IR 05000454/1987028
| ML20237K760 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/18/1987 |
| From: | Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20237K733 | List: |
| References | |
| 50-454-87-28, 50-455-87-26, IEIN-87-004, IEIN-87-012, IEIN-87-12, IEIN-87-4, NUDOCS 8708270287 | |
| Download: ML20237K760 (17) | |
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U. S. NUCLEAR' REGULATORY COMMISSION
REGION III
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' Report Nos. 50-454/87028(DRP);50-455/87026(DRP)
Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2 Inspection At: Byron Station, Byron, Illinois Inspection Conducted: July 1 - 31, 1987 Inspectors:
P. G. Brochman L. N. 01shan
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Approved B -
M. Hinds, hief 81887
Reactor Projects Section IA Date l
Inspection Summary Inspection from July 1 - 31, 1987 (Report Nos. 50-454/87028(DRP):
50-455/87026(ORP))
Areas Inspected:
Routine, unannounced safety inspection by the resident
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inspector and a headquarters inspector of licensee action on previous
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inspection findings; LERs; operations summary; training; surveillance; maintenance; operational safety and ESF walkdown; startup testing; event
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followup; and' allegations.
Results: Of.the nine areas inspected, no violations or deviations were l
identified in eight areas.
One violation resulting in a Notice of Violation
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was identified in the remaining area (failure to follow approved procedures
- Paragraph 3.b).
This violation was of more than minor safety significance and indicated that multiple individuals were failing to follow administrative procedures for the control of equipment status.
This violation is also indicative of the station management's inadequate resolution of problems which had been previously identified by the licensee's quality assurance program.
One additional violation (failure to follow Technical Specifications -
Paragraph 3.a) was identified in this area; however, in accordance with 10 CFR 2, Appendix C, Section V.A, a Notice of Violation was not issued.
8708270297 870020 PDR ADOCK 05000454 G
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DETAILS'
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Persons Contacted Commonwealth Edison Company L
R. Querio, Station Manager
- R. Pleniewicz,. Production Superintendent
- R. Ward, Services Superintendent
- W. Burkamper, Quality Assurance Superintendent
- L. Sues, Assistant Superintendent, Operating j!
G. Schwartz, Assistant Superintendent, Maintenance
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- T. Joyce, Assistant Superintendent, Technical Services J
D. St. Clair, Assistant Superintendent, Work Planning i
T. Higgins, Operating Engineer. Unit 0
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J. Schrock, Operating Engineer, Unit 1 0. Brindle, Operating Engineer, Unit 2 T. Didier, Operating Engineer, Rad-Waste
- M. Snow, Regulatory Assurance Supervisor
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- F. Hornbeak, Technical Staff Supervisor
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- R. Flahive, Radiation / Chemistry Supervisor
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P. O'Neil, Quality Control Supervisor
- A. Chernich, Training Supervisor
- E. Zittle, Regulatory Assurance Staff
- A. Britton, Quality Assurance Inspector
- D. Berg, Nuclear Safety
- *D. Flowers, Inservice Inspection Coordinator The inspector also contacted and interviewed other licensee and
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contractor personnel during the course of this inspection.
- Denotes those present during the exit interview on July 31, 1987.
2.
Action on Previous Inspection Findings (92701 & 92702)
a.
(Closed) Open Item (454/86010-05(DRP)):
Problems with the Diesel Generator (DG) isochronous relays. Within the last two years, there were four instances of erratic DG operation when a DG was operated in the speed droop mode.
In each case, due~to the large fluctuations in DG load, the diesel was automatically or manually tripped. The licensee's investigation determined that the problem was due to high contact resistance on isochronous relays.
The licensee replaced the 1A and 18 DG isochronous relays (31M Agastat)
with General Electric Model #12HFA151A2H relays under modification M6-1-86-082.
The inspector reviewed the modification package and verified that the new relays had been installed and that post-maintenance testing had been completed satisfactorily.
Based on this corrective action, this item is considered closed, i
b.
(Closed) Violation (454/87017-05(DRP); 455/87016-05(DRP)):
Failure to include the shiftly channel check of the pressurizer pressure channels in the Mode 3 daily and shiftly surveillance. The i
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'E inspector reviewed the licensee's response and verified tnat the corrective actions had been accomplished as stated. The-inspector verified that.the shiftly surveillance.for both units, 1805 0.1-1,2,3 and 2B0S 0.1-1,2,3, had been revised to require that a r
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charnel check be performed for the pressurizer pressure channels whi W in Mode 3.
