IR 05000266/1982010
| ML20058F148 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/15/1982 |
| From: | Bob Fitzpatrick, Guldemond W, Hague R, Hague R, Konklin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20058F129 | List: |
| References | |
| 50-266-82-10, 50-301-82-09, 50-301-82-9, IEB-82-01, IEB-82-1, NUDOCS 8207300309 | |
| Download: ML20058F148 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-266/82-10(DPRP); 50-301/82-09(DPRP)
Docket No. 50-266; 50-301 License No. DPR-24; DPR-27 Licensee: Wisconsin Electric Power Company 231 West Michigan Milwaukee, WI 53203 Facility Name:
Point Beach Nuclear Power Plant, Units 1 and 2 Inspection At: Point Beach Site, Two Rivers, WI Inspection Conducted: April 1 to May 31, 1982 a
Inspectors:
W. G. Guldemo e
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R. L. Hague ck'f r% ub
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B. E. Fitzpa f Z$)h Approved By:
J. E. Kon -
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Project
'ection 2A Inspection Summary Inspection on April 1 to May 31, 1982 (Report No. 50-266/82-10(DPRP)
50-301/82-09(DPRP))
Areas Inspected: Routine resident inspection of Operational Safety Verification, Monthly Maintenance Observation, Monthly Surveillance Observation, Followup on Licensee Event Reports, IE Bulletin Followup, Review of Plant Observations, Followup on Items of Noncompliance, Inde-pendent Inspection, Containment Integrated Leak Rate Test, Inspection During Long Term Shutdown, and Refueling Activities. The inspection involved a total of 252 inspector-hours onsite by three inspectors in-cluding 35 inspector-hours on off-shifts.
Results: Of eleven areas inspected, no items of noncompliance were identified in ten areas. One item of noncompliance was identified in one area (failure to maintain proper fire barrier conditions for pro-tection of a safety related area; Paragraph 2).
8207300309 820716 PDR ADOCK 05000266 G
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DETAILS 1.
Persons Contacted
- G. A. Reed, Manager, Nuclear Operations J. J. Zach, General Superintendent T. J. Koehler, Operations Superintendent J. C. Reisenbuechler, I & C Superintendent W. J. Herrman, Maintenance & Construction Superintendent R. S. Bredvad, Health Physicist
- R. E. Link, EQR Superintendent.
- F. A. Zeman, Staff Services Supervisor The inspectors also talked with and interviewed members of the Operations, Maintenance, Health Physics, and Instrument and Control Sections.
- Denotes personnel attending exit interviews.
2.
Operational Safety Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the months of April and May. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of both auxiliary and containment reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector, by observation and direct interview, verified that the physical security plan was being imple-mented in accordance with the station security plan. This included monitoring the searches of several incoming vehicles.
Unit I reactor was taken critical at 8:55 a.m. on April 12, 1982, completing a steam generator inspection outage which had started at 3:30 a.m. on March 26, 1982. The inspection included a 2000 psi primary to secondary hydrostatic test, an 800 psid secondary to primary hydrostatic test, and an eddy current inspection of essenti-ally 100'. of all tubes in both steam generators. As a result of the inspection, the licensee plugged 38 tubes in the "A" steam gen-erator, one of which had been sleeved in November 1981 and had developed a small leak in the cold leg side, and 14 tubes in the B steam generator. Three explosive plugs were weld repaired in the
"B" steam generator. Post recovery operation will be restricted by the licensee to a hot leg temperature of 557*F and 80% power.
In addition, during the Unit 1 outage the following deficiencies were corrected.
A primary Power Operated Relief Valve which had been leaking was lapped and returned to service.
Source Range Channel N-31, which
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had caused two reactor trips on December 9, 1981 (Ref. IE Inspection Report 81-23) was returned to service by placing a new detector in the spare detector well. The faulty detector, which is still located in the normal detector well, will be replaced during the fall 1982 Unit i refueling. A leaking safety injection check valve canopy seal weld, which had been previously peened over to reduce leakage contributing to elevated Unit I containment airborne particulate activity, was weld repaired.
During a routine plant tour at 7:30 p.m. on April 16, 1982, the inspector discovered the fire door between the 4D diesel generator room and the air compressor room open. No fire watch was present. The closure mechanism failed to close the door. The inspector secured the door and notified the licensee. The licensee was unaware of the open door and promptly repaired the closure mechanism. The length of time the door was open was indeterminate. This is an item of noncompliance with respect to Technical Specification 15.3.14.E (266/82-10-01; 301/82-09-01). This is a recurrent item of noncompliance previously identified in IE Inspection Report Nos. 50-266/81-11 and 50-301/81-13.
