IR 05000454/1987039
| ML20236S907 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 11/19/1987 |
| From: | Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236S897 | List: |
| References | |
| 50-454-87-39, 50-455-87-36, NUDOCS 8711300171 | |
| Download: ML20236S907 (13) | |
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- e U.S.' NUCLEAR REGULATORY COMMISSION REGION III'
.I Reports No. 50-454/87039(DRP);50-455/87036(DRP)
Docket Nos.. 50-454;.50-455 Licenses.No. NPF-37; NPF-66
' Licensee: -Commonwealth Edison Company Post Office Box 767 Chicago, IL-60690
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Facility Name: Byron Station, Units 1 and 2 Inspection At:
Byron Station, Byron, Illinois Inspection Conducted:
October 2 through 30, 1987
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Inspectors:
P. G. Brochman W. J. Kropp.
J. M. Hinds,'Jr.
T. D. Reidinger R
~A. Wes urgh g
W Nd Approved By:
. M.
nds, Jr., Chief
// //9(j/)
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. Reactor' Projects Section.1A Date Inspection Summary Inspection from October 2 through 30, 1987 (Report Nos. 50-454/87039]DRP);
'50-455/87036(DRP))-
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Areas Inspected:
Routine, unannounced safety inspection by the resident
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- Tnspector and regional inspectors of licensee action on previous. inspection findings,' operations summary,. design changes and modifications, training, surveillance, maintenance, operational safety and ESF walkdown, deviation reports, event followup, and management meetings.
Results:
No violations or deviations were identified.
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DETAILS 1.
' Persons ~ Contacted.
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-Commonwealth Edison Company
- T. Maiman, Vice President, PWR Operations
- R. Querio,. Station Manager
- R. Pleniewicz, Production Superintendent
- R. Ward, Services Superintendent W. Burkamper, Quality Assurance Superintendent
- L. Sues, Assistant Superintendent, Operating
- G. Schwartz, Assistant Superintendent, Maintenance
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T. Joyce, Assistant Superintendent, Technical Services
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D. St. Clair, Assistant Superintendent, Work Planning
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T. Higgins, Operating Engineer, Unit 0 l
J. Schrock, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit 2 j
T. Didier,- Operating Engineer, Rad-Waste
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- M. Snow, Regulatory Assurance Supervisor
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- F. Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor P. O'Neil, Quality Control Supervisor W. Pirnat, Regulatory Assurance Staff
- E. Zittle, Regulatory Assurance Staff
- M. Whitemore, Rad-Chem /GSEP Coordinator
- K. Yates, Nuclear Safety
- A. Britton, Quality Assurance Inspector
- J.
Langon, Regulatory Assurance
- K. Groesser, General Manager, PWR Operations
- W. Roberts, Industrial Relations Manager, Nuclear Operations
- R. Flessner, PWR Operations Staff
- E. Lamken, Quality Assurance, Byron
- W. Dean, Nuclear Safety, Byron The inspectors also contacted and interviewed other licensee and contractor personnel during the course'of this inspection.
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- Denotes those present during the management meeting on October 21, 1987.
- Denotes those present during the exit interview on October 30, 1987.
2.
Action on Previous Inspection Findings (92701 and 92702)
(Closed) Open Item (455/86046-02(DRP)):
Freeze protection surveillance did not verify that heat tracing for the Unit 2 refueling water storage tank (RWST) and condensate storage tank (CST) was energized.
The inspector reviewed Revision 3 to the freeze protection surveillance, OBOS XFT-A1, and verified that provisions were made to verify that heat tracing is energized for the Unit 2 RWST and CST.
Based on this action, this item is considered closed.
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Summary of Operations:
l t 1 operated at power levels up to.98% for the. entire ' report period.
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- Unit 2 began the report period shutdown and-was -taken critical at 9:11 p.m.
on October 1, 1987.
At 4:46 a.m. on October 2, 1987, a turbine trip / reactor trip on'Hi-Hi steam. generator level occurred at 13% power
immediately.after the unit was synchronized.
During the recovery, the
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. disconnect switch for the Unit 2 station auxiliary transformer was
' incorrectly opened.
This caused alloss of offsite power and an. Unusual.
Event was declared. : Offsite power.was restored at 1:12 p.m..The Unusual
~ Event was terminatec' at 2:18. p.m. on the same day.. The reactor was taken critical at 3:55 p.m. on October 3, 1987.
