SNRC-1238, Annual Operating Rept for 1985
| ML20141E483 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/31/1985 |
| From: | Leonard J, Steiger W LONG ISLAND LIGHTING CO. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| SNRC-1238, NUDOCS 8604220023 | |
| Download: ML20141E483 (19) | |
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LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION ANNUAL OPERATING REPORT DECEMBER 7, 1984 THROUGH DECEMBER 31, 1985 0[
Mk'd Approved // ' '
Date W.
E.
Steiger Plant Manager 8604220023 851231 DR ADOCK 0500 2
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1 TABLE OF CONTENTS 1.0 Summary of Operating Experience 2.0 Summary of Operating Experience Relating to Safety-Related Non-Corrective Maintenance (Preventive Maintenance) 3.0 Summary of Events Conducted During Outage 3.1 Reportable Occurrences 3.2 Corrective Maintenance Pertaining to the Outage 3.3 ECC Systems Outages 4.0 Personnel and Man-Rem by Work and Job Function 5.0 Fuel Performance 2
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1.0 Summary of Operating Experience Low Power (.001%) Operating License NPF-19 was issued to the Shoreham Nuclear Power Station by the U.
S.
Nuclear Regula-tory Commission on December 7,
1984.
Following receipt of this license, the operational sources were loaded into the reactor vessel on December 17.
Fuel loading commenced on December 21, and was completed on January 19, 1985.
Initial reactor criticality occurred at 6:25 P.M.
on February 15, which was followed by two days of core physics testing.
The core reactivity was found to be within.06% of its design value, well within the acceptance criteria.
All testing that could be performed under the license limi-tation of low power testing at atmospheric pressure was completed on February 17.
For the next four and a half months Shoreham was maintained in cold shutdown pending receipt of an operating license that would authorize Phase III and,IV activities as described in several licensing submittals before the Low Power Atomic Safety and Licensing Board.
During this outage a vessel leak test was performed and the Control Rod Drive mechanism cooling water orifices were cleaned and replaced to correct a low flow problem caused by spalling of scram inlet valve seats which were made of Teflon.
Following the issuance of the Partial Initial Decision on emergency diesel generators LBP-85-18, operating license NPF-36 was issued on July 3,
- 1985, enabling low power testing at power levels not to exceed 5% of rated power.
Plant heatup commenced on July 7 and continued for three months.
On July 14 the first reactor scram occurred (one of only four to occur during the year).
Rated reactor pressure was reached on August 7,
followed by a gradual heatup.
Testing was performed at intervals during this and subse-quent heatups to ensure that all systems were operating correctly.
The events comprising initial criticality and low power testing are described in the Startup Test Report (Reference SNRC-1216).
Repeated problems were experienced with the reactor level instrumentation, the most serious of which resulted in an Unusual Event being declared on September 8 and the fourth reactor scram on September 12.
The root cause was traced to inadequate slope for portions of the reference legs of both loops.
Interim corrective action was taken and a permanent modification is now complete.
On September 27 an alert was declared, in accordance with Emergency Planning Procedure EPIP l-0, when the eye of Hurricane Gloria approached to within 10 miles of the plant.
The reactor had been shut down in anticipation of this occurrence.
No damage occurred to the plant.
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On October 6 the main turbine generator was rolled to rated I
speed, but was not synchronized to the grid.
On October 8 1
essentially all testing possible under the existing low power testing license had been completed.
The reactor was shut down on that date, and remained shut down through the end of the year.
Major activities during this outage in-cluded replacement of the neutron sources and implementation of the remaining Environmental Qualification Modifications (Reference SNRC-1235).
At the years end the reactor was in Cold Shutdown, with the t
vessel head removed and the reactor cavity fully flooded while plant modifications were being performed.
t 2.0 Summary of Operating Experience Relating to Safety-Related I
Non-Corrective Maintenance N
During 1985, Safety Related Maintenance performed during non-outage periods consisted primarily of Preventive Maintenance Program implementation and minor corrective Maintenance tasks in support of the Power Ascension Test Program.
