health physics technician performed the survey and released the valve. He then directed individuals to install the valve without protective equipment, but he failed to update the protective measure requirements of the RWP.
Since the valve had been released, there was no safety concern.
However, concerns with the failure to update RWP's and actions that the licensce is taking to correct the problems are addressed in NRC Inspection Report 50-382/86-33.
No violations or deviations were identified.
14. Cold Weather Preparation The NRC inspector verified that work items that affect the operability of freeze protection circuitry have been completed.
lio violations or deviations were identified.
15.
Exit Interview The inspection scope and findings were summarized on March 2, 1987, with those persons indicated in paragraph 1 above. The licensee acknowledged the NRC inspectors findings. The licensee did not identify as proprietary any of the material provided to or reviewed by the NRC inspectors during this inspection.
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