The. unit was synchronized to,
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the grid at 11:47 p.m.
'During synchronization', a generator trip occurred.
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- The unit was again synchronized to the grid at 9:13 a.m. on October 4~,
1987.
At 5:10 p.m. on October 5; 1987, an Unusual Event was declared due to excessive reactor coolant system leakage.
The leak waszlocated, and the Unusual Event was terminated at 6:05 p.m. on the same day.
The unit-
-operated at' power levels up to 95%' for the rest of the report period.
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4.
Design Changes and Modifications (37700).
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The inspectors reviewed the following design changes and modifications to verify that they were accomplished in.accordance with Technical l
Specifications, 10 CFR 50.59, administrative procedures, and the
licensee's. quality assurance program.
j Modification Description
M6-1-87-076 Add inductor and capacitor to Safety Injection (SI)
relays to: avoid / revise problem.
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M6-1-86-007 Change setpoints on condensate booster pump Hi flow alarm.
M6-1-85-201 Replace limit switches on valve ICV 111A.
M6 1-85-508 Revise the diesel generator (DG) jacket water cooling
system based on the engine air intake manifold
temperature.
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M6-1-85-035 Add off gas (0G) metering valves, flowmeters, and pressure regulators.
M6-1-86-110 Install larger operator for essential service water (SX) discharge valves 1SX001A and ISX001B.
M6-1-85-556 Revise sway braces for Unit 1 Residual Heat Removal (RHR) "A" and "B" pumps to reduce vibrations.
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M6-1-85-063 Replace battery chargers for diesel-driven SX makeup pumps.
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M6-1-87-023 Replace splices on reactor vessel resistance temperature detectors (RTDs).
M6-1-85-028 Add new alarm to main generator over-excitation circuit.
M6-1-85-180 Install reactor cavity water level instruments.
M6-1-85-515 Replace limit switches on reactor coolant (RC) and chemical.and volume control (CV) valves.
M6-1-86-108 Reroute SI line ISI08JD.
M6-1-85-195 Install isolation valve'for bus duct differential pressure (DP) switches.
M6-0-85-360 Install new power outlet for dry active waste sorting area.
During the review of these design changes and modifications, the inspectors verified the following activities:
design changes and modifications were controlled in accordance with administrative and quality control procedures; 10 CFR 50.59 safety evaluations were completed; test criteria ~were established to evaluate the performance of the modifications; test procedure results were reviewed and evaluated to verify that the acceptance criteria had been met; revisions to operating procedures were completed; training programs were revised to reflect the changes to plant systems; as-built drawings were updated and control room drawings revised (prior to placing the change in operation) to reflect the changes to plant systems; and inservice testing and inspection requirements were reviewed for the modified equipment.
The inspectors reviewed the licensee's program for temporary modifications (mechanical alterations [ blind flanges, hoses, dutchmen],
lifted leads, and jumpers) to verify that administrative controls required Senior Reactor Operator (SRO) approval for installation and removal of the temporary alterations, that formal records and in-field identification tags were maintained, that independent verification requirements were met, and that periodic audits of all temporary alterations were performed to verify the continued presence of and continued need for the alterations.
The reviewed modifications were determined to have been installed in accordance with Technical Specifications, 10 CFR 50.59, and the licensee's quality assurance program.
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l No violations or deviations were identified.
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Training (41400 and 41701)
- The effectiv'eness of training-programs for licensed and nonlicensed
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personnel was reviewed by the inspectors during witnessing'of the-licensee's. performance of routine surveillance, maintenance, and-o'perational activities and during review of-the licensee's-response to events which occurred:during October 1987.. Personnel appeared to be knowledgeable of the' tasks being performed, and nothing.was observed which indicated ineffective 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 or reviewed to ascertain that they were conducted-in accordance with approved procedures and in conformance with Technical _ Specifications.
Calibration of feedwater pressure gauge 1PI-FW059.
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Analogue channel operational test of source range nuclear instrument
1NR031.
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The following items were considered during this review:
the limiting l
conditions for operation were met while affected components or systems j
were removed from and restored to service; approvals were obtained prior
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'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;
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and any deficiencies identified during the. testing ^were properly documented, reviewed, and resolved by appropriate management personnel.
During performance of-the source range nuclear instrument test, the
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inspector identified two concerns.
The reactor operator performing the test was directed to perform only sections 1, 2, 4, 5, and 10 of the procedure by his supervisors.
