IR 05000219/1986017
| ML20212K021 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/01/1986 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212J996 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-TM 50-219-86-17, NUDOCS 8608190025 | |
| Download: ML20212K021 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-219/86-17 Docket No.
50-219
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License No.
DPR-16 Priority Category C
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Licensee:
GPU Nuclear Corporation 100 Interpace Parkway Parsippany, New Jersey
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Facility name: Oyster Creek Nuclear Generating Station Inspection At: Forked River, New Jersey Inspection Conducted: June 1 - July 6, 1986 Participating Inspectors:
W. H. Bateman, Senior Resident Inspector J. F. Wechselberger, Resident Inspector W. H. Baunack, Project Engineer Approved by:
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.P- /~d'C A. R. B16d'gh, Chief Date Reactor Projects Section IA Inspection Summary:
Routine inspections were conducted by the resident inspectors and a Region based inspector (179 hours0.00207 days <br />0.0497 hours <br />2.959656e-4 weeks <br />6.81095e-5 months <br />) of activities in progress including outage management, maintenance, modifications, QC inspection activity, radiation control, physical security, housekeeping, and fire protection. The inspectors also reviewed licensee's response to selected safety issues and investigation of isolation condenser snubber operability. The inspectors followed up on outage problems including high radiation door control, inadvertent starts of the Standby Gas Treatment System and emergency sirens.
Results:
One violation was identified.
The violation involved failure to maintain positive control of a locked closed high radiation door and failure to conduct required radcon briefings.
The review of the licensee's documentation on isolation condenser snubbers revealed that more inoperable snubbers may be found once the licensee's inspection is completed and that some snubbers on the system were deficient prior to contractor work commencing during the Cycle 10R outage. Two examples of failure to proceduralize required operator actions were identified and discussed with the licensee.
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DETAILS 1.
Survey of Licensee's Rasponse to Selected Safety Issues A survey was conducted primarily to determine the actions that the licensee is taking to address a selected sample of safety issues. These issues have been identified in IE Bulletins, Information Notices, and. in the Institute of Nuclear Power Operations (INPO) Significant Operating Event Peports (SOERs).
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The licensee's responses to the following documents were reviewed with regard to the issue of biofouling of heat exchangers.
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IE Information Notices 81-21 and 83-46
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INPO SERs 63-81, 42-83, 47-83, and 16-84 INPO SOER 84-1
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Generally, the survey concerned the capability to monitor safety related equipment cooled by open-cycle service wa?.er systems. Specifically,' the survey addressed the following questions:
1.
Is instrumentation available on safety related equipment to mon-l itor flow changes and heat exchanger performance?
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l The systems of interest for this inspection are the Emergency Service Water system (ESW) and the Service Water system (SW).
The ESW systen. supplies cooling water to the containment spray heat exchangers while the SW system supplies the reactor build-ing closed cooling water (RBCCW) and the turbine building closed cooling water (TBCCW) heat exchangers. The RBCCW system
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supplies the drywell coolers, shutdown cooling and clean-up i
system heat exchangers, recirculation pumps, and fuel pool cooling system loads.
The ESW side of the containment spray heat exchangers was modified during the Cycle 10 outage with a differential pressure gage across each division plate.
In addition, an annubar has been installed on ESW System I for flow indication.
System II has not been modified with an annubar. The System II modifica-tion was planned for Cycle 10 outage but was deferred to the
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present Cycle 11 outage. Additional engineering design work may delay the modification until Cycle 12. The licensee will continue to use an ultrasenic flow measuring device to measure ESW system II flew until the modification is completed.
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The licensee is modifying the RBCCW heat exchanger service water side to incorporate flow measuring devices during the current Cycle 11 outage. The service water side has inlet and outlet pressure gages.
2.
Are readings recorded and reviewed agcinst design parameters on a routine basis?
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The licensee performs an inservice test (quarterly requirement)
J of the ESW system on a monthly basis. This test compares and trends ESW sytem flow and delta pressure across the containment spray heat exchangers. Alert and action levels are determined.
The SW inservice test is performed quarterly. The test measures SW flow and total head which are compared against baseline established alert and action levels.