At 6:57 p.m. on May 26, 1982, Unit 2 was phased onto the grid after a 41 day refueling outage. During the outage the major work accomp-lished was replacement of the "A" reactor coolant pump motor, replace-ment of one core flux mapping thimble, steam generator eddy current inspection, plugging of 10 tubes in the "A" steam generator and 6 tubes in the "B" steam generator, completion of an upgraded fire pro-tection system in the containment, and various TMI related work.
The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. During the period of April 1 to May 31, 1982, the inspector walked down the accessibic portions of the auxiliary feedwater and emergency electrical systems to verify operability. The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedures.
3.
Monthly Maintenance Observation Station maintenance activities of safety related systems and com-ponents listed below were inspected to verify that they were conducted in acccrdance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications.
The following items were considered during this review; the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were
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Inspected as applicable; functional testing and/or calibrations were d
performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by quali-fled personnel; parts and materials used were properly certified;
radiological controls were implemented; and fire prevention controls were implemented.
Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance.
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The following maintenance activities were observed; replacement of the Unit 2 "A" reactor coolant pump motor, and replacement of one core flux mapping thimble.
Following completion of maintenance on the Unit 2 "A" reactor coolant pump motor, the inspector verified that these systems had been returned to service properly.
During a routine inspection of the "A" reactor coolant pump motor on April 28, 1982, the licensee found several cracks in the braze joint between the rotor shorting bars and lower shorting ring. After dis-cussions with Westinghouse, it was decided to replace the motor with
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an on hand spare. The licensee will attempt to decontaminate the original motor and ship it back to Westinghouse for repair.
If the
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motor cannot be sufficiently decontaminated for shipment, Westinghouse will do the repair on site if necessary.
4.
Monthly Surveillance Observation The inspector observed technical specifications required surveillance testing on the reactor protection and safeguards analog system and
the control room heating and ventilation system and verified that
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testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with
technical specifications and procedure requirements and were re-i viewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly
reviewed and resolved by appropriate management personnel.
The inspector also witnessed portions of the following test activities:
IT-1025 Ten-year pressure test of main steam and main feed systems, Unit 2 IT-1045 Ten-year pressure test of containment spray system, Unit 2 IT-1055 40-month pressure test of safety injection system.
IT-1095 Ten-year pressure test of auxiliary feedwater system, Unit 2
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ORT-3 Safety injection actuation with loss of engineered safeguards A-C, Unit 2 IT-04 Inservice testing of low head safety injection pumps, Unit 2 REI-20.0 Control rod worth, boron worth and end point measurement, Unit 2 5.
Licensee Event Reports Followup Through direct observations, discussions with licenseo personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications.
82-007/01TO Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment, Unit 1 82-008/03LO Observed inadequacies in the implementation of administrative or procedural controls, Unit 1 82-002/01T0 Degraded steam generator tubes, Unit 2 6.
IE Bulletin Followup For the IE Bulletin listed below the inspector verified that the Bulletin was received by licensee management and reviewed for its applicability to the facility.
If the Bulletin was applicable the inspector verified that the written response was within the time period stated in the Bulletin, that the written response included the information required to be reported, that the written response included adequate corrective action commitments based on information presented in the Bulletin and the licensee's response, that the 11-censee management forwarded copies of the written response to the appropriate onsite management representatives, that information discussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as described in the written response.
82-01 Alterations of radiographs of welds in piping subassemblies 7.
Followup on Items of Noncompliance Licensee actions in response to the items of noncompliance documented in the inspection reports below were reviewed to verify that the actions were completed as committed and were in conformance with regulatory requirements.
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DOCKET NUMBER NUMBER TITLE 50-301 81-21 Safety injection valve improperly positioned 50-266 81-19 Failure to follow approved procedures resulting in a system pressure transient 50-266 81-19 Failure to follow approved procedures in that boric acid tanks were not sampled as required 50-266 81-19 Failure to follow approved procedures resulting in an improper valve line-up 50-266 81-19 Failure to follow approved procedures in that a procedural change was made without concurrence 50-266 81-19 Failure to follow approved procedures in that RWP guidelines were not adhered to 8.
Plant Trips Following the Unit 2 plant trip at 0903 May 28,1982, the inspector determined the status of the reactor and safety systems by observa-tion of control room indicators and discussions with licensee per-sonnel concerning plant parameters, emergency system status and reactor coolant chemistry. The inspector verified the establishment of proper communications and reviewed the corrective actions taken by the licensee.
All systems responded as expected, and the plant was returned to operation.