Based on.this corrective action, this item is con.idered closed.
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c.
(Closed) Violation (454/87017-06(DRP); 455/87016-06(DRP)):
Failure to follow procedures during the shutdown of the DGs..The inspector-
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had discovered three instances in which the arming switch for the auto-reclose circuit for-the DG output circuit breaker, was not returned to the normal position after the DG was shutdown. The l
inspector reviewed the licensee's response and verified that the j
corrective actions had'been accomplished as stated. The inspector reviewed the permanent changes to the DG shutdown procedure, BOP DG-12, to verify that clearer guidance was provided to the operators.
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Additionally, a requirement to independently verify the position of
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the switch was added to the procedure..The inspector also verified l
that caution cards had been placed on the control switches and that
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training was completed for the licensed operators.
Based on the s
corrective actions taken, this item is considered closed.
d.
(Closed) Violation (454/87017-07(DRP)):
Failure of a fire watch to have a readily accessible fire extinguisher near welding activities l
in the Unit I containment. The inspector reviewed the licensee's
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response and verified that the corrective actions had been accomplished as stated. The inspector interviewed contractor personnel and verified that meetings had been held and additional guidance provided on the requirements for fire extinguishers and fire watches near cutting, grinding, or welding activities. Based on the corrective actions taken, this item is considered closed.
-I The inspector will review the performance of contractor personnel
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during subsequent outages.
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(0 pen) Open Item (455/86046-02(DRP)):
Freeze protection surveill-ance does not verify that the heat tracing for the Unit 2 condensate storage tank level transmitter is energized. The licensee has not yet completed the revision to the freeze protection' surveillance, Byron Operating Surveillance OBOS AFT-A1; consequently, this item will remain open. Additionally, during discussions with the licensee's staff, the inspector noted that the heat tracing circuit for the Unit 2 refueling water storage tank (RWST) vent piping was also not verified to be energized in OBOS AFT-A1, but the Unit 1
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RWST vent piping heat tracing was verified to be energized. The licensee agreed to revise the BOS to include a verification of the (.
heat tracing for the Unit 2 RWST vent piping.
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3.:
Licensee Event Report (LER) Followup (92700)'
(Closed) LERs (454/85027-3L;.454/87012-LL; 455/87008-LL; 455/87009-LL;
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-455/87010-LL): Through direct observation, discussions with licensee personnel,-and review of records,.the following LERs were reviewed to determine that the deportability requirements were culfilled,'immediate corrective action was accomplished,. and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
LER'No.
Title Unit 1 454/85027-03 Failure of main steam isolation valve to close due to the failure of an air line check valve.
454/87012 Two trains of the component cooling system inoperable due to loss of water inventory caused by a personnel error.
Unit 2 455/87008 Two. lines of containment mini-purge system simultaneously open due.to a personnel error and an inadequate procedure.
455/87009 Manual reactor trip due to the inability to control steam generator levels following the loss of a main feedwater pump.
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455/87010 Manual reactor trip due to the loss of fluid in the EH system due to a pipe failure at an EH system filter.
The events described in LER 455/87010 are discussed further in paragraph 12.b.
a.
With regard to LER 455/87008, this LER discusses an event from I
June 6 - 8, 1987, with Unit 2 in Mode 4, in which the containment mini-purge supply and exhaust ventilation (VQ) system was placed in operation contrary to Technical Specifications. Technical Specification 3.6.1.7.b requires that the eight-inch containment purge (mini-purge) supply and exhaust isolation valves be open no
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more than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per calendar year and that no more than one line be open at a time, while in Modes 1, 2, 3, and 4.
i At 1204 on June 6, 1987, the radiation chemistry department requested that a purge be conducted of the Unit 2 containment to reduce iodine levels. This request was reviewed by the shift control room engineer (SCRE [SRO licensed control room supervisor]), who interpreted the Technical Specification requirement for no more than one "line" to mean no more than one " train." However, only l
one train of VQ mini-purge system exists. After obtaining approval
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from the Shift Engineer (SE [SRO licensed shift supervisor]) to commence a containment' release, the SCRE directed a reactor operator to start the VQ system. The operator started the VQ system in accordance with procedure BOP VQ-6, " Containment Mini-Purge System Operation"; however, he overlooked a caution note, before step F.6, that both supply and exhaust flow paths may not be used simultaneously,.
while in Modes 1, 2, 3, and 4..No reference of this requirement was made in the precautions or limitations and actions sections of the 80P.
