ML20212D367
| ML20212D367 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/31/1986 |
| From: | Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 8703040103 | |
| Download: ML20212D367 (42) | |
Text
-
Cm WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968
- 3000 George \Yashington \Vay
- Richland. \Yashington 99352 6,
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- =rk o
$ 9 Docket No. 50-397 75 Q o
February 18, 1987 F 9
Mr. J. B. Martin Regional Administrator Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596
Dear Mr. Martin:
5
Subject:
NUCLEAR PLANT N0. 2 1986 ANNUAL REPORT
Reference:
- 1) Title 10, Code of Federal Regulations, Part 50.59(b)
- 2) WNP-2 Technical Specifications, 6.9.1.4 and 6.9.1.5
, 3) Regulatory Guide 1.16, Reporting of Operating Information -
! Appendix A In accordance with the above listed references, the Supply System hereby submits the Annual Report for calendar year 1986. Should you have any questions or comments please contact M.R. Wuestefeld, WNP-2 Plant Engineering Supervisor, Reactor Systems.
Very truly yours, 1
w&
- .M. Powers Plant Manager CMP
- MRW:TRW i
attachments B703040103 861231 PDR R ADOCK 05000397 pg
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o ANNUAL OPERATING REPORT OF O WNP-2 FOR 1986 0
00CKET NO. 50-397 FACILITY OPERATING LICENSE NO. NPF-21 O
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Washington Public Power Supply System O 3000 George Washington way Richland, '4ashington 99352 O
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TABLE OF CONTENTS o 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . 1-1 o
A. 1986 Power History for WNP-2 Graph ....... 1-4
- 2. REPORTS ....................... 2-1 O A. Annual Personnel Exposure and Monitoring Report .............. 2-2 B. Main Steam Line Safety / Relief Valve Challenges . . . . . . . . . . . . . . . 2-3 O C. Sumary of Plant Operation ........... 2-5
- 0. Sumary of Significant Maintenance Performed on Safety Related Equipment . . . . 2 - 13 E. Indications of Failed Fuel . . . . . . . . . . . 2 - 21
'O F. Plant Changes and Tests . . . . . . . . . . . . . 2 - 22
- 1. Plant Modifications and Design Changes . . . 2 - 23
- 2. Plant Procedure Changes . . . . . . . . . . 2 - 31 0 2 - 32
- 3. Plant Tests . . . . . . . . . . . . . . . .
G. Reactor Coolant Activity Cumulative Iodine Levels . . . . . . . . . . . . . . . . 2 - 33 0 1. tlNP-2 Dose Equivalent Iodine Graph . . . . . 2 - 34 0
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- 1. INTRODUCTION The Annual Operating Report of Washington Public Power Supply System Plant O Mumber 2 (WNP-2) for 1986 is provided as a supplement to the Monthly Operation l Report. I additionalIt required should bereports noted that, for ease are also of reference included. and completeness,in This report is submitted 1 accordance with the requirements of Federal Regulations and Facility Operating l License HPF-21.
O WHP-2, a 3323 MWt 9WR-5, began commercial operation on December 13, 1984. !
Ouring March 1986, NNP-2 was shutdown after running continuously for a record 116 days. The Bonneville Power Administration (BPA) which receives all the power from the plant, requested that the Supply System remain in a unit reserve status until the start of the annual maintenance and refueling outage in April. The decision was based on two major factors: high streamflow in Q the Columbia River system and the " drying up" of the California market for Pacific Northwest surplus electricity. In June, after the outage, the plant was restarted. For the remainder of the year, the generator was removed from the grid for a total of 13 days. In November, higher-than-acceptable vibra-tions were experienced on Reactor Recirculation Pump "A". As a result, the plant ran in single loop operation mode through the balance of 1986. Refer to Page 1-4 for the 1986 Power History Graph.
O During the year, there were several examples of major accomplishments which required significant effort on the part of the Supply System to successfully complete. The following is a sumary of those efforts:
n (a) The first refueling outage was successfully completed. Significant
- activities included:
o Disassembly, inspection and repair of Reactor Recirculation Pump "B".
It was determined that the internal assembly would be replaced.
o Removal of spent fuel assemblies and refueling the reactor. The O refueling activity included 128 new fuel bundles and, more signif t-represented a transition from G.E. to Exxon fuel. The cantly, transit ion required a significant Technical Specification Amendment and implementation of a new core monitoring program.
o Disassembly and inspection of one of the low-pressure turbines.
O Although no cracks were found, three blades were replaced as a precautionary measure, o Replacement of the impellers in the three circulating water pumps that circulate water from the cooling towers through the condenser.
The original bronze impellers were replaced by a more durable stain-O less steel natorial.