3.
Do procedures and training address operator action if signifi-cant heat exchanger performance degradation is detected?
The licensee's abnormal operating, diagnostic and restoration, and system operation procedures address operator actions if sys-tem degradation is detected. The procedures specify plant pa-rameters to monitor and required operator action if a specific plant parameter limit is exceeded. The operators are trained on these procedures and, in particular, on past ESW heat exchangers biofouling/ debris problems.
e 4.
Are periodic inspections performcd?
The licensee conducts periodic inspections of the heat exchang-ers as required by Nuclear Mutual Limited Insurance Company, which usually occurs every refueling outage.
In addition, in-spections are procedurally required based upon system operating performance. The licensee is replacing the RBCCW heat exchang-ers during the current outage. The containment spray heat ex-changers were inspected in October 1985 and one will be inspected during this outage. This inspection will determine if the other containment spray heat exchanger will be inspected.
The diesel generators at Oyster Creek are air cooled and, there-fore, were not considered. The fire protection system water source is a fresh water pond and therefore also was not consid-ered.
The licensee plans to install a new chlorination system using a liquid bleach application. This is projected to be com-l pleted 'oy April 1987.
l No unacceptable conditions were identified.
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2.
Isolation Condenser Snubbers Inspection Report 86-12 discussed missing welds on pipe clamps that form a part of the overall snubber assembly on two snubbers. The previous monthly report discussed the question of snubber and isolation condenser system operability. The licensee has performed an initial simplified analysis on the isolation condenser to determine system operability with the deficient snubbers removed from the system and found the system to be operable. The licensee is contemplating performing additional analysis to make a more reasonable evaluation of system operability by rerunning the analysis with the loads carried by the missing snubbers redistributed to the remaining snubbers and supports.
The licensee determined that these snubbers were made inoperable during the Cycle 10 outage as a result of extensive repair work conducted on the isolation condenser piping. The repair work comprised weld overlay and pipe replacement necessitated by indications of intergranular stress corrosion cracking (IGSCC) found in some of the piping welds. Apparently the snubbers were not reassembled properly by the contractor, omitting the weld attachment of the shear lug to the pipe clamp. The licensee had ex-perienced numerous quality control problems with the contractor on the Isolation Condenser system repair and as a result increased their quality control monitoring of the contractor's activities. Partially in response to these problems, an inspection was performed.of the system's snubbers ar.d supports on August 6, 1984 by a Plant Materiel engineer who identified shehr lug weld problems with snubbers NE-1-SS, NE-2-S13, NE-2-S9, NE-2-S10, and ME-1-S4 among other problems. The following snubber shear lug welds were not inspected during this walkdown because they were l
covered over by pipe isolation:
NE-2-S21
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NE-1-S7
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632-R7
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NE-1-S12 NE-1-59
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632-R3
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NE-1-511 At the end of this report period, the licensee had six snubbers remaining to be inspected to verify proper shear lug welds:
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NE-2-S9 (Shear lug weld identified as missing in 8/6/84 inspection)
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NE-2-S10 (Shear leg weld identified as missing in 8/6/84 inspection)
NE-1-S7
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632-R7
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NE-1-S12 NE-2-S21
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As indicated, two of these snubbers were previously reported in the August 6, 1984 inspection as having shear lug welds missing. Thus, additional snubber deficiencies are suspected.
Prior to the licensee's August 6, 1984 inspection, the contractor who performed the IGSCC weld overlay repairs performed an inspection of the snubber and hanger assemblies associated with the isolation condenser to document existing conditions. This inspection identified three snubbers with missing shear lug welds as follows:
(Note: These weld deficiencies were reportedly corrected by the contractor.)
NE-2-S9
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NE-2-S10 NE-2-S11
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These inspection results identified existing snubber problems prior to the contractor performing any work on the Isolation Condenser system and may be indicative of problems with other safety systems. The inspector will follow up with the licensee to determine future inspection plans to verify that other safety systems do not have the same problem.
