IR 05000400/1988038
| ML18005A721 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/23/1988 |
| From: | Bradford W, Fredrickson P, Shannon M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18005A720 | List: |
| References | |
| 50-400-88-38, NUDOCS 8812060271 | |
| Download: ML18005A721 (11) | |
Text
REPORT DETAILS Persons Contacted Licensee'mpl oyees- -. ---: "
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Matson,"'Vice Pr'esident,'arr'is Nuclear"Prb'ject" "
Hinnant, Plant General Manager'ibson, Director, Programs and Procedures Tibbitts, 'Director; Regulatory Compliance-.
-"
Bohanan, Director, Special Programs Van Metre, Manager.",~Technical"Support":".."~" "" '.
Morton, Manager, Maintenance Collins, Manager, Operations Sipp, Manager, Environmental and Radiation Monitoring Braund, Supervisor, Security Lentz, Systems Engineering Milson, Reactor/Performance Engineering Moods, Testing and Maintenance Support Batts, Supervisor,. Mechanical Maintenance Smith Supervisor, Operations Support Olexik, Supervisor, Shift Operations Forehand, l3i'rector, gA/gC Millett, Manager, Outages and Modifications Other licensee employees contacted included technicians operators, mechanics, security force members, engineering personnel and office personnel.
'Attended exit interview.
Acronyms and initialisms used throughout this report are listed in paragraph 12..
Operational Safety Verification t71707)
The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the report period.
Also, the operability of selected emergency systems was ver>fied, tagout records were reviewed and proper return to service of affected components was verified.
Tours of the plant were conducted to observe plant equipment conditions; including fluid leaks, excessive vibrations, and general housekeeping efforts.
The inspectors verified compliance with selected LCO and results of selected surveillance tests.
The verifications were accomplished by direct observation of monitoring instrumentation, valve positions, switch positions, accessible pipe snubbers, and review of completed logs, records, and chemistry results.
The licensee's compliance with LCO action statements was reviewed as events occurred.
8812060271 SS1123 PDR ADOCK 05000400
The inspectors routinely attended meetings with certain licensee manage-ment and observed various shift turnovers.
These meetings and shift turnovers provided a daily status of plant operations, maintenance,-
and testing activities'n progress, as well as discussions of signi.ficant problems.
The inspectors =reviewed-the shift foreman-'s log,. control room operator's log, clearance
.center t'ag. out.=logs, system'tatus logs, chemistry
~ and health physics.logs,. and~control: status -board:-::The -
inspectors noted that the operators appeared to be alert and aware of.
changing plant conditions.
The following events and conditions were reviewed:
'.
Pressurizer Pow'er Operated'Relief Valve - Sp'urious Opening On October 20, 1988, at 8:03 a.m., pressurizer pressure transmitters 444 and 445 indicated erratic low and high readings.
The high readings actuated power operated relief vaTve 444B and approximately 28 gal]ons of reactor coolant were discharged into the pressurizer relief tank.
The valve reclosed following the instrument perturbations and no further actuations were experienced.
The initial root cause was attributed to a radio transmission in the area of the pressure transmitter inside of containment.
Discussions with personnel indicated that no previous problems had been experienced in this area with radio transmissions, It was also noted, that the security radio bands experienced chatter at the same time as the erratic instrument behavior.
The licensee declared an unusual event and made the appropriate notifications in a timely manner.
The licensee is continuing the'nvestigation of this event and the findings will be reviewed by the inspectors.
b.
Loss of "A" Train Electrical Power On September 21, 1988, at j.:00 p.m.
with the unit in mode 5,
a complete loss of power occurred on the
"A" train electrical distribution system.
A dispatcher supervised relay crew was performing switchyard testing and inadvertently lifted the wrong lead in the fault circuitry which caused offsite supply breakers to trip and lock out.
The
"A" diesel generator was inoperable due to maintenance and the
"A" RHR pump, which was running at the time, tripped.
The
"B" RHR pump was started at 1:07 p.m.
and. shutdown cooling flow was established at 1: 12 p. m.
The recovery of shutdown cooling and the "A" train electrical distribution was witnesses by the inspector and appeared to be excellent.
The licensee was questioned about the operability of
"A" RHR pump without its emergency power source being available, in that the
"A" diesel generator was inoperable.
The inspectors performed a detailed review of the shutdown RHR and electrical distribution Technical Specifications and reviewed the licensee's previous Technical Specification changes.