As a consequence of these errors, both the supply and exhaust. lines of the'VQ mini purge system were opened.
At 0819 on June 8, 1987, a technical staff engineer discovered the incorrect alignment during a routino system walkdown, and operating department valves were restored to their proper positions.
Both lines were open for approximately '44.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. As corrective action, a temporary change was issued to procedure B0P VQ-6,'and a permanent change is being made to provide more explicit guidance about opening both the supply and exhaust lines in Modes 1, 2, 3, and 4.
This event.has been included in the licensed operator training program.
Disciplinary action was taken against the SCRE and the reactor operator.
The failure to maintain isolation of at least one of the VQ mini-purge paths is a violation of Technical Specification 3.6.1.7.b (455/87026-01(DRP)).
However, this violation meets the tests of 10 CFR 2, Appendix C, Section V.A.; consequently, no Notice of Violation will be issued, and this matter is considered closed.
b.
With regard to LER 454/87012, this LER discusses an event on April 8, 1987, with Unit 1 in Mode 6, when both trains of the component cooling water (CC) system were inoperable due to the.
improper' operation of a valve.
The 1A residual heat (RH) removal exchanger had been taken out of service (005) for gasket replacement.
The CC side of the heat exchanger was drained, and one of the points of isolation for this work was valve ICC9412A, "RH heat exchanger outlet isolation valve." Out-of-service cards were hung on the valve handwheel and on the circuit breaker for the valve motor operator.
On April 6, 1987 a work supervisor (utility non-licensed) requested that the
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00S card hung on the handwheel be temporarily lifted to flush the
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gearbox of the valve's motor operator and to replace the grease.
An operating department shift foreman (licensed SRO) approved this request with a verbal agreement that work was only to be performed on the motor and that the valve was not to be stroked open for any reason. At approximately 1725 on April 8,1987, the contractor maintenance crew performing the work on the valve motor stroked the valve partially open to release the torque in the motor gear train.
The work crew was subsequently interviewed, and its members stated
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that they had received verbal arnroval to stroke the valte from the shif t engineer's of fice;. however, they could not identify 'the specific individual who approved their actions. The personnel in the shift engineer's office stated they had not given permission to the_ work crew to stroke the valve.
The licensee was unable to reconcile these two statements.
Valve ICC9412A connects the outlet of the RH heat exchanger to the CC pump suction header.
The CC surge tank is also connected to the CC pump suction header.
Consequently, when valve ICC9412A was partially opened the contents of the CC surge tank drained'to the empty RH heat exchanger and caused the 1A CC pump to trip on low surge tank level. The low level in the CC surge tank also would have prevented the IB CC pump from starting; however, it was already out of service for unrelated reasons. Therefore, both trains of the CC system were inoperable, which prevented reactor core decay heat from being transferred to the ultimate heat sink.
An equipment operator, dispatched to investigate, discovered the cause of the loss'of surge tank level and shut discharge throttle valve ICC9507A (a butterfly-type valve, not typically used for isolation) to the RH heat exchanger, which stopped the loss of water from the CC system. The CC surge tank was then refilled and the 1A CC pump restarted.
Both trains of CC were inoperable for 17 minutes; however, residual heat removal system temperature did not change by more than 5 degrees during these 17 minutes.
The inspector met with operating and maintenance department managers to discuss this event.
The current procedure which controls these activities is BAP 331-1 " Administrative Requirements for Temporary Lifting of 00S cards and/or piccing Equipment in Test," which states that temporary lifts are intended to allow testing of equipment following maintenance.
The inspector expressed a concern to licensee management that removing the 00S tag on valve ICC9412 and.
imposing a verbal requirement to maintain it in the closed condition appeared to be inimical to an operating philosophy which requires written controls over changes in equipment status. The inspector asked if, as a matter of policy, the SRO who approved the temporary lift needed to consider what would be the consequences on the plant, if the valve were to move or moved, either deliberately or inadver-tently.
Licensee management was unable to provide evidence of a formal policy on this subject or to provide any assurances that any informal policies were being consistently implemented by SR0s.
Additionally, the licensee's policy on performing work on components which are being used as physical isolation points for other work activities is not clear.
Consequently, the licensee is requested to provide a written response which clarifies its policies and how they are implemented and disseminated.