Preventive Maintenance has been classified into two major categories, Mechanical / Electrical, and Instrument and Controls (I&C).
1 In the category of Mechanical / Electrical, 924 preventive I
tasks were completed and in the I&C category, 413 tasks were completed.
Additionally, LILCO continues to perform surveillances and preventive maintenance activities as suggested by the engine manufacturer on the Transamerica
- DeLaval, Incorporated emergency diesel generators.
The specific preventive maintenance activity which began in the November - December 1985 time period fulfills the eighteen month surveillance requirement contained in Section 4.8.1.1.2, paragraph e,
item 1, and is expected to be complete by mid-April 1986.
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3.0 Summary of Events Conducted During Outage Between February 17 and July 7, 1985, Shoreham was shut down pending issuance of the low power testing license NPF-36.
During this outage a vessel leak test was performed and the control rod drive mechanism cooling water orifices were cleaned and replaced.
There were no reportable occurrences pertaining to this outage.
On October 8,
1985 Shoreham's heatup. testing under the existing low power testing license NPF-36 was completed except for the synchronization of the main turbine generator to the grid.
The reactor was placed in Cold Shutdown to replace the neutron sources and to perform several plant modifications.
LILCO installed various pieces of equipment necessary to satisfy the - Commission regulation concerning environmental qualification of electrical equipment, 10 CFR 50.49; increased the number. of fire detectors and relocated others to enhance the fire protection program; made necessary changes to the post-accident sampling facility; and performed other minor modifications.
3.1 Reportable Occurrences Licensee Event Reports (LER)85-006, 85-017,85-020, 85-038,85-042 and 85-043 prompted an engineering evaluation for modifications to the "A"
and "B"
water level instrumentation reference legs.
Materials were ordered in December for these modifications and work is now complete.
LER 85-055 reports on the completion of replacement of electrical equipment that was not environmentally qualified per 10CFR50.49 and license condition 2.C.8 of NPF-36.
3.2 Corrective Maintenance Pertaining to the Outage 3.2.1 Corrective Maintenance for Environmental Qualification Corrective Station Modifications for the
(
environmental qualification of electrical l
equipment that were completed in 1985 are as
[
follows:
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I Design Output Station Return Package No. (DOP)
Mod. No.
Title to Service 8 3 -052 85-048 Addition of Circuit Breakers 10-10-85 8 3 -074 85-084 Restore Beck Actuator 11-23-85 Wiring 8 3 -084 85-014 Replace (2) Barton Level 11-29-85 Transmitters with Rosemount Transmitters 8 3 -085 85-029 Replace Bailey Pressure 9-05-85 Transmitters 8 3 -094 85-049 Addition of 125 VDC 8-29-85 Series Breaker to Recirculation Pump Trip Switchgear 8 3 -095 85-045 High Radiation Area Monitor 12-27-85 Cable Assembly Replacement 8 3 -096 85-077 Low Pressure Coolant 12-18-85 Injection Motor-Generated Sets Feeder Brea.kers Trip Control Circuits 8 3 -113 85-085 EQ Modification to 11-30-85 Charcoal Filter Train 8 3 -126 85-090 Change-Out Components on 12-26-85 Hydrogen Recombiner 8 4 -002 85-095 Replace Raymond MASR Series 12-20-85 Damper Actuators (9)
(with SURE series) 8 4 -007 85-079 Replacement of Flow 12-12-85 Transmitter in the Reactor Building Standby Ventilation System 8 4 -008 85-052 Install a Breaker to 11-06-85 Isolate Radiation Monitor Panel in Turbine Building 8 4 -009 85-088 Replace High Pressure 11-30-85 Coolant Injection Control Panel to Reduce Radiation Dose 8 4 -031 85-054 Replace Low Range 12-24-85 Accident Radiation Monitoring Panel 6
Design Output Station Return Package No. (DOP)
Mod. No.