Performance of portions of surveillance procedures is acceptable, if allowed by the procedure and if all of the steps in a section are performed.
The inspector observed satisfactory performance of sections 1, 2, 4, 5, and 10; however, when the exiting section (10) was completed, the temporary power supply was still installed.
The operator stopped and received guidance to perform section
7.(which contains instructions for removing the tenporary power supply)
and to reparform sections 1 and 2 (entry) and section 10.
Section 7 l
specifies that a stopwatch be used to record the'two phi time for the
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boron dilution protection system (BDPS) actuation and reset.
The special equipment section of the procedures requires the use of one stopwatch.
During a review of section 7, the inspector identified a second concern.
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4-After the voltage'is increased to the' trip setpoint, the length of time j
for the BDPS signal to actuate is measured and recorded; additionally,
the length of time for the-BDPS signal to reset is also measured.
Both measurements use thefsame starting point.
The inspector expressed:
l to the operator the concern that the stopwatch to be-used did not have the capability to record two separate times.
The operator agreed with the inspector and stopped the procedure until a second stopwatch was obtained. The surveillance was.then completed satisfactorily.
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inspector discussed these concerns with licensee management, and the licensee agreed to revise the procedure.
Revision of the procedure will be followed as an open; item (454/87039-01(DRP); 455/87036-01(DRP)).
No violations 1 or deviations were identified.
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Monthly Maintenance Observation (62703)
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Station maintenance activities of the safety-related systems and
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components listed below were observed or reviewed to. ascertain that they were conducted in accordance with approved procedures, regulatory guides,
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and industry codes or standards, and in conformance with: Technical l
- Specifications.
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l Retubing of-containment chiller 1W001CA
- Cleaning of moisture trap.on air dryer 1DG01 DAD i
Repair of flow indicating switch for radiation monitor 1PR003J J
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Replacement of rocker arms on diesei generator 1DG01KB
Following completion of maintenance on the diesel generator, air dryer, and radiation monitor, the inspectors verified that these systems 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. prior 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 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 Walkdown.(71707, 71709, 71710, and 71881T The inspectors observed control room operation, reviewed applicable logs and conducted discussions with control room operators during October 1987.
During these discussions and observations, the inspectors ascertained
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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 inspectors 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, rad-waste, and turbine buildings were conducted to obr.,erve 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.
During a tour of the Unit 1 containment chiller room to observe maintenance on radiation monitor 1PR003J, the inspector noted that there was approximately one inch of service water standing on the floor, covering an area of approximately 30 square feet, and that water was still spraying out of the pump for the 1PR002J monitor skid.
The mechanic working on the flow switch was forced to stand on other equipment to avoid standing in the water while he was working on electrical equipment.
The inspector discussed this concern with licensee management and was informed that maintenance was scheduled for the next day to repair the leak.
The inspector expressed concern that the leak had not been contained or directed, but had been allowed to spread over the floor in an uncontrolled manner.
The leak was subsequently fixed and the area cleaned up.
The inspectors verified by observation and direct interviews that the physical security plan was being implemented in accordance with the station security plan.
The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.
The inspectors also witnessed portions of the radioactive waste system controls associated with rad-waste shipments and barreling.
During October 1987, the inspectors walked down the accessible portions of the hydrogen
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recombiner systems to verify operability.
During the walkdown the inspector noted that valve 00G066 (discharge valve for the "B" hydrogen recombiner) was closed.
The mechanical lineup
for the hydrogen recombiner required valve 00G066 to be open.
The Byron Safety Evaluation Report, Supplement 5, identified that the licensee committed to administrative controls to maintain the hydrogen recombiner disc.harge valves open.
The valves are required to be maintained open during normal operation to prevent backflow through a failed hydrogen recombiner.
The licensee issued Deviation Report 6-1-87-133 to document this event.
Pending the licensee's evaluation of the cause of valve 00G066 being closed, an an unresolved item will track this finding (454/87039-02(DRP); 455/87036-02(DPP)).
The observed facility operations were verified to be in accordance with the requirements established under Technical Specifications, 10 CFR, and administrative procedures.
No other violations or deviations were identified.
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Deviation Reports (92700)
The inspector selected several licensee' Deviation Reports (DVRs) to verify compliance with regulatory requirements and the licensee's quality requirements.
The inspector's review was particularly directed-toward the licensee's corrective actions.