Submission of this response will be followed as an unresolved item (454/87028-01(DRP);
455/87026-02(DRP)).
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During a review of the completed temporary lift record sheets,. BAP 330-1T11, for the month of April 1987,.the inspector identified several documents that were missing required approval or concurrence signatures. BAP 330-1, " Station Equipment Out-of-Service Procedure," was the controlling procedure during April 1987, and paragraph C.6.a(2) required that m erence be obtained from thu individual (or his supervisor) 1.no requested'the original 00S; otherwise, the temporary lift may not be authorized.
Paragraph C.6.b(2) required that the. Shift Engineer or the Shift Foreman
'(licensed SRO) shall authorize the temporary lift only after he
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is satisfied that plant conditions allow'this activity, that no 00S holder restrictions have.been violated, and that all concurrences have'been.made. After the temporary lift is approved, a package is assembled which includes the 005 form, the master'00S card, and the temporary lift record sheet.
Paragraph C.6.b(5) requires that;the control room supervisor (licensed SRO) review the package and then issuc it to the applicable unit operator (licensed RO) for imple-mentation.
For the temporary lifts which had been completed in April 1987 (approximately 200), ten record sheets were identified as missing required signatures.
Seven of these sheets did not have concurrence
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initials for.,the~ individuals who had originally requested the 00S.
Eight of these sheets were missing the Shift Engineer or Shift Foreman authorization signatures.
Five of these sheets were missing both signatures. The forms were dated March 32, April 7, 8, 9, 13, 25 and 30, 1987.
Discussions with the quality assurance (QA) supervisor indicated that the QA department had twice previously (November 1986 and mid-April 1987) identified problems with missing signatures on the temporary lift record sheets to station management, as part of surveillance that the QA department had performed.
10 CFR 50, Appendix B, Criterion V, as implemented by Commonwealth Edison Company's Quality Assurance Manual, Quality Requirement 5.0, requires that activities affecting quality shall be accomplished in accordance with approved instructions, procedures, and drawings.
The control of equipment status is an activity affecting quality.
The failure to obtain the necessary concurrences from the individuals who initially requested the 00S cards for seven temporary lift record sheets, prior to the temporary lifting of the 00S cards, is a violation of 10 CFR 50, Appendix B, Criterion V (454/87028-02a(DRP); 455/87026-02a(DRP)).
The failure to obtain the necessary SRO approval signatures for eight temporary lift
record sheets, prior to the temporary lifting of the 005 cards, is j
a violation of 10 CFR 50, Appendix B, Criterion V (454/87028-02b(DRP);
455/87026-03b(DRP)).
In the response to this violation the licensee is requested to j
address three additional questions:
(1) why did the the control
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room supervisor tail to detect these missing signatures and permit the temporary lift to proceed without the required approvals; (2)
why did the unit operator perform the temporary lift when the required signatures were not present on the record sheets; and (3)
why did this problem recur after it had been previously identified by QA department personnel.
No violations or deviations were identified with the other LERs.
4.
Summary of Operations Unit 1 operated at power levels up to 98% until 2211 on July 29, 1987, when a reactor trip on power range negative high flux rr.te occurred due to a lightning strike (see paragraph 12.e). The reactor was taken critical at 1641 on July 30 and synchronized to the grid at 2201 on the same day.
During power ascension an identical reactor trip caused by a lightning strike occurred at 0153 on July 31, with reactor power at 30%
(see paragraph 12.f). The unit remained shutdown for the rest of the report period. An Unusual Event was declared at 1005 on July 13, 1987, when a shutdown required by Technical Specifications was commenced.
The Unusual Event was terminated at 1055 prior to the completion of the shutdown (see paragraph 12.c).
Unit 2 operated at power levels up to 35% until 0628 on July 1,1987, when the reactor was manually tripped due to an electro-hydraulic (EH) system leak (see paragraph 12.b).
The reactor was restarted at 1403 on the same day. However, due to problems with the turbine throttle-stop valves, the unit was shut down at 2147 on July 2.
After the problem with the throttle-stop valves was corrected, the unit was taken critical at 1301 on July 4, and synchronized to the grid at 0115 on July 5.
The unit subsequently operated at power levels up to 100% until 1845 on July 14, when the unit was tripped from 100% power as part of a planned startup test.
Following the test the unit was restarted at 0832 on July 16, and synchronized to the grid at 1346 on the same day. The unit operated at power levels up to 98% until 1216 on July 25, when a reactor trip on over temperature delta t (0 TDT) occurred (see paragraph 12.d).