Title to Service 84-098 85-055 Replacement of Flexible 11-26-85 Conduit for (93) GE Instrument Panels in Reactor Building 85-006 85-070 Replacement of SMB-3 10-30-85 Actuators with SB 3 Type 85-017 85-092 Replacement of Ametek 12-30-85 Flow Element Transmitter 85-038 85-099 Component Change Out 480V 11-29-85 AC MCC (Affects 4 MCC Units)85-039 85-076 HVAC Instrumentation 11-30-85 85-046 85-064 Replacement of Magnetr,ol 11-25-85 Level Switches (6)85-062 85-050 Fuse Isolation for 11-15-85 Unqualified Position Switch on HPCI Turbine Skid 85-246 85-121 6plicing Instrument 11-29-85 Cables to Eliminate Terminal Blocks for EQ 85-263 85-122 Three (3) ASCO Pressure 11-30-85 Switches to be Eliminated because of a lack of suitable Dynamic Qualification Data EEAR 85-047***
Replacement of ASCO 11-29-85 Pressure Switches (last of 15) l LDR 85-118 Replacement of GE 11-29-85 Radiation Detector Waterproof Kits.
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l 3.2.2 Corrective Maintenance for NRC Unresolved Item 84-46-05 1
l Corrective maintenance in response to NRC Inspection Report 84-46
-(Unresolved Item 84-46-05, non-conformance to NFPA 72 D/E) for the installation of additional fire detectors and the relocation of others were evaluated under Design Output Package 85-016 and per-formed under Station Modification 85-012.
Work conducted under this Station Modifica-tion was performed throughout the year and completed in December 1985.
Six (6) Function B type heat detectors were relocated in the HPCI skid area.
A total of 305 fire detec-tors have been installed.
A break down of type and location are as follows:
Location Fire Detector Type Additional Installation Control Building Function B Heat 18 I
Function A Smoke 57 Function A Flame 9
Rx. Bld. Secondary Function A Heat 29 Containment Function A Smoke 190 Screenwell Pump Function A Smoke 2
House l
A license amendment has been requested via SNRC-1220 to reflect the revised detector locations.
3.2.3 Corrective Maintenance to the Post Accident Sampling Facility Corrective maintenance to the Post Accident Sampling Facility were made to provide a more representative coolant gas sample, prevent nitrogen overpressurization, provide a back-up water and air supply, replace the oxygen analyzer sensor with a new high pressure sen-sor, replace regulating air metering valves with fine control metering valves, and to modify the PASS Control Panel mimic diagram to depict the as-built condition.
The Station Modifications for the Post Accident Sampling Facility are as follows:
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Dasign Output Station Return Pcckage No. (DOP)
Mod. No.
Title to Service 85-043 85-071 Tubing Modification 10-11-85 85-072 85-072 Nitrogen Overpressuriza-9-12-85 tion Prevention 85-134 85-086 Addition of an Ultimate Condensate Water Supply 10-31-85 85-149 85-083 Addition of an Ultimate Compressed Air Source 10-17-85 85-286 86-005 PASS Oxygen Analyzer &
Subsidiary Modification
- Work is currently being performed.
Return to Service will be within the first quarter of 1986.
3.3 ECC Systems Outage From December 7, 1984 to December 31, 1985 there were 65 limiting conditions for operations (LCO's) for the ECC Systems.
32 of the 65 LCO's were actual require-ments per Technical Specifications.
The remaining 29 were for tracking.
At Shoreham, an LCO is categorized as a " tracking LCO" if the technical specification LCO occurred while the reactor was in an operational mode other than that for which the LCO is applicable.
It is for this reason that the HPCI inoperability due to the turbine exhaust check valve problem is not included in the table describing the thirty-two (32)
LCOs.
Tracking LCOs are none-the-less important because they provide assurance that operability will be restored prior to entering the applicable ope, rational mode.
Of the 32 actual LCO's the breakdown is as follows.