The trip was caused by an electrical operator trouble-shooting a ground fault in the 125 volt DC system. As the operator was checking for loose or grounded connections a lead became disconnected from its terminal connector. This wire supplied power to one of the 1 out of 2 safety injection reactc trip relays.
9.
Refueling Activities Prior to fuel handling the inspector verified that all surveillance testing required by technical specifications and licensee procedures for fuel handling had been completed. The inspectors witnessed several shifts of fuel handling during the. core off-load. Staffing during off-load was in accordance with technical specifications. Good house-keeping practices were observed in the fuel handling area and contain-ment integrity was maintained as required by technical specifications.
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On May 4, 1982, while removing assembly J-67 from the core the oper-ator received an overload alarm. After the assembly had been trans-ferred to the spent fuel pit a detailed inspection was done and it was found that a corner piece of the grid strap which is third from the top was missing. A search for the missing piece of grid strap.
was conducted all along the path the assembly had taken and on May 6, 1982 an approximately 2 inch piece of the strap was found in the transfer canal on the spent fuel pit side.
It was determined that this was only part of the missing piece. During decontamination of the transfer canal on the containment side a second piece about 4 inches long was found.
It was determined that these two pieces make up the missing section from the fuel assembly.
After discussions with Westinghouse the licensee decided not to return assembly J-67 to the core and Westinghouse redesigned the core load
to accomodate a low burnup assembly that had been removed during a previous refueling.
At 7:13 p.m. on May 4, 1982, while removing the burnable poison rod assembly handling tool from fuel assembly F-61 in the spent fuel pit, the guide pins on the tool apparently became bound to the fuel assembly nozzle block and the fuel assembly was inadvertently lifted approxi-
mately 4 feet. At this point the spent fuel pit operators observed
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that the fuel assembly was being lifted and stopped their operation.
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On stopping the hoist the assembly dropped 4 feet back into the spent fuel racks. A precautionary auxiliary building evacuation was accom-plished. Health Physics Personnel took air samples in the area and a spent fuel pit water sample was obtained. All sample results indi-cated normal activities and the evacuation was terminated.
Examination of the burnable poison Rod Assembly Handling Tool revealed some damage to the guide pins which had to be repaired before any further steps involving the tool could be accomplished.
Inspection of the dropped fuel assembly revealed that, although most of the fuel rods in the assembly had been displaced downward, very few actually made contact with the bottom nozzle block. Assembly F-61 was not scheduled to be returned to the core.
10.
Containment Integrated Leak Rate Test The resident inspectors, in conjunction with a Region III inspector, monitored activities associated with the Unit 2 containment integrated leak rate test.
Included in the inspection were reviews of the procedure for technical adequacy, determination that the procedure was available and being used by test personnel, verification that the special test equipment required by the procedure was calibrated and in service, verification of initial conditions and system lineups, monitoring of key parameters during the test, and verification that post-test blowdown was within technical specification radioactivity release limits. The resident inspectors also conducted an independent post-test containment inspection for equipment degradation.
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l 11.
Inspection During Long Term Shutdown
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l The inspector monitored Unit 2 shutdown operations, reviewed applica-l ble logs,-
and conducted discussions with operators during April and
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May 1982. Tours of Unit 2 areas were conducted to verify tagout procedure implementation and to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibra-tions. Also monitored were housekeeping / cleanliness conditions, impicmentation of radiation protection controls, and implementation of the station security plan.
On several occasions while in transit to the Unit 2 containment, the inspector found the fire doors to the stairwell adjacent to the Unit i facade elevator on the 26 and 66 foot levels to be open. After questioning licensee personnel and finding no apparent reason for the doors being left open, the matter was brought to the attention of licensee management.
12.
Independent Inspection a.
In a recent daily report an event was documented wherein chemical stratification in the Refueling Water Storage Tank (RWST) led to erroneous conclusions concerning boron concentration in the RWST and resulted in an RCS dilution upon injection during plant shut-down. The inspectors reviewed this item for applicability at Point Beach and determined the following. At Point Beach, the RWST's are sampled every two weeks. Prior to sampling, the tanks are placed on a two day recirculation using the refueling water circulating pump. During this recirculation, approximately 61% of the RWST contents are pumped from the bottom to the top of the tank.
Recirculation prior to sampling is enforced through a routine call up system.
Recirculation appears on the routine call up list on alternate Tuesdays.
Sampling appears on the following Thursdays.
In addition to identifying the above items, the review also de-termined that the in-place procedures for returning the Residual Heat Removal System (RHR) to normal do not include provisions for verifying boron concentration in either the RHR or RWST.