The DVRs reviewed by the inspector were:
l DVR Issue Date Subject 06-02-87-071 7/16/87 Failure of valve 2FW009C to open.
06-02-87-068 7/13/87 2B Diesel Generator did not achieve 4000 volts in less than 10 seconds.
06-02-87-080 8/11/87 211 inverter tripped.
06-02-87-076 7/30/87 AFW flow indicators failed low (2A steam generator).
06-01-87-077 6/05/87 Blown fuses on instrument loop cards 1PB9500/C and IPB403C/U.
In each case the corrective action appears appropriate for the event.
The DVRs reviewed by the inspector were processed in accordance with the licensee's established procedures.
No violations or deviations were identified.
10.
Onsite Followup of Events at Operating Reactors (93702)
The inspector performed onsite followup activities for events which j
occurred during October 1987.
These followups included reviews of l
operating logs, procedures, Deviation Reports, and Licensee Event Reports
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(where available), and interviews with licensee personnel.
For each
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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 that licensee investigations had identified root causes of equipment malfunctions and/or personnel errors and had taken appropriate corrective actions prior to plant restart.
Details of the events and the licensee's corrective actions developed through inspector followup are provided in Paragraphs a through c below:
a.
Unit 2 - Reactor Trip due to Feedwater Pump Trip At 1:23 p.m. on October 1, 1987, with reactor power at 92%, a reactor trip on Lo-Lo level in the 20 steam generator occurred.
Contractor personnel working on the 28 main feedwater pump high pressure stop valve had inadvertently bumped the overspeed trip plunger for the feedwater pump turbine, which is located next to the stop valve.
Total ler.gth of travel for the trip device to actuate is
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approximately 1/4 inch.
Control room operators ran the turbine back
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to 50% power but were unable to stabilize level.in the 2D steam generator before the rea.ctor trip occurred.
The plant was stabilized
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in Mode 3.
All safety systems functioned normally following the trip,
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j As corrective action the licensee has disseminated information to make operations and maintenance personnel aware of the location and sensitivity of the trip mechanism.
The licensee is evaluating the l
need to install a guard ~over the trip mechanism.
The reactor was taken critical at 9:22 p.m. on the same day.
During an attempt to synchronize the generator to the grid, a turbine trip and reactor trip occurred (see Paragraph b).
No violations or deviations were identified.
b.
Unit 2 - Unusual Event Declared due to a Loss of Offsite Power Following a Reactor Trip At 4:46 a.m. on October 2, 1987, with reactor power at 13%, a turbine trip and reactor trip occurred due to Hi-Hi level in the 2C steam generator (SG).
Operators were synchronizing the turbine to the grid when the trip occurred.
The main feedwater regulating valve (FRV) for SG 2C is 2FW530.
This valve is used from 20% to 100%
power to control the water level in the 2C SG.
Below 20% power a smaller valve, main feedwater regulating bypass valve 2FW530A, is used to control SG water level.
Valves 2FW530 and 2FW530A are in parallel.
Valve 2FW530 was leaking past its seat, to the extent that the level was still rising in the SG, even though 2FW530A was fully shut.
As a consequence, operators were manually controlling level
in the 2C SG by :nanually cycling valve 2FWO39C.
2FWO39C is downstream of both 2FW530 and 2FW530A.
The water levels in the other SGs were being controlled automatically by their respective FRV bypass valves.
Following the synchronization, the water levels in all four SGs began to swell rapidly.
The levels in SGs A, B, and D were stabilized by the automatic system.
However, operators were not able to stabilize the level in the 2C SG before the Hi-Hi level trip setpoint was reached, resulting in a turbine trip and reactor trip.
Immediately following the trip all systems functioned normally.
At 5:01 a.m. orders were radioed to the equipment operator (E0) in the switchyard to open the Unit 2 main power transformer disconnect switch.
This is a normal operation which is performed after a L
reactor trip to allow the generator output breakers to be closed l
(without the generator being connected to the grid) to reestablish the ring bus.
The E0 incorrectly opened the disconnect switch which l
feeds the Unit 2 station auxiliary transformer (SAT).
The SAT is the normal power source for the engineered safety feature (ESF)
electrical busses and for the reactor coolant pumps when the reactor is shutdown.
Consequently, the ESF and reactor coolant pump busses were deenergized when the SAT disconnect switch was opened.