16 are for Ell, Residual Heat Removal (RHR) 6 are for E21, Core Spray (CS) 9 are for E41, High Pressure Coolant Injection (HPCI) 1 is for all the ECC systems l
These 32 actual LCO's are detailed in Table 3.3-1 indicating for each component by system the outage dates, duration, and the reason for the outage.
24 of the 32 actual LCO's placed the system inoperable due to surveillance testing.
8 were due to equipment failure.
The corrective action for these are described below.
LCO 85-325 Problem description:
Panel lH21-PNL-103B power supply inverter IB21-INV-625D output failed low-no power light on.
Corrective Action:
Replaced with new inverter.
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LCO 85-350 Problem description:
Thermal blew as valve lE41*320MOV-041 went shut.
Valve was found fully shut when thermal was reset.
Corrective action:
Found valve operator torqueing too hard into seat.
Torque currents were 70-118 amps.
Lowered torque switch to 1.0 from 1.5 and torque currents dropped to 15.8 - 24.9 amps.
LCO 85-442 Problem description:
With flow on RHR "A"
Loop, valve
. Ell *MOV-42A did not open when closed down.
Corrective action:
Replaced two broken electrical terminal connection lugs for this valve.
LCO 85-453 Problem description:
Water level indicators for the "B" reference leg read high and diverged from the water indicators for the "A"
reference leg due to excess condensate accumulation in the steam line leading to the condensing chamber.
Corrective action:
Added additional insulation and installed a hanger on the line to assure that line slope remains constant.
LCO 85-476 Problem description:
Locking screw broke on flow switch making the switch repeatedly inaccurate.
Corrective action:
Replaced flow switch with new one.
LCO 85-482 and 85-483 Problem description:
Leak discovered at pipe weld test connection near MOV-40A.
Corrective action:
Repaiaed leak in weld pipe to coupling connection.
LCO 85-487 Problem description:
HPCI suction relief valve lE41*320RY-145 leaks.
Found disc and seat pitted.
Corrective action:
Repaired valve:
machined and lapped seat and disc.
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TABLE 3.3-1 DATE OUP DATE RE'IURNED
' DURATION
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SYSTDI C&IPONENP ICO #
OF SERVICE
REASON OUP Ell Icop "B" A85-040 1/30/85 1/31/85 1
8 Removed loop frcm service -
for ILRT Ell IOV 47/48 A85-048 2/4/85 2/14/85 9
12 84.002.03 inplcsnentation of this surveillance procedure E11 MOV 47/48 A85-088 3/4/85 3/4/85 5
I&C lii/ Press isolation surveillance.
Ell Snubber A85-126 3/24/85 5/22/85 59 10 Snubber was removed to allow for CRD maintenance Ell RHR sys A A85-127 3/24/85 5/22/85 59 12 Snubber removed RIE A injection line i
Ell RHR Shutdown cooling A85-229 5/28/85 5/30/85 2
1 Repack MOV-047 Ell B LPCI Iow A85-325 7/4/85 7/4/85 14 B LPCI low pressure Permissive permissive affected by EOCS l
analog trip cabinets l
Ell Shutdown cooling A85-353 7/20/85 7/22/85 Perform TP 41.012.03 Ell Shutdown Cooling A85-368 7/27/85-7/27/85 9
Shutdown cooling isolation surveillance E11 MOV 042B A85-417 8/22/85 8/22/85 1
11 Routine Operation of RHR for Suppression Pool cooling antirotational device Ell Shutdown cooling A85-424 8/25/85 8/26/85 1
10 Surveillance disables Shutdown cooling mode Ell IOV-042A A85-442 9/4/85 9/4/85 While running A loop in Suppression Pool cooling mode Ell Shutdown Cooling A85-465 9/13/85 9/13/85 13 In ;5eparation of "B" Ref leg hydro 11 l