(219/86-17-01)
The above status update was based on discussions with the licensee personnel and review of the following dccuments:
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GPUN Quality Deficiency Report 84-060
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Contractor Non Conformance Report OC-8
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Maintenance and Construction Critique 84-0
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Contractor and Non Conformance Report OC-10
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Plant Material Inspection Report (8/6/84)
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Material Non Conformance Report 84-523
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Licensee Meeting Minutes
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Contractor - 100% Verification of Job Complete
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Material Non Conformance Report 84-443
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Inservice Inspection Reports.
3.
Radiation Protection During entry to and exit from the RCA, the inspectors verified that proper warning signs were posted, personnel entering were wearing proper dosimetry, personnel and materials leaving were properly monitored for l
radioactive contamination, and monitoring instruments were functional and in calibration.
Posted extended Radiation Work Permits (RWPs) and survey status boards were reviewed to verify that they were current and accurate.
The inspector observed activities in the RCA to verify that personnel com-plied with the requirements of applicable RWPs and that workers were aware of the radiological conditions in the area.
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During a routine tour of the reactor building on June 25, 1986, the in-spectors discovered the normally locked closed high radirtion door to the 51' cleanup room open. The floor radiological controls technician (RCT)
was notified, who called into the area, attempting to verify the presence of personnel.
The door was then secured, the Group Radiological Control Supervisor (GRCS) was notified and a high radiation key obtained from the radiation work permit (RWP) office. One RCT monitored the cleanup room door while another RCT donned protective clothing to make an entry into the cleanup room. The RCT verified the absence of personnel in the cleanup room and exited the area.
The inspector reviewed licensee radiation surveys and confirmed that the cleanup room contains areas of radiation greater than 1000 milliroentgens per hour; therefore, controls for a locked high radiation area applied.
The inspector reviewed the licensee's subsequent Radiological Investigative Report to determine the cause of the event. The licensee determined that there were two procedural violations involving the proper issuing and con-trolling high radiation keys. The first violation involved issuance of a key to a contractor worker unfamiliar with proper high radiation key con-trol procedures without a proper briefing from the issuing RCT. The RCT was involved with a personnel decontamination problem at the time and fail-ed to properly brief the worker. Upon receiving the key, the worker is
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recuired to sign an acknowledgement signifying that the worker has been properly briefed and understands his responsibilities for proper key control. On the reverse side of this Locked High Radiation Area Control and Key Usage Log is printed the particulars of the brief.
In addition, the licensee has posted this list of briefing points outside the RWP office.
The worker signed the Key Usage Log without receiving a briefing.
The second procedural violation involved the transfer of the key to the worker's foreman without first notifying the RWP office.
Procedure 9300-ADM-4110.06, Rev. 2, Control of Locked High Radiation Areas, requires the RWP office be notified of each key transfer.
Reportedly, the worker gave the key to his foreman who did not maintain positive control of the
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l high radiation door. The contractor foreman did not have visual control of the door, did not notify the RCT that all personnel had exited the area, did not lock the access door upon work completion, nor did he sign the High Radiation Key Usage Log.
Further, a second contractor foreman who had entered the high radiation area with the contractor worker to review work activities, left the area after the worker and neglected to ensure the high radiation door was locked.
l The licensee's corrective actions thus far have been to issue a written
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reprimand to the contractor workers involved and discuss the incident with all field personnel, in addition to exploring mechanical methods to insure proper high radiation door control. This had been a problem during the Cycle 10 outage and the licensee has already identified other incidents during this outage. The failure of the licensee to maintain a high radia-tion area (in which the intensity of radiation is greater than 1000 milli-roentgens per hour) locked with positive control over each entry is con-trary to the requirements of Technical Specifications 6.13 and forms the
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first part of a single violation. Additionally, the failure to properly brief workers entering high radiation areas on locked high radiation key control is contrary to the requirements of Station Procedure 9300-ADM-4110.06 and forms the second part of a single violation. (219/86-17-02)
In addition to the response required by 10 CFR 2.201, the licensee should also address in his response the adequacy of radcon staffing to support ongoing work activities and the apparent ignorance or poor attitude of contractor workers that led to this noncompliance.
4.