The inspectors have concluded, as the
C.
d.'icensee had previously done, that the plant was in full compliance with the applicable Technical Specifications prior to the loss of electrical power event.
Reactor Coolant System'. Unidenti:fied Leakage.
On October 6;:..1988;,at 2 30.p'.m':: with. the-,unit...in-mode.4;,
unidenti=
- ':-fied 'rea'ctor ':cool'ant:system:
l'eakage was-'"..found 'to"-'be '2:4'GPM;
-The
~
licensee was unable-to locate the"source. of the RCS leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as per Technical Specifications.
Accordingly, as required by the site emergency plan, an unusual event.was, declared...- Letdown flow was isolated'nd the unidentified leakage was reduced to.39 GPM; thereby allowing.terms,nation,:os,the-unusual event:.
Plant.-personnel found the
"A" mixed bed demi'ne&alizer drain valve, FCS-74, cracked open 1 1/2 turns.
This valve had been previously checked closed using the remote handwheel.
Letdown was re-established and the RCS leak rate for unidentified leakage remained within the Technical Specification limits of 1 GPM.
'ain feedwater valve 1FM-307 is a
3 inch air operated valve and serves as the isolation valve for the bypass line around the main feedwater isolation.valve.
This valve is also a
containment isolation valve and is closed by'
main feedwater isolation signal.
It is opened between 19'nd 20K power in order to warm up the main feedwater line prior to opening the main feedwater isolation valve, and is required to be open for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to prevent thermal shocking of the steam generator inter~als.
The valve was found to have an excessive stroke time which exceeded the surveillance test requirement of 10 seconds.
The failure of the valve to meet the ASME Section NI stroking requirements made the valve inoperable and consequently was required to be closed per TS 3.6.3.
The licensee determined that the stroking problem was due to a deficiency with the valve internals, which could only be repaired with the plant in mode 5.
Unable to perform the necessary feedwater warming and remain in compliance with the TS 3.6.3 Action Statement, the licensee formally requested the NRC to grant enforcement discretion until 1FM-307 could be repaired.
Enforcement discretion was granted on October 14, 1988.
No violations or deviations were identified.
3.
Monthly Surveillance Observation (71709)
The inspectors witnessed the licensee conducting maintenance surveillance test activities on safety-related systems and components to verify that the licensee performed the activities in accordance with licensee requirements.
These observations included witnessing selected portions of each surveillance, review of the surveillance procedure to ensure that
administrative controls were in force, determining that approval was obtained prior to conducting the surveillance test, and verifying that the individuals conducting the test were qualified in accordance with plant approved.procedures.
Other observations included ascer'taining that test instrumentation used was. calibrated, data collected was within, the specified
'requirements'of -:Technical Specifications,-. any-identified-=
discrepancies'were,properly.'noted;,
and the systems were:correctly,.returned to service.
Portions"of the fo11owing-test'ctivities were observed or reviewed by the inspectors:
OST-1015 Emergency. Service Mater. Operability "-
OST-1026
.
Reactor Coolant SystemLeakage:.".'.-"~
OST-1080 Auxiliary Feedwater Full Flow OST-1081 Containment Visual Inspection OST-1106 Chemical Volume Control System/Safety Injection System Operability OST-1108 Residual Heat Removal System-Oper ability OST-1216 Component Cooling Mater quarterly Test.(Vibration Test Only)
OST-1315 Emergency Service Mater Valves In Service Inspection OST-1803 Containment Sump Visual Inspection No violations or deviations were identified.
Monthly Maintenance Observation (62703)
Station maintenance activities of safety-related systems and components were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and were in conformance with Technical Specifications.
Items considered during the review included:
verification that limiting conditions for operations were met while components or systems were removed from service; approvals were obtained prior to initiating the work; approved procedures were used; completed work was 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 were properly certified; and radiological and fire prevention controls were implemented.
Mork requests were also reviewed to determine the status of outstanding jobs to assure that priority was assigned to safety-related equipment maintenance which may affect system performanc Portions of the following activities were observed or reviewed:
1 HS-72; The main steam supply to the turbine driven AFM pump was found to have excessive 'leakage; The plant re-,entered mode 5.and the
,valve was rebuilt.