The procedure does require that the RHR be recirculated with the RWST to establish a 2000 ppm boron concentration in the RHR as called for in safety analyses. This was brought to the attention of the licensee on~ April 7, 1982. The licensee agreed to review the matter and make a determination as to whether such sampling should be procedurally required.
b.
On April 9, 1982, the licensee informed the inspectors that review of the test data for the loss of voltage protection relays for the 4.16 KV safeguards busses revealed that the test results were outside Technical Specification limits.
Specifically, Technical Specification 15.3.5. A calls for a trip time delay
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cf 0.3 seconds plus or minus 5% when bus voltage falls to zero.
Measured time delays ranged from 0.317 seconds to 0.35 seconds.
Licensee investigation of the discrepancy revealed that the Technical Specification limit, submitted as part of a Technical Specification change request in 1977 and approved in 1981, was in error. The 0.3 second limit was extracted from the relay characteristic curve. However, the relay operates with a nominal minimum trip point of 20 cycles or 0.33 seconds. Thus, the limit submitted and approved was not consistent with the operation of the relay installed.
Based on this information, a conference call between the licensee, the Resident Inspector, and a representative of the Office of Nuclear Reactor Regulation (NRR) (C. Trammel) was held on April 9, 1982. The specifics of the problem were discussed.
In addition, the licensee presented information which showed that the installed relays were operating properly in accordance with their calibra-tion curve. Further, the licensee committed to begin preparation of a Technical Specification change on April 12, 1932 to correct the error and to submit a 14-day LER documenting the details of the problem. Based on the information presented and the commit-ments made, NRR concurred in continued operation with the existing relay setpoints.
c.
On April 14, 1982, the licensee informed the inspector that a routine procedure review by the instrument and control super-intendent had identified rhat the procedurally prescribed set-point for reactor protection system bistable P-6 was noncon-servatively incorrect. This bistable is designed to automatically energize the source range instruments and enable the source-10 range high flux reactor trip to function at 10 amperes on the intermediate range on a reactor startup.
P-6 setpoints are pre-scribed in the Technical Specification Limiting Safety Systems Settings and Limiting Conditions for Operation. The procedure l-10 for setting P-6 requires that the bistable trip at 10 amperes under such conditions to allow blocking the source range high flux trip. The reset point, that point which energizes the source range and enables its trip, is determined by the minimum bistable band width. This results in a nominal setting of 8 x-11
amperes. The result is that the source range trip function, as described in the Final Safety Analysis Report, is not made available until two tenths of a decade below the required value.
The licensee, in response to this situation, and after performing a review pursuant to 10 CFR 50.59, has established administra-tive controls to ensure that the source ranges are energized and-10 the trip function is enabled at 10 amperes on the intermediate range. Also, other bistable setting procedures have been re-viewed to determine if they suffer from similar deficiencies.
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d.
Based on the incident at Ginna the licensee decided to have Westinghouse perform a routine annular search of the steam generator tube bundles with fiber optics. On May 8, 1982 an object which appeared to be a welding rod was found on the hot leg side of the "A" steam generator. The bottom of the rod was not visible from the angle of inspection but it appeared to be standing vertically with its upper end resting against one of the outer periphery tubes.
The initial attempts to remove the rod were unsucce=sful in getting the bottom end to break free. However the top was pulled loose from the tube it was resting against and the pulling on the rod had bent it in about the middle and left it laying on the tube sheet. The inspector viewed the rod at this point and concurred with the licensee in that it appeared to be a welding rod and that it had in fact been in contact with one of the tubes about 12 inches above the tube sheet. This was indicated by an obvious discoloration of the tube at the point of contact.
Westinghouse then brought in a pneumatically operated gripper and on May 12, 1982 the rod was removed.
Inspection verified that it was in fact a 14 inch welding rod which appears to have been corroded to the tube sheet at one end and to the tube at the other end.
The last work involving welding in the steam generator was done in 1975. A reevaluation of the eddy current results for all tubes in the vicinity of the rod was done with no abnormalities noted. No other objects were found in either steam generator.
e.
The licensee disassembled the.four core deluge check valves for Unit 2 to inspect for the possibility of lockwire binding.
This was a commitment made in July 1981 inspection reports No. 50-266/81-13; 50-301/81-15. Three of the valves had un-secured lockwires as was found in Unit 1.
These were repaired.
One valve had its lockwire secured as per the valve drawing.
13.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) throughout the month and at the conclusion of the inspection and summarized the scope and findings of the inspection activities. The licensee acknowledged the findings.
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