The reactor was restarted at 2143 on the same day and synchronized to the grid at 0118 on July 26.
The unit subsequently operated at power levels up to 98% for the rest of the report period.
5.
Training (41400 & 41701)
The effectiveness of training programs for licensed and nonlicensed personnel was reviewed by the inspector during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response to events which occurred during July 1987.
Personnel appeared to be knowledgeable of the tasks being performed, and nothing was observed which indicated any ineffectiveness of training.
No violations or deviations were identified.
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6.
Monthly Surveillance Observation (61726)
Station surveillance activities of the safety-related systems and components listed below were observed / reviewed to ascertain that they
.were conducted in accordance with approved procedures and in conformance with Technical Specifications.
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2A Component Cooling Pump ASME quarterly test 1A Diesel Generator Monthly test 1A Safety Injection pump ASME quarterly test Valve ISI8814 stroke' time testing The following items were considered during this review:
the limiting conditions for operation were met while affected components or systems were removed.from and restored to service; approvals were obtained prior to initiating the testing; testing was accomplished in accordance with approved procedures; test instrumentation was within its calibration
' interval; testing was accomplished by qualified personnel; test results conformed with Technical Specifications and procedural requirements and were reviewed by personnel other than.the individual directing the test; and any deficiencies identified during the testing were properly documented, reviewed, and resolved by appropriate management personnel.
The inspector reviewed the surveillance records for the 1A Diesel generator and the 1A and IB Contai.. ment Post-LOCA (Loss of Coolant Accident) Hydrogen monitors to verify that surveillance were being l performed within their required intervals.
No violations or' deviations were identified.
7.
Monthly Maintenance Observation (62703)
Station maintenance activities of the safety-related systems and components listed below were observed / reviewed to ascertain that they.
were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications.
Modification of the IB Diesel Generator input to the ESD system Following completion of maintenance on the diesel generator, the inspector verified that it had been returned to service properly.
The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from and restored to service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prier to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire
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prevention controls were implemented. Work requests were reviewed to I
determine the status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performance.
No violations or deviations were identified.
8.
Operational Safety Verification and Engineered Safety Features System i
Walkdown (71707, 71709, 71710, & 71881)
The inspector observed control room operation, reviewed applicable logs and conducted discussions with control room operators during July 1987.
During these discussions and observations, the inspector ascertained that the operators were alert, cognizant of plant conditions, and attentive to changes in those conditions, and that they took prompt action when appropriate.
The inspector verified the operability of selected emergency systems, reviewed tagout records and verified the proper return to service of affected components. Tours of the auxiliary, fuel-handling, turbine, and rad-waste 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 verified by observation and direct interviews that the physical security plan was being implemented in accordance with the station security plan.
The inspector reviewed the following security event report received during July 1987. At approximately 1022 on July 2, 1987, suspected controlled substances were found during the search of a vehicle prior to its entry into the protected area.
The vehicle was a tractor trailer for a trucking firm.
The vehicle and its occupants were denied access to the site. The suspected controlled substances were transported to the local law enforcement agency, where testing indicated the presence of marijuana.
The driver and his helper have been permanently barred from access to the site and the trucking firm was notified of this event.
The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. The inspector also witnessed portions of the radioactive waste system controls associated with rad-waste shipments and barreling.
During the month of 1987, the inspector walked down the accessible portions of the 2A Safety Injection (SI) system to verify operability.
During the walkdown, the
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inspector obsersed a large puddle of water on the floor of the 2A SI pump I
room.
The water had come from a packing leak on valve 2SI8814. A review of the outstanding work requests indicated that a work request had been written on March 13, 1987, to repair the leak. The inspector expressed i
concern to licensee management over the length of time that an identified
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leak in a contaminated system had gone unrepaired.
Facility operations observed were verified to be in accordance with the requirements established under Technical Specifications, 10 CFR, and administrative procedures.
No violations or deviations were identified.
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9.
Startup Test Witnessing and Observation (72302)
L The inspector witnessed. performance of portions of the following Unit 2 startup test-procedures in order to~ verify that testing was conducted in accordance with the operating license and procedural requirements,.that test date was properly recorded, and that performance of licensee personnel. conducting the tests demonstrated an understanding of assigned duties and responsibilities.
Turbine / Reactor Trip from 100*4 power No violations'or deviations were identified.