- '
Containment 'AirCooler; The 'containment:-ai'r ':cooler:hadrexperienced leak'age o'n -the-.tube-manifold A'r'azing"proce'ss'as'-used'to make'he necessary'epairs.
"'alve Live Load -Packing; The plant completed.approximately 700, live load backings during this outage.
Initial results appear to be excel Ient.":-: ':: '.:-.
Diesel Generators; Problems were experienced with the Moodward governor.
The manufacturer's technical representative was on site to assist in repairs.
During an inspection of the main control board internals, various discrepancies were identified such as loose debris missing cover plates and loose amphenol connectors.
Operations personne)
took immediate action in writing deficiencies for the identified items.
The loose debris was removed immediately and cover plates were replaced, Maintenance personnel secured loose spare amphenols and.verified proper locking for other main control board amphenols.
No violations or deviations were identified.
5.
Radiological Protection Progr am (71709)
Selected activities of the licensee's Radiological Protection Program were reviewed by the inspectors to verify conformance with plant procedures and NRC regulatory requirements.
The areas reviewed included:
organization and management of the plant's health physics staff;
"ALARA" implementation; Radiation Work Permits (RWPs) for compliance to plant procedures; personnel exposure records; observation of work and personnel in radiation areas to verify compliance to radiation protection procedures; and control of radioactive materials.
No violations or deviations were identified.
6.
Physical Security Program (71881)
Licensee's compliance to the approved security plan was reviewed by the inspectors.
The inspectors verified by observation and interviews with security force members that measures taken to assure the physical protection of the facility met current requirements.
Areas inspected included:
organization of the security force; establishment and maintenance of gates, doors, and isolation zones; access control; and badging procedure No violations or deviations were identified.
Licensee Event Reports (92700)
The following Licensee Event-..Reports (LE%) were:.reviewed fov:potential generic problems to determine'rends, to determine whether information
. included in -.'-the. report"meets 'he-NRC reporting'equirements"-and.-to*
consider whether the covrective action. discussed in the report -appears appropriate.
The licensee action was reviewed to verify that the event has been reviewed and evaluated by the licensee as required by the Technical Specifications; that corrective action was taken by the licensee; and that safety limits, limitiqg safety settings and LCOs were not exceeded.'he inspectot"'examined the incident report; lo'gs and records, and interviewed selected personnel.
The following reports are considered closed:
LER 87-25 LER 87-26 LER 87-27 LER 87-29 LER 87-30 LER 87-31 LER 87-32 LER 87-33 LER 87-35 LER 87-36 LER 87-38 LER 87-39 LER 87-40 LER 87-43 LER 87-44 LER 87-46 LER 87-47 LER 87-48 LER 87-49 LER 87-50 LER 87-51 LER 87-54 Reactor Trip Auxiliary Feedwater Actuation Pressurizer Pressure Channels Inoperable Inoperable Piping Snubber on Steam Generator Blow Down Line Missed Survei lance Test Reactor Trip REM-21ML-3541 Radiation Monitor Inoperable Emergency Diesel Generator Missed Air Roll Reactor Trip Emergency Diesel Generator Missed Air Roll Reactor Trip Intermediate High Range Trip Setpoint Out of Calibration Time Interval Exceeded for Caloyimetric Containment Isolation Valves Omitted from Monthly Position Check Missed Surveillance Test for Axial Flux Difference Failed Main Feed Recirculation Flow Control Valve Auxiliary Feedwater Actuation Missed Surveillance Test for Axial Flux Difference Auxiliary Feedwater Actuation Reactor Shutdown - Unidentified Leakage Greater Than
GPM Auxiliary Feedwater Actuation Reactor Shutdown - Unanalyzed Condition No violations or deviations weve identi fied.
Reactor Trip On October 14, 1988, at 12:27 p.m., the plant lost the "B" condensate pump and
"B" condensate booster pump.
The control room operator manually tripped the
"B" main feed pump after observing "0" PSIG suction pressure.
The operators manually tripped the reactor and main turbine, started the auxiliary feedwater system, and closed the main steam isolation valves.
The reactor plant was stabilized using the steam generator power operated
re1ief valves and auxiliary feedwater.
Attempts were made to restart the
"A" and
"B" condensate pumps with no success.
After troubleshooting the contro1 circuitry and performing system inspections, the licensee found that the contro1 board. instrument"for.:hotwel.l 1evel.,-was -,in error and.-
indicating higher-than actual.