O I-1 0
o Induction Heating Stress Improvement (IHSI). The remaining 35 welds designated as requiring the IHSI process, which were not performed during the previous outage, were completed during the 1986 refueling outage. As a part of this task, ultrasonic exami-o nations (for the detection of Intergrant.lar Stress Corrosion Cracking) were performed by EPRI-qualified examiners before and after IHSI. No unacceptable conditions were found.
(b) WNP-2 continued to have an excellent record for limiting worker radia-tion exposure of all boiling water reactor nuclear plants in the o country, according to a published report by the Nuclear Regulatory Commission. In 1986, total radiation exposure at the plant was 221.5 man-rem, an improvement over the 230 man-rem goal set by the Supply System and less than half the long-range goal set by the nuclear industry for boiling water reactors. The Institute of Nuclear Power Operations has set 460 man-rem as the industry goal for 1990 for BWR's.
U (c) Several conditions to WNP-2 Operating License NPF-21 were either sat-isfied or completed and included:
o Condition No. 4 - Seismic Qualification o Condition No. 8 - Fission Gas Release 3 o Condition No.10 - Thermal Hydraulic Stability lO o Condition No.12 - Alternate Remote Shutdown System l o Condition No.16 (3-b) - Emergency Response Capability l
(Regulatory Guide 1.97) o Condition No.19 - Relocation of Engine-Mounted Controls o Condition No. 20 - Diesel Air Dryers O (d) Following a two-week inspection in December, a team of INP0 inspectors evaluated overall operations at WNP-2 as " exemplary". They noted six areas of " good practice" which included a maintenance program for maintaining environmental qualification of equipment, maintenance practices regarding the valve packing program, calibration of radia-tion monitoring equipment, chemistry programs designed to protect U coolant systems, organizational and administrative measures to track comnitments made to regulatory agencies, and implementation of a well-organized " fitness for duty" program.
(e) WNP-2 continued to have a positive trend in the reduction of Licensee Event Reports (LERs). A total of 44 LERs were written during 1986 as O compared to 63 LERs in 1985 and 131 in 1984. It should be noted that, of the LERs written during 1986, nine were associated with inadvertent or spurious Chlorine Monitoring System actuations and four were asso-ciated with Reactor Water Cleanup System spurious isolations.
During the outage, the method of circulating water treatment was O changed from gaseous chlorination to chemical addition of sodium hypo-chlorite due, in part, to the industrial safety hazards associated with chlorine gas. The chlorine gas was removed from the WNP-2 site and, as a result, the WNP-2 chlorine detection system was no longer needed to ensure control room habitability in the event of a non-credible chlorine gas release. During April, an amendment to the O
1-2 0
O WNP-2 Technical Specifications was requested to delete chlorine detec-tion requirements. On January 18, 1987, the Technical Specification amendment was approved, and the detectors were isolated to prevent spurious challenges to Emergency Safety Feature Control Room ventila-O tion equipment.
Regarding Reactor Water Cleanup System isolations, a Plant Engineering study was conducted which resulted in changing instrumentation range settings to preclude system sensitivity causing the spurious isola-tions. The change has successfully reduced challenges to the
.O isolation logic.
The actual and adjusted capacity factors for the year are listed in the following table. For the periods of time when the plant was operating with one recirculation pump, the adjusted capacity factor was based on a maximum power output of 71.7% rather than 100%.
O" MONTH CAPACITY FACTOR ADJUSTED CAPACITY FACTOR January *** 68.4 95.4 February 54.9 76.6 b b O May 0 0 June ** 14.6 14.6 July 81.0 81 .0 August 90.7 90.7 September 75.8 75.8 Oct ber 88.0 88.0 O November *** 56.4 59.8 December =57.9 94.6 Overall 51 .5 58.8 O
- Started Maintenance / Refueling Outage
- Ended !!aintenance/ Refueling Outage
- Entered Single t.oop Operation
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i O MAXIMUM POWER OUTPUT LIMITED TO APPROXIMATELY 72% BASED ON SINGLE LOOP OPERATION DATA BASED ON AVERAGE POWER GENERATED PER DAY. THEREFORE, RECOVERY FROM A SCRAM THAT OCCURED WITHIN A 24 HOUR PERIOD WILL NOT INDICATE A ZERO PERCENT POWER LEVEL. _
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- 2. REPORTS f
t The reports provided in this document meet the requirements of Federal IO Regulations (10 CFR 50.59) and the W4P-2 Operating License. Complete data for the year 1986 has been included.
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2.B MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES This section contains information concerning main steam line safety / relief valve challenges for calendar year 1986 in accor-dance with the requirements of NUREG 0737, Item II.K.3.3.
TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATION POWER ASSOCIATED DATE COMPONENT ID (CODE) (CODE) (CODE) LEVEL LER 03/15/86 MS-RV-30 B D C 15% --
0 03/15/86 MS-RV-40 B D C 15% --
03/15/86 MS-RV-3C B 0 C 15% --
03/15/86 MS-RV-SC B D C 15% --
03/15/86 MS-RV-1A B D C 15% --
0 03/15/86 MS-RV-3A B D C 15% --
03/15/86 MS-RV-4A B D C 15% --
03/15/86 MS-RV-28 B D C 15% --
03/15/86 MS-RV-3B B D C 15% --
0 03/15/86 MS-RV-4B B 0 C 15% --
03/15/86 MS-RV-lC B D C 15% --
03/16/86 MS-RV-2A B D C 15% --
03/16/86 MS-RV-1B B D C 15% --
0 03/16/86 MS-RV-5B B 0 C 15% --
03/16/86 MS-RV-lC C + Hydroset B C 3% --
03/15/86 MS-RV-2C C + Hydroset B C 3% --
03/16/86 MS-RV-4C C + Hydroset B C 3% --
0 03/16/86 MS-RV-SC C + Hydroset B C 3% --
03/16/86 MS-RV-10 C + Hydroset B C 3% --
03/16/86 MS-RV-40 C + Hydroset B C 3% --
06/07/86 MS-RV-3C C + Hydroset B C 3% --
0 06/07/86 MS-RV-SC C + Hydroset B C 3% --
06/07/86 MS-RV-20 C + Hydroset B C 3% --
06/16/86 MS-RV-5B C + Hydroset B C 3% --
06/16/86 MS-RV-1A C + Hydroset B C 3% --
06/16/86 MS-RV-1B C + Hydroset B C 3% --
0 06/16/86 MS-RV-28 C + Hydroset B C 3% --
06/15/86 MS-RV-3B C + Hydroset B C 3% --
06/16/86 MS-RV-4B C + Hydroset B C 3% --
_g 06/29/86 MS-RV-1 B C + Hydroset B C 3% --
06/29/86 MS-RV-3B C + Hydroset B C 3% --
06/29/86 MS-RV-5B C + Hydroset B C 3% --
2-3 0
O.
i 2.8 MAINSTEAMLINESAFETY[RELIEFVALVECHALLENGES(Continued)
CODES:
O Type of Actuation A. Automatic B. Remote Manual C. Spring O
Plant Condition A. Construction B. Startup or Power Ascension Tests in Progress C. Routine Startup O 0. Routine Shutdown E. Steady State Operation F. Load Changes During Routine Operation G. Shutdown (Hot or Cold)
H. Refueling O Reason for Actuation A. Overpressure
- 8. AOS or Other Safety System C. Test
! 0. Inadvertent (Accidental / Spurious)
E. Manual Relief 19 r
- 1) Remote manual actuations occurred in support of acoustic monitor
.g NOTES:
position indication calibration testing required by Technical Specifications LCO 3/4.4.2.
- 2) Spring set testing was performed in accordance with ASME Section XI and Technical Specifications requirement in appifcabiitty paragraph 4.0.s.
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2.C StMERY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS GENERATOR GUTAGE OFF-LINE CAUSE SituTDOWN LER DATE TYPE HOURS CODE ETH00 NutBER SYSTEM COW ONENT CAUSE AND ACTION TO PREVENT RECURRENCE 7/1/85 F 0 A 5 -- CB PUWX Power output limited to 72% due to thru inoperability of "B" Recirculation 3/14/86 Pump due to excessive vibration.
3/14/86 F 16.2 G 3 86-003 RB INSTRU The reactor scranused at 51% power during the perforiaance of the Average Power Range Monitoring (APRM) channel surveillance. Another APRM channel was inadvertently placed in a test mode with the logic system in test not reset causing a full RPS actuation.
3/16/86 5 381.7 F 3 86-004 RB INSTRU During reactor shutdown, a pressure transient was caused by the pressure control system. This transient caused m the nuclear instrumentation Intermedi-e ate Range Monitor (IRM) channels G and m 11 to exceed their high upscale trip setpoint. Unit remained off-line due to a lack of power demand.
3/31/86 5 2208 C 9 -- -- -- Unit entered scheduled refueling /
thru maintenance outage. During the month 6/10/86 of June, startup criticals were per-formed for training purposes.
6/10/86 5 48.9 8 1 -- HA ECFUN Generator was removed from the grid to perform turbine overspeed testing.