Laundry Shipment On June 24, 1986 the licensee received a laundry shipment from Interstate Nuclear Services without the appropriate shipping papers. The shipment was classified as low specific activity (LSA) containing 11.605 milli-curies and was inspected and accepted by the licensee noting no other discrepancies. Apparently the shipping papers were left in another truck when immediately prior to departure the trucks were changed. The shipper electronically transmitted the shipping papers to the licensee later in the day. The inspector discussed this event with Region I radiation protection specialists, who determined the event was under State of New Jersey jurisdiction. No inspector concerns remained after follow up of this matter.
5.
HFA Relays The licensee, in a letter to Region I dated May 22, 1986, requested a par-tial deferral of HFA relay replacement until Cycle 12R.
Bulletin 84-02 (March 12, 1984) required the replacement take place within two years of the date of the Bulletin. The licensee responded to Bulletin 84-02 in a letter dated July 18, 1984 stating, in part, that all nylon HFA relays on both normally energized and normally de-energized safety related systems would be replaced during the 11R refueling outage (April - October 1986).
The licensee's May 22, 1986 letter requested that replacement of 97 DC excited nylon spool HFA relays in safety-related applications (normally de-energized) be deferred until Cycle 12R outage scheduled to commence April 1988. The inspectors discussed with licensee personnel the inade-quacies of the May 22 letter, including lack of justification for the deferral and lack of compensatory testing or inspection.
In response the licensee intends to provide additional justification for this deferral request in a letter to the NRC, as well as their plans for replacement of some of the 97 normally de-energized HFA relays indicated above during the current Cycle 11R outage. NRC Region I will delay its decision on this matter until receipt of the supplemental correspondence.
6.
Safeguard System Actuated by Severe Thunderstorms On June 11, 1986 the facility experienced severe thunderstorms in the area which caused the Standby Gas Treatment System (SBGTS) to initiate and the primary and secondary containment to isolate several times. Voltage fluc-tuations caused by lightning strikes resulted in the automatic bus transfer (ABT) switch shifting from its selected power supply vital AC power panel
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1 (VACP-1) to its alternate supply (VMCC 1A2).
This caused a relay (6K37)
to de-energize and initiate SBGTS and isolate Primary and Secondary Containment. The ABT is a break before make design with a transfer time of sufficient duration to allow the relays to de-energize and actuate the safeguard systems.
The licensee evaluated the transients and found the plant response to be normal for the voltage fluctuations.
In addition, the licensee has tasked Plant Engineering to evaluate the ABT transfer time, preventive mainte-nance, and condition to determine if the transfer time is adequate for this application.
The NRC inspectors considered licensee action to address this event to be adequate.
7.
Spurious Initiation of Standby Gas Treatment System
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On two separate occasions during this report period, the Standby Gas Treatment system (SBGTS) was determined by the licensee to have been actuated by faulty power supplies to an area radiation monitor and a process radiation monitor. On June 9, 1986 while performing maintenance on RN04 A-6, isolation condenser vent radiation monitor, the power supply, RN37, caused its associated process raciation monitors to fluctuate both upscale and downscale.
RN37 also supplies the reactor building vent monitors RN04 A-1 and A-2, wnich isolate normal reactor building ventilation and initiate SBGTS on an upscale signal of 13 mr/hr. After the initiation signal, the instrumentation and control technician reset the trip, clearing the actuating signal for SBGTS. The licensee did not perform the proper notification in accordance with 10CFR 50.72, nor did they notify the radiological controls department of the event. The shift personnel assumed that since the automatic initiation was due to a spurious signal generated by a faulty power supply, there were no reports required. The Operations Director counseled the shift personnel on the notification requirements and issued a memorandum to all control room operations personnel concerning proper reporting requirements.
The proper notification was made by the licensee the following day.
The power supply (R011B) to the reactor building refueling floor area radiation monitor failed on June 19, 1986 causing the SBGTS to initiate and normal reactor building ventilation to isolate. The licensee replaced the power supply shortly after the failure occurred and restored normal ventilation to the reactor building. The proper notification was made by the licensee.
Licensee actions were considered adequate.
8.
Outage Activities A.