The--hotwell--level had actua1ly been 1ow and ca'used-'a loss=. of suction in-the condensate.
pump.-which=caused::a.-low==:-==
- "
- condensate"discharge'-.
pressure,'pump -trip.: - Repairs were made and theplant
. ":;returned.,to"'operate'on;
':":.'.-.
""-.-.:
Additional problems were encountered during the loss of condensate pump event:
9.
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.
The masn feed'pump=:motor.".-,'ifai'led,"to <trip on'loss os the'running:
condensate or condensate booste'r pump motors as required.
It was found that the logic trip relay had failed and was unable to trip the main feed pump motor.
V CExcessive and repeated water hammers occurred in the secondary plant condensate and feedwater piping. It appears that the condensate and feedwater systems were partially drained after the condensate pump trip and before the main feed pump was tripped.
The water hammers appeared to be generated at the 2B feedwater heater and lasted for a period of approximately 20 minutes with a frequency of about once every 8 to >0'econds.
Technical support and site engineering inspected piping supports during and after the event and there was no apparent damage to the plant.
No violations or deviations were identified.
Part 21, Report No.
145 - From IMO Delaval, Inc.
On April 29 1988, IMO Delaval, Inc.
reported to NRC and the affected licensees or a potential problem with certain en@inc control devices in the air start, lube oil, jacket water systems, and crankcase systems.
The component manufacturer's equality Assurance program did not test the product.
The components, as listed below, are identified by IMO part number and description.
KR-001-000 Air Start Valve KR-004-001 Air Start Val ve KR-004-002 Air Start Va1ve F-573-156 Low Pressure Lube Oi1 Trip F-573-330 High Temperature Jacket Water Trip F-573-359 High Pressure Crankcase Trip Another problem was identified on May 12, 1988, with Part No. F-573-156, Low Pressure Lube Oil Trip.
The diaphram can be held solid against the pressure head, creating a smaller surface area, and requiring a higher pressure to activate the valve.
Corrective action was to add an additional 1-1/8 inch diameter with.030 inch deep counterbore in the pressure hea.
12.
The licensee has initiated corrective action as follows:
The low pressure lube oil trip, high pressure jacket water trip, and high pressure crankcase.tnp devices'.were"returned to 'IMO-Delaval for remachining;:
inspection, and testing." -'-
Mork 'Request.88-AKNT1 was'-wiiitten'.to:. di.sassemble the."air start valves in stock and-.verify-oper'ability"and'leakage')ates
"" """"*'
his item is closed.
Allegation Investigation Allegation 88-A-0068 was received by the NRC and concerned Environmental gualsfication (Eg) considerations in that the plant was operated for some time without Raychem heat shrink on two triax cable connections.
Cables No.
10062R and No.
10062( are located in the containment building inside a
junction box in the seal table room.
Resolution:
After a
review of a
Harris Nuclear Plant significant operati'onal occurrence report and discussions with responsible management personnel, the inspectors determined that the cables listed above are connected to the Intermediate Range Nuclear Instrument. Detectors.
These detectors are not required for post accident conditions and, therefore, do not have to meet Eg requirements.
The Raychem heat shrink was placed on the connections by the licensee in order to prevent moisture intrusion into
.the amphenol and possibly rendering the detector inoperable.
The cables
.listed above are not required to be maintained per Eg requirements.
This item is closed.
Exit Interview The inspection scope and findings were summarized during management interviews throughout the report period and on October 20, 1988, with the plant manager and selected members of his staff.
The inspection findings were discussed in detail.
The licensee acknowledged the inspection findings and did not identify as proprietary any material reviewed by the inspectors during this inspection.
List of Acronyms and Initialisms AFM ALARA ASME CFR Auxiliary Feedwater As Low As Reasonably Achievable American Society of Mechanical Engineers Code of Federal Regulations Environmental gualsfication Feedwater
GPM LER LCO No.
NRC OST PSIG RCS REM RHR RMP TS WL on Gallons Per Minute Licensee Event Report Limiting Conditions for Operati Number Nuclear Regulatory Commission
.
=Operation 'Surveillance-Test.:-:-
" Pounds per Square-Inch
"'Gauge'-
'eactor C'oola'nt"-System=-rl-.~
'adiation Monitor'esidual Heat Removal
, Radiation Mork Permit Technical Specification
, Liquid 'Ha'ste
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