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TABIE 3.3-1 (Con't) i DATE OUP DATE RETURNED DURATION I
SYSTD4 COMPONErff 100 #
OF SERVICE
REASON OUP Ell Low Pressure Coolant A85-482 9/21/85 9/22/85 1
4 Icak at pipe weld in line Injection "A" loop ntmber WR 230 (WR 230 pipe)
)
E11 Suppression Pool A85-483 9/21/85 9/22/85 1
4 Isak at pipe weld in line l
spray ntsnber WR 230 i
E11 RHR Shutdown Cooling A85-560 11/13/85 11/18/85 5
Removed fran service for j
I&C surveillance preventive maintenance i
E21 Core Spray System A85-075 2/24/85 2/28/85 4
15 Operations moved check j
"B" Loop valve nust recheck local leak rate
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E21 W Vs 33A/B A85-144 4/4/85 4/4/85 2
Core spray injection valve j
surveillance l
E21 MOV 33A/B A85-200 5/11/85 5/11/85 4
Core spray injection valve open permissive i
surveillance
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E21 Core Spray injection A85-401 8/14/85 8/14/85
'Ib perform SP 44.203.09 valve s
j E21 Core spray injection A85-459 9/11/85 9/11/85 7
Surveillance j
valve "A" Icop E21 Core Spray injection A85-462 9/11/85 9/11/85 4
Surveillance valve "B" Icop E41 HPCI system A85-329 7/9/85 7/9/85 11 Reactor pressure 150 psig E41 HPCI system A85-333 7/12/85 7/12/85 Performance of STP-15 i
de-energization of r
system motor operated valves and turbine quick start i
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TABLE 3.3-1 (Con't) i.
DUPATION SYS'Ind OOMPONENI' IfD #
OF SERVIG
REASON OUT' E41 MJV-041 steam inlet A85-350 7/18/85 7/20/85 2
18 Upon receipt of isolation signal valve themals blew E41 Turbine stop valve A85-378 8/3/85 8/10/85 7
7 SP 44.202.01 being performed E41 HPCI Pmp A85-412 8/18/85 8/23/85 4
8 STP-15 E41 MOV-41 A85-464 9/12/85 9/16/85 3
8 MOV closed on autmatic isolation l
l E41 FS-003 A85-476 9/19/85 9/20/85 1
5 Technician broke off l
locking screw while j
performing test E41 HPCI system A85-487 9/23/85 9/24/85 Outage for repair of leaking relief valve E41 HPCI system A85-488 9/24/85 9/28/85 4
2 R m oved frm service for hot restart & hot alignment E11/E21/
ADS, core spray A85-453 9/8/85 9/8/85
.4 "B" RPV level reference leg E41 leak l
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4.0 Personnel and Man-Rem by Work and Function j
4.1 The following table entitled "SNPS Annual Dose (20.407)
Report" was prepared to provide the information required by Technical Specification Section 6.9.1.4 and 10 CFR 20.407 (a) (2).
The total number of individuals for whom personnel monitoring was provided exceeds the total number of individuals for whom personnel moni-toring is required under 10 CFR 20.202(a).
This infor-1 mation is included in this annual operating report and will be transmitted to the Director, Nuclear Regulatory Research under another transmittal letter.
SNPS ANNUAL DOSE (20.407) REPORT START DATE - 01/01/85 END DATE
- 12/31/85 WHOLE BODY EXPOSURE NUMBER INDIVIDUALS RANGES IN RANGE NO MEASURABLE EXPOSURE (Rem) 1606 l
MEASURABLE EXPOSURE 40.1 577 0.1 TO 0.25 31 0.25 TO 0.5 4
0.5 TO 0.75 3
0.75 TO 1.00 1
1.00 TO 2.00 0
2.00 TO 3.00-0 3.00 TO 4.00 0
i 4.00 TO 5.00 0
5.00 TO 6.00 0
6.00 TO 7.00 0
4 7.00 TO 8.00 0
8.00 TO 9.00 0
9.00 TO 10.00 0
l 10.00 TO 11.00 0
11.00 TO 12.00 0
12.00+
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4.2 The following table, entitled " Personnel and Man-Rem by Work and Job Function", was prepared to comply with the requirements of Section 6.9.1.5 of the SNPS Technical Specifications.