10.
Followup of Headquarters Requests-(92701)
a.
NRC Information Notice (IN) 87-12 discussed potential problems with General Electric Type AFK-2-25 circuit breakers failing to open on demand. A review of licensee records indicates that this
. type of breaker is not used at Byron; consequently, this IN is not specifically applicable. However, the licensee has applied the recommendations given in the IN to other circuit breakers in the plant. Based on this. review, the inspector believes that the licensee has satisfactorily addressed the concerns of IN 87-12.
b.
NRC IN 87-04 addressed a problem pertaining to long-term storage of fuel for diesel engines for emergency service.
Problems have been identified in the industry with the plugging of diesel fuel oil filters and strainers due to the growth of microorganisms and generation of sediment.
The inspector reviewed the licensee's program for preventing plugging of fuel oil filters and strainers and verified that surveillance are performed every eighteen months to inspect the strainers and filters for diesel-driven emergency equipment. ' For the diesel generators and auxiliary feedwater pump diesel, the differential pressure across both strainers and filters are measured and displayed on the local control. panel.
High differential pressure across either strainers or filters will-actuate an alarm on the local control panel.
Based on this review, the inspector concludes that the licensee has satisfactorily addressed the concerns in IN 87-04.
No violations or deviations were identified.
11.
Follow p of Region III Requests (92701)
The inspector received a request from Region III in a memo from C. E. Norelius, dated June 26, 1987, to review the licensee's program for monitoring ambient temperatures of areas which contain electrical equipment and instrumentation.
Some electrical equipment and instrumentation is affected by high temperatures, which can result in
accelerated aging, setpoint drift, or equipment failure.
The inspector
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met with the licensee's staff to discuss the licensee's program.
Technical Specification 3.7.12 requires that the ambient temperatures of
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13 areas, which contain safety-related equipment, be controlled to specified limits.
If the limits are exceeded, LCOs (Limiting Conditions for Operation) are entered. Surveillance requirements specify that these areas be checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and that the temperatures be recorded in the shiftly surveillance B05 0.1-1,2,3 every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
These surveillance records are reviewed by the licensee's technical staff to check for trends and abnormal operations.
These requirements and procedures appear to establish an adequate program for monitoring ambient temperatures.
No violations or deviations were identified.
12. Onsite Followup of Events at Operating Reactors (93702)
a.
General The inspector performed onsite followup activities for events which occurred during July 1987. This followup included reviews of operating logs, procedures, Deviation Reports, Licensee Event Reports (where available), and interviews with licensee personnel.
For each event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions, and reviewed licensee actions to verify consistency with procedures, license conditions, and the nature of the event. Additionally, the inspector verified for each event that the licensee investigation had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restart.
Details of the events and licensee corrective actions developed through inspector followup are provided in Paragraphs b through f below, b.
Unit 2 - Manual Reactor Trip on July 1,1987 (LER 455./87010)
At 0605 on July 1, 1987, with reactor power at 35%, the control room operators received an EH reservoir low level alarm.
The EH system provides the hydraulic pressure to operate the main turbine governor and throttle-stop valves.
The reservoir stores EH fluid to supply EH pumps. A shift foreman (licensed SRO) was dispatched to the scene and reported that EH fluid was spraying out of a pipe at a fitting upstream of the Fullers earth filters.
The reactor operators began to ramp down the turbine, and maintenance personnel were requested to tighten the fitting. At 0628 the pipe failed completely, and the shift foreman directed the control room to secure the EH pumps, which were beginning to cavitate, and to manually trip the reactor.
Reactor power was at approximately 20%.
With the EH pumps secured, the turbine would have eventually tripped
on low EH header pressure, causing an automatic reactor trip.
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unit was stabilized in Mode 3 and all systems responded normally after the trip.
The licensee's investigation determined that a 360-degree circumferential failure had occurred in a piece of thick walled tubing at the Fullers earth filters on the EH pump skid.
The crack
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was immediately adjacent to the ferrel for a union fitting.
Metallurgical examination of the failed tubing determined that the cause was due to fatigue failure due to rigid supports. The licensee has isolated the filter until the supports can be redesigned. The licensee inspected the piping in Unit 1 and at s
Braidwood and did not identify any cracks in the same area, and also verified that the piping configuration and supports were different from those in Byron Unit 2.
The reactor was restarted at 1403 on July 2.