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Following the loss of offsite power, the emergency diesel generators for Unit 2 started and sequenced the safe shutdown loads onto the-Unit 2 ESF busses.
The control room operators reenergized selected non-safety related busses and at 5:35 a.m. began a natural circulation cooldown of the of the reactor coolant system. The cooldown was accomplished by bleeding steam to the atmosphere via the power oper-ated relief valve (PORV) for the 2D SG, in accordance with emergency procedure 2BEP ES-0.2.
Unit I was not.affected and continued to-operate at rated power. The NRC Region III office dispatched two regional inspectors and a supervisor to the site to evaluate the event.
As a conservative action the licensee decided to activate and man the Technical Support Center (TSC). At 6:10 a.m. licensee management decided to close the PORV and continue the cooldown using the Unit 2 steam generator blowdown system.
This method of cooldown is not mentioned in emergency procedure 2BEP ES-0.2.
The licensee based its decision on the fact that the PORVs would still be available, if needea, and using the SG blowdown system would recycle water to the condensate storage tank (CST) rather than release the water to atmosphere, thereby conserving the CST's water inventory.
The CST is the normal source of water for the auxiliary feedwater pumps, and was at a lowered level when this event began.
The inspector dis-cussed this decision with operations managers and agreed that the deviation from emergency procedure 2BEP ES-0.2 did not place the plant in an unsafe condition; however, the inspector noted that the documentation in the' operating department logs (shift engineer and Unit 2 reactor operator) was less than adequate in discussing the reason why the cooldown method was changed and what level of authority had approved this decision. The licensee agreed to review the level of documentation made to support this decision and to discuss with shift supervisors the need to adequately document decisions which occur during an event.
When the loss of offsite power occurred, pressurizer pressure rose
above 2400 psig. The pressurizer PORVs have a nominal setpoint of
2335 psig.
The reactor operator noted the pressure rise and manually opened PORV 2RY455A to relieve the pressure. The operators did not expect this to happen. A subsequent review by the licensee has q
indicated that portions of the control circuits for the PORVs are powered by non-class 1-E power supplies; consequently, the PORVs will
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not actuate automatically following a loss of offsite power, even if
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an overpressure condition should exist.
However, the control room
operators can operate the valves manually, as the manual actuation I
switch for each PORV has its own class 1-E power supply. The inspector discussed this information with NRR and was informed, in a telephone conversation on November 13, 1987, between the inspector and Mr. R. Jones, Chief, Reactor Systems Branch (NRR), that the design of the PORVs' control circuits is consistent with the FSAR design and the Byron Safety Evaluation Report, NUREG-0876, section 5.2.2.
The inspector discussed this with licensee management, and the J
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licensee agreed to revise the training program for licensed reactor
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operators to heighten their awareness of the operation of PORVs
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following a loss of offsite power.
The licensee is also reviewing
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b operating procedures and annunciator response procedures to determine if further guidance is needed for the operation of PORVs following a loss of offsite power event. Consequently, the inspector had no further concerns regarding this matter.
At 1:18 p.m. the control room received indication that the diesel-driven auxiliary feedwater pump (2B AF pump) was surging (the diesel's speed was erratic). The control room operators secured the 28 AF pump and took it out of service. They informed the TSC of this situation. Operators had been in the process of refilling the diesel oil day tank, which was 18% full, when the surging occurred. A maintenance crew was dispatched to the pump and determined that the fuel oil line was air bound. The maintenance crew used a vacuum pump to bleed air from the fuel oil line, and after several hours and three unsuccessful start attempts, it was able to get the AF diesel to run. A subsequent investigation by the licensee has indicated that the fill line for the day tank is directly above the suction line to the diesel. The licensee believes that the combination of the low level in the day tank and the fuel oil falling into the tank entrained air into the fuel oil, and this entrained air was then sucked into the suction line, causing the diesel to become air bound.
The licensee has reviewed the preoperational test results for the 2B AF pump and determined that the level in the day tank did not drop below 48% during the 48-hour performance run.
Consequently, the licensee was unaware of this design problem. As temporary corrective action the licensee has posted instructions in the control room to direct the operators to fill the day tank when the level drops to 50%. The inspector will follow the questions of the adequncy of the day tank design, the effect on other diesel day tanks,
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g and the licensee's permanent corrective actions to resolve the problem, i
asanunresolveditem(454/87039-03(DRP);455/87036-03(DRP)).