It was prepared using the guidance contained in Regulatory Guide 1.16, Appendix A,
and presents data which goes beyond our Technical Specifi-cation requirements.
Minor differences are due to rounding off values.
The column entitled
' Station Employees' applies to all LILCO employees whose normal work location is the Shoreham Nuclear Power Station.
The column entitled
' Utility Employees' applies to 1
LILCO employees whose normal work location is not at the Shoreham Nuclear Power Station.
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i PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION Number of Personnel
(>- 100 mrem)
Total Man-Rem (2)
Contract
' Work and (1)
Station Utility Contract Worker Station Utility Worker Job Function Employee Employees and Others Employees Employees and Others Reactor Operations and Sur/eillance Maintenance Personnel i
l a) Maintenance 3
None 3
1.568 Insignificant 3.214 b) Instrument & Control None 1
0.525 Insignificant 0.448 0.030 Insignificant c) Computer Engineering None Operations Personnel 1.109 Insignificant 0.076 a) Operations 1
None b) Reactor Engineering None 10 0.212 Insignificant 3.545 c) Systems Engineering None 0.144 Insignificant 0.068 Radiological Controls a) Health Physics 11 None 8
3.73 Insignificant 2.04 b) Radiochemistry None 0.06 Insignificant 0.13 c) Radwaste 1
None 0.17 Insignificant Supervisory / Management None 1
0.16 Insignificant 0.410 Engineering (Not Operations)
I 0.070 Insignificant a) Rad. Protection None 0.140 Insignificant 0.010 b) Nuclear Systems None 0.140 Insignificant 0.290 c) Project Engineering None Outage / Modifications Insignificant i
a) Outage Planning None 0.010 Insignificant b) Planning & Scheduling None 0.050 Insignificant 0.10 c) Modi fications None l
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. PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION
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Number of Personnel
(> 100 mrem)
Total Man-Rem (2)
Contract III Work and Station Utility Contract Worker Station Utility Worker Job Function Employee Employees and Others Employees Employees and Others Nuclear Operations Support Department 0.116 Insignificant 0.012 a) QA/QC None 0.010 Insignificant 0.430 b) Security None 0.07 Insignificant 0.110 c) Training None t
Function Total Maintenance 3
None 4
2.12 Insignificant 3.66 Operations 1
None 10 1.47 Insignificant 3.69 l
l Radiological Controls 12 None 8
3.96 Insignificant 2.17 l
(includes H.P.)
i Supervisory / Management None 1
0.16 Insignificant 0.41 Il Engineering None 0.35 Insignificant 0.30 (not Operations)
Outage / Modifications None 0.06 Insignificant 0.10 0.19 Insignificant 0.56 l Nuclear Operations None Support 1 GRAND TOTAL 16 23 8.32 10.89 l
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Doses associated with low power testing (2)
Represents 100% of cumulative man-rem exposures l
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5.0 Fuel Performance There were no indications of fuel failure at Shoreham for 1985.
No special irradiated fuel examinations were conducted.
Such testing was unwarranted due to the low operational power level.
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A 9%emamuung t:
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LONG ISLAND LIGHTING COMPANY axwme manauw rd SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 JOHN D. LEONARD.JR.
VICE PRES 10ENT NUCLEAR OPER ATIONS MAR 0 31986 SNaC-1238 Dr. Thomas E. Murley Office of Inspection and Enforcement Region 1 U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 Annual Operating Report Shoreham Nuclear Power Station Docket No. 50-322
Dear Dr. Murley:
Pursuant to the requirements of Shoreham Technical Specification Sections 6.9.1.4 and 6.9.1.5, attached is a copy of the Annual Operating Report.
If there are any questions, please contact this office.
Very truly yours, f
.m ht-
$b?Vi&jl
-J.k D. eonard, Jr, Vice President-Nubic ar Operations t
)
V NEGF:ck Attachment cc:
R.
Caruso J. A.
Berry
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