However, the licensee I
experienced problems with the surveillance test for the turbine throttle-stop valves, which did not stroke within the required time limit.
The unit was shut down at 2147 on the same day.
Following discussions between the licensee and the manufacturer, Westinghouse, the licensee modified the test procedure t., ensure that the valves were at thermal equilibrium before performing the test. After modifying the procedure, the licensee successfully tested the valves, j
and the unit was taken critical at 1301 on July 4 and synchronized to the grid at 0115 on July 5.
c.
Unit 1 - Unusual Event Declared on July 13, 1987 At 0800 on July 13, 1987, with reactor power at 98%, a surveillance on the solid-state protection system (SSPS) failed. The failed
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light illuminated on the logic test panel for switch C, position 5, when the test was performed.
Pressurizer low pressure was the particular function which was being checked. With this function
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inoperable, Technical Specifications 3.3.1 and 3.3.2 require that the function be restored to an operable condition or that the
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reactor be in Hot Standby (Mode 3) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
At 1005 the licensee declared an Unusual Event and began to reduce turbine load to place the unit in Mode 3.
By 1051 the licensee had located a loose wire on switch C.
The wire was reconnected and the surveillance successfully performed.
The licensee terminated the Unusual Event at 1055 prior to the completion of the shutdown and returned the unit to its rated output. The licensee is investigating the cause of the loose wire, and the inspector will review the licensee's report after it is issued.
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d.
Unit 2 - Reactor Trip on July 25, 1987
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At 1216 on July 25, 1987, with reactor power at 98%, a reactor trip on over temperature delta T (0 TDT) occurred during the performance of a 30% load rejection test.
The licensee was performing a test, as part of the startup program, to verify the ability of the unit to perform a 30% load rejection test (SPP 87-80) from 100% power at a rate of approximately 200'e per minute, without causing a reactor trip.
Approximately one minute into the transient, a reactor trip on OTDT was received.
All systems functioned normally after the trip.
However, a pressurizer relief valve did open for 9.12 seconds.
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The unit was stabilized in Mode 3.
The licensee had previously completed a 30% load reduction on July 11, which did not result in a reactor. trip.
The licensee was not'able to identify the specific cause of the L
reactor trip, but believes that the problem lies with the digital electro-hydaulic control (DEHC) system for the turbine.
To complete
.the testing of the DEHC' system and also to check out the steam dump valves, the licensee had to start up the unit.
The reactor was taken critical at 2143 and synchronized to the grid at 0118 on July 26.. The licensee limited power to 50% until the problem could be solved. The' licensee subsequently performed testing on the steam dump ' valves to verify that they were operating properly. Adjust-ments were made to the current to pressure converters and the volume boosters. The valves were stroke time tested and verified to travel their full strokes, under flow, without binding.
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Following~ discussions with Westinghouse, the licensee gradually increased power to 98%, with strip chart recorders installed to-monitor various parameters. Governor. valve surveillance tests were performed at 80% and 90% power.
Power was then reduced to approximately 60%, and load reductions of a smaller size and lesser-rate were performed while recording DEHC, turbine, and steam dump performance. At the end of this report period, the information collected during this testing had been forwarded to Westinghouse for review and evaluation.
After the end of the report period, recommendations were received from Westinghouse on changes to the DEHC program. After installing the changes in the DEHC program and completing tuning of the control loops for various secondary plant components, the licensee performed load rejection tests at various rates and of various power changes to verify the acceptability of the Westinghouse recommenda-tions.
Following completion of these troubleshooting and tuning efforts, the licensee successfully completed a 50% load rejection test on August 12, 1987. The inspector witnessed performance of l
the test and will review the LER in a subsequent report.
e.
Unit 1 - Reactor Trip on July 29, 1987 At 2211 on July 29, 1987, with reactor power at 98%, a reactor trip on power range negative high flux rate occurred.
i The licensee's investigation discovered that nine of ten circuit l
breakers were tripped on the power supplies for the five rod drive l
cabinets. With these power supplies deenergized, all of the control i
rods were unlatched and fell into the core.
The control rods falling into the core caused a negative high flux rate reactor trip on the power range nuclear instruments.
The licensee believes that a lightning strike which hit the station caused a power surge on the
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ground grid, and this surge tripped the power supply breakers.
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review of records indicated that a spike had occurred on the ground j
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alarm for DC bus 112'350 msec before the reactor trip. The spike was approximately 14 msec long. The lightning strike also tripped
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the prime and process computers. All other equipment functioned f.
normally after the trip, and the unit was stabilized in Mode 3.