MSIV External Closing Springs Atwood and Morrill Company, Inc. notified GPUN in a letter dated June 13, 1986 of the potential failure of the external closing assist springs on the MSIVs. Detroit Edison's Enrico Fermi Unit 2 experi-enced a failure of the inner spring of a spring set which fit around
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the yoke rods joining the air cylinder to the valve cover. Metal-lurgical examination of the broken springs confirmed that the failure was caused by quench cracks developed during the manufacturing process. Atwood and Morrill recommended the external closing springs on the MSIVs be magnetic particle inspected at the first available opportunity.
General Electric has issued an advance information notice on MSIV spring recommendations prior to releasing a service information let-ter (SIL). General Electric recommends performing a visual inspec-tion of the inner and outer springs for cracks in lieu of MPT during a scheduled outage and further states that the potential failure of a MSIV external closing spring is of no safety significance.
The licensee has accepted the General Electric recommendation and has performed a visual inspection of the MSIV springs.
No cracks were identified by the licensee in the MSIV springs. An Inspection and Enforcement (I & E) Information Notice is planned on this subject.
If the I & E Information Notice is issued, the inspector will verify that the licensee actions are in accordance with the Notice recommendations.
(219/86-17-03).
B.
Isolation Condense Weld Inspection
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The licensee found indications of intergranular stress corrosion cracking (IGSCC) on weld NE-1-27 in the isolation condenser system piping. NE-1-27 is located in the "A" isolation condenser steam pip-ing. This discovery occurred during inservice inspection activities of the weld overlays which were completed on the isolation condenser piping during Cycle 10R. The weld overlays were performed as a re-sult of IGSCC indications in certain piping welds. The licensee presently plans to increase the scope of their inspection to include 20 additional welds. The increased scope includes welds in both "A" and "B" isolation condenser steam and condensate piping outside the d rywell.
Repairs will be made to NE 1.-27.
The inspectors will continue to follow Isolation Condenser piping weld inspection activity.
C.
Scram with Mode Switch in Refuel The licensee attempted to lock the Mode switch in the Refuel position on June 25, 1986 prior to commencing control rod drive exchanges, but was unable to lock the switch in this position. During this attempt a full scram was actuated, apparently as a result of the shutdown contacts closing from the switch manipulation.
The hydraulic control units were valved out due to the cold shutdown plant condition with no fuel present in the reactor vessel.
Thus, only the scram relays actuated. The licensee is presently planning to repair the Mode switch prior to commencing refueling operatations.
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D.
Limitorque Valve Operators The licensee purchased new Limitorque valve operators for use in plant safety systems. The Limitorque coerators arrived with grease in the spring packs. Grease in the spring pack area may change the spring constant of the Belleville washers and thus affect the valve operating characteristics. The vendor was contacted regarding this problem and informed the licensee that the valves had a modification which incorporated a bypass line to pass the grease from the spring pack volume to the valve operator gear box. The licensee is awaiting a confirmation letter from the vendor documenting the Limitorque valve operator modification.
d.
Loose Control Room Wires The licensee discovered a loose wire (#783) off of terminal TB 11-7 for the overvoltage relay for BUS 1A. An immediate maintenance action was implemented to re-land the wire. The licensee could not determine how the wire became loose. The inspectors questioned the licensee as to what action they intended to take to ensure there were no other loose conneccions in Control Room panel wiring.
In response the licensee committed to a QC inspection, prior to restart from the 11R outage, of wire terminations in areas where maintenance and
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modifications were performed.
9.
Emergency Sirens
The licensee experienced problems with the emergency sirens during this t
report period. One of these problems involved inoperable sirens as a l
result of a lightening strike and the other resulted from the failure of the system encoder.
Failure of the encoder rendered all 47 sirens
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The licensee in each case made the appropriate notifications and was able to return the sirens to an operable status within a short i
period. The licensee is continuing to investigate if action can be taken
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to prevent these problems in the future.
The inspectors had no further questions at this time.
10.