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By 1:12 p.m. the licensee had completed an inspection of the SAT disconnect switch and determined that no damage occurred.
The disconnect switch was closed, and the SAT was reenergized. Had the
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- 7 diesel generators failed to function properly, the disconnect switch could have been closed almost immediately, thereby allowing the SAT
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to be reenergized. However, the licensee chose a conservative d
course of action, to inspect the switchyard and the SAT for damage
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before reenergizing the SAT, as long as the DGs were running le properly and the unit was in a stable cor.dition.
F,y 1:19 p.m. the
reactor coolant pump busses were reenergized. By 1,:12 p.m. both ESF busses had been parallehd with their respective DGs, and the DGs r
were then secured. The linusual Event was terminated at 2:18 p.m.
At 5:25 p.m. the 2D reactor coolant pump was restarted, reestablishing l
. forced circulation through the reactor core The hector coolant l
system had been cooled to approximately 400'F from 15i'F.' The reactor coolant system was feated up M normal operating temperature,andthereactorwastpencriticalat6:15p.m.on October 3, 1987. The unit was synchronized to the grid at 11:47 p.m.
on the same day and began 6 return to rated power.
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Unit 2-Unusual: Eve $tDeclaredduetoReactorCoolantSystem(RCS)
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R Leakage in Excessiof Tsclnical specifications.
Atapprbately4N0p.m.onOctober5,1987,withreactorpowerat
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Q p,25%, indications of decreasing. level in the volume control. tank (VCT)
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ere rechived;in the cnntrol room,/and an auxiliary operator in the O
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T Unit 2 containment mechanical penetration area heard a loud y
sq6ealing'nhise. The' operator tielieved the noise was caused by water from the chemic'al and volume rentrol-(CV). system. leaking past N
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a valve sdat..After exiting the area, he reported this information
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-to the, control rocm.
The reactor; operators entered the abnormal operating procedure,1'or excessive RCS leakage, 280A PRI-1,: isolated
normal letdown, and placed tne excess letdown system in service. :At 5 d 0.p.m. the licensee had determined that the leak was approximately
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40 gym, and an Unusual Event was declared. The licensee's review of-y
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records after.the event indicated that the leak rate was actually 30 s
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gpm. At 5:fi0 p.m the' inlet and vent valves for the 2B seal injection ff;1ter were found notifully closed. The valves were closed, stopping the leakage. By 5:55 p.m. normal letdown had been restored.
The Upusual Event was terminated at 6:05 p.m., and the unit was returned to rated power';
i The licensee's investigation determined that no work was in-progress
in the area when the leak started. The 28 seal injection filter had been replaced on October 2,.1987.
The post-maintenance tests were
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corepleted on October 5, and at 12:50 p.m. the filter was placed in a standby condition (inlet, outlet, vent, and drain valves verified shut).. The Mcensee believes that the vent' and inlet valves opened by themselves. Byron has ha'd previous occurrences of Kerotest
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valves sudde.nly starting to leak. The licensee's search of computer data bases indicates that other licensees have had the same problem with Kerotest valves. The licensee is evaluating long-term
corrective action and is working with the vendor on a resolution to the problem.
No violatt'ons or deviations were identified.
The inspectar will review these events in a subsequent report after the LERs arc issued.
11. ManagementMeetings(30702)'
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On October 21, 1937, Hr. A. Bert Davis, NRC Region III Administrator, and
!
members of his staff met with Mr. T. J. Maiman, Commo, wealth Edison Vice f(
President, PWR OpePations, and members of his staff (denoted in Paragraph I
%
of this report). This meeting wus held to discuss NRC concerns related to
"hd the recent loss of'offsite power event and to the excessive reactor
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sl coolant system leakage event (see Paragraphs 10.b and 10ic).
12. Unresolved Items
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Unresolved items are matters about which more information is required in j
order to ascertain whether they are acceptable items, violations, or
'
deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 7 and 10.b.
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13. Open Items Open items are matters which have been discussed with the licensee, which will be' reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. An open item disclosed during the inspection is discussed in Paragraph 6.
14. Exit Interview (30703)
The inspectors met with the licensee representatives denoted in Paragraph I at the conclusion of the inspection on October 30, 1987. The inspectors summarized the purpose and scope of the inspection and the findings.
The inspectors also discussed the likely informational content of the inspection report, with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents or processes as proprietary.
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