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.Th'e licensee inspected the power supplies and other plant equipment j
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.which had previously.been; susceptible to damage from lightning
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strikes; no damage was identified. The prime and process computers
- were restarted. The licensee-established a task force to examine ways to reduce the-susceptibility of.the rod drive system.to lightning strikes. The licensee reset the power supply breakers and the reactor was taken critical at 1641 on July 30 and synchronized to the grid at 2201 on the same day.
f.
Unit 1 - Reactor Trip on July 31, 1987 At 2211 on July.29, 1987, with reactor power at 30% during the return to rated power, a reactor trip on power-range nuclear instrument' negative high flux rate occurred. This trip was identical.to the trip on July 29; however, only three power supp1y.
breakers opened. Nevertheless, a sufficient number.of control rods
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fell into the core to cause the negative flux rate trip. All other
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equipment functioned normally after the trip, and the unit was -
stabilized in Mode 3.
The unit remained shutdown at the end of this-report period pending a resolution of this problem. The inspector will review the two lightning-induced trips in a subsequent report after the LER is issued.
No violations or deviations were identified.
13. Allegations (99014)
(Closed) Allegation (RIII-87-A-0017): An anonymous allegation was received at the Region III office on February 24, 1987. The alleger stated that there was a crack in a concrete wall and steel plate in Byron Unit 2 and that it had not been resolved satisfactorily.
The alleger included a drawing which indicated the crack was in a vertical wall near grid location F21 at the 396' elevation.
Plate #3 of structural steel drawing S-1522 was also referenced.
NRC Review: Grid location F21 at the 396' elevation is in the basement
.i of the turbine building, which is a non safety-related, seismic category II structure.
Consequently, the NRC forwarded this concern to the licensee in a letter from C. E. Norelius to C. Reed, dated April 14, 1987, for review and evaluation.
The licensee's response, in a letter from K. A. Ainger to A. B. Davis,
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dated May 22, 1987, stated that there was no concrete at location F21 in i
the turbine building. However, drawing S-1522 provides a detail of the secondary shield wall in the Unit 2 containment.
The containment building is a safety-related, seismic Category I structure. A crack was located at the 395'-9" elevation, 80 azimuth of the Unit 2 containment,
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secondary ' shield wall. A field examination by the architect engineer
(Sargent & Lundy [S&L]) determined that the crack was a small horizontal surface defect.approximately 1/4' inch wide by 6 inches long by 1/2 inch
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-deep. The concrete surface'. adjacent to the defect was repaired.
The allegation also indicated a crack in an embedment plate at this i
location.
The field examination by S&L determined that the embedment i
plate had'not cracked, but that two separate plates had not been-completely butted together when the concrete was initially poured.
Additionally, the gap between the embedment plates is about one_ inch above the concrete defect and therefore is not related'to the defect in the concrete. There are no attachments near the edge'of the embedment plates.
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.The licensee has determined, based on the field inspection and review,
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that the structural integrity of the secondary shield wall. is not
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affected by this defect, The. gap between the embedment plates does not affect the integrity of any attachments to the embedment plates.
Conclusion: The allegation was substantiated, in that a defect in the concrete wall did exist; however, the allegation that the embedment plate was cracked was not substantiated.
The defect was repaired, and a subsequent analysis determined that it did not affect the structural integrity of the secondary shield wall or of any attachments to the embedment plates. Based on this review, this allegation is considered closed.
'14.' Violations for which A " Notice of Violation" Will Not Be Issued i
The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requirement. However, because the NRC wants to encourage and support licensees' initiatives l
for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.A.
These tests are:
(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonable time period; and (5)
it was not a violation that could reasonably be expected to have been
. prevented by the licensee's corrective action for a previous violation.
A violation of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued is discussed in Paragraph 3.a.
15. Unresolved Items Unresolved items are matters about which more information is required I
in order to ascertain whether they are acceptable items, items of noncompliance, or deviations. An unresolved item disclosed during the inspection is discussed in Paragraph 3.b.
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16.
Exit Interview (30703)
The inspector met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on July 31, 1987. The inspector summarized the purpose and scope of the inspection and the findings.
I The inspector also discussed the likely informational content of the I
inspection report with regard to documents or processes reviewed by the
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inspector during the inspection.
The licensee did not identify any such documents or processes as proprietary.
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