Containment Spray System The licensee reported the Containment Spray system would not meet 831.1 design standards during a seismic event. A 10 CFR 50.72 report was made as a result of the discovery that the Containment Spray System ring headers inside the drywell were not seismically qualified. During the l
October 1985 cutage, the licensee inspected the ring header supports and i
recently performed a computer analysis of the system using that inspection data. At the end of this report period, the licensee was performing
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l an analysis of the system to determine operability. Modifications to correct the deficiencies to the Containment Spray system are planned to be
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accomplished prior to plant startup from the current Cycle 11R refueling outage. The inspector will review the results of the computer analysis and the licensee's actions during a future inspection. (219/86-17-04)
11.
Low and Low-Low Reactor Water Level Instrument Replacement The licensee is planning on replacing the existing low (RE05s) and low-low (RE02s) reactor water level instruments with Foxboro analog trip units and Rosemount transmitters.
The present SOR Inc. differential pressure switches (model number 103 AS-BB212-NX-JJTTX6) have experienced signifi-cant drift problems at the RE05 position.
Previously, the licensee had installed model number 103 AS-8212-NX-JJTTX6 at the REOS position but had to replace it with "BB" model because of the drift experienced in the switch setpoint. The RE02 position also has a SOR "BB" model installed and likewise will be replaced.
Inspection Reports 86-02, 86-04, and 86-06 provide additional information on the SOR differential pressure switch performance.
The licensee has completed most of the instrument rack work necessary for changeover for the RE05s. Present plans are to replace the RE05s after refueling.
The SOR differential pressure switches would remain installed during the refueling activities. The RE02s would be replaced when new equipment arrives on site and prior te restart from 11R.
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12. Housekeeping During routine tours of the power block, the inspector observed that housekeeping was deteriorating on elevations 23' and 51'.
The inspectors discussed these observations with plant management who stated they felt the problem was not so much housekeeping as it was clutter of items needed t
I to accomplish various jobs and unavoidable dirt from work in progress.
The inspectors stressed their point that they felt housekeeping was deteriorating, and the licensee agreed to upgrade the housekeeping on
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these two elevations.
Housekeeping in other areas of the plant was accep-
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table. The inspectors will continue to observe housekeeping during the remainder of the outage.
13. Proceduralization of Required Operator Actions In a letter from NRC licensing (J. Zwolinski) to GPUN (P. Fiedler) dated 4/24/86 and entitled " Schedular Exemption - Compliance with 10 CFR 50.44(c)(3)(tii) - Isolation Condenser High Point Vents (TAC 59342)," it was clarified that the licensee was taking credit for operator action to open the existing high point vents.
The inspectors reviewed plant proce-i dures to ensure this action, involving lifting leads and installing jumpers, was incorporated. The inspectors found the action was not pro-ceduralized.
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In a letter from NRC licensing (J. Donohew) to GPUN dated 5/22/86 and en-titled " April 2, 1986, Meeting with GPU Nuclear Corporation (GPUN) to Dis-cuss the Post-Accident NUREG 0737 Items II.F.1.1 and II.F.1.2," it was clarified that there are plant procedures to load the turbine building ventilation fans onto an emergency diesel generator if needed. The in-spectors reviewed plant procedures and found the action required was not proceduralized.
The inspectors discussed the initial example of failure to proceduralize action for which they were taking credit with the licensee who stated they would correct the problem. After determining there was a second example, the inspectors again questioned the licensee who stated they also had picked up on the oversight after the fact. The licensee has stated both of the missing operator actions will be proceduralized. The inspectors will follow-up to ensure this action is completed (219/86-17-05).
The inspectors will continue to review licensee commitments to ensure required operator action is incorporated into applicable procedures.
14. Observation of Physical Security Dur'.ng daily tours, the inspectors verified access controls were in accor-dance with the Security Plan, security posts were properly manned, pro-tected area gates were locked or guarded, and isolation zones were free of obstructions. The inspectors examined vital area access points to verify that they were properly locked or guarded and that access control was in accordance with the Security Plan.
No inspector concerns were identified.
15.
Exit Ir.terview l
A summary of the results of the inspection activities performed during
this report period were made at meetings with senior licensee management i
at the end of the inspection. The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was included.
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