O O O O O O O O O O O 2.C StItuRY & FLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER RE00CTIONS (Continued)
I 1
l GENERATOR OUTAGE OFF-LINE CAUSE SHUTDOW1 LER DATE TYPE HOURS CODE METHOD NU!EER SYSTEM COIPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 6/15/85 5 23.1 B 1 --
MS VALVEX Reactor power was decreased to 30% and i the generator was removed from the grid for main generator voltage regu-lator stability testing. While at a reduced power level, mechanical tur-bine overspeed testing and SRV testing was performed.
l 6/21/86 F 197.0 A 1 86-021 CF VALVEX The plant was shutdown to repair a RHR system primary coolant isolation valve due to excessive leakage.
l l
l 7/10/86 F 17.1 G 3 86-023 EG INSTRU A reactor scram occurred due to momen-i tarily reaching the APRM high neutron flux setpoint of 125% of indicated
" scale. The high flux condition was
- caused by turbine governor valve
- closure at 100% power dae to inadver-tent loss of DC power to the 500KV main generator output breaker control circuitry when an incorrect circuit breaker was operated during electrical ground isolation.
7/15/86 5 0 H 5 N/A RB CONR00 Decreased power to perfonn control rod sequence exchange.
9
O O O O O O O O O O O 2.C SIM4ARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS /POER REDUCTIONS (Cec.tir.ued)
GENERATOR OUTAGE OFF-LINE CAUSE SHUTCOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 7/25/86 F 17.7 A 3 86-025 HA VALVEX The reactor scrammed at 72% power on RPV high pressure while performing weekly turbine valve tests. A broken pin to linkage on a governor valve caused a false valve position signal which rendered DEH system incapable of controlling pressure. The pin was replaced and the plant returned to operation.
9/2/86 F 67.8 A 3 86-030 CH TURBIN Reactor scrammeo due to low water level from loss of Reactor Feed Pump 1B Turbine as a result of an erroneous electronic overspeed trip signal. A separate failure of a single reactor recirculation flow control valve to runback precluded the capacity of the
'd remaining RFP to maintain level. The failed components were replaced and the plant was returned to service.
9/12/86 S 5.2 B 1 ---
EB CKTBRK The generator was removed from service to correct a high temperature condition on the transformer side of generator disconnect.
9/26/86 5 0 H 5 ---
RB CONROD Decreased power to perform control rod sequence exchange.
11/10/86 F 0 A 5 ---
CB PUMPXX Reactor Recirculation Pump 1A was removed from service due to excessive vibration.
O O O O O O' O O O O- O s
2.C SIM4ARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Cctinued)
ENERATOR OUTAE OFF-LINE CAUSE SHUTDOWN LER DATE TYPE HOURS CODE ETH00 NUleER SYSTEM C0ff0NENT CAUSE AND ACTION TO PREVENT RECURRENCE 11/20/86 F 122.5 D 1 86-037 CC INSTRU During a plant shutdown due to lack of 86-038 qualification of a relief valve acous-tic monitor connector (see LER 86-037),
an automatic scram occurred at 157, power. The cause of the scram was low RPV water level after loss of a Reactor Feessater Pump on low suction pressure (see LER 86-038).
11/10/85 F 0 A 4 -- CB PUf@XX Power output limited to 727, based on thru singleloopoprationddetoexcessiva 12/31/86 vibration of A" Reactor Recirculation Pump.
.sa a
m TOTAL GENERATOR OFF-LINE CAUSE CODE TOTAL FOR 1986 HOURS A 6 282.5 8 3 77.2 C 1 2,208.0 D 1 122.5 F 1 381.7 G 2 33.3 H 2 0 s ., , s TOTAL 3,105.2 -
's
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O O O O O- 0 0 O'- 0 2.C SIDNutY OF PLANT OPERATION INCLUDING UNIT SHUTD06tlS/ poler llEDUCTIONS (Continued)
SINNutY OF CODES OUTAGE TYPE ,
F- Forced -
S- Scheduled CAUSE CODE A- Equipment Failure B- hintenance or Test C- Refueling D- Regulatory Restriction E- External Cause n
. F- Administration e
G- Personnel Error H- Other SHUTDOWN PETH00 1- m nual 2- mnual Scram 3- Auto Scram 4- Continued 5- Reduced Load 9- Other
F 0 O O O O O O- O -02
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k- 2.C :I ui44ARY S OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS /POWE_R REDUCTIONS (Continued)
SYSTEM CODE STANDARD CODE SYSTEM DESCRIPTION CA Reactor Vessels & Appurtenances CB' ' Coolant Recirculation Systems & Controls CF Residual Heat Removal Systems & Controls CH Feedwater Systems & Controls IA Reactor Trip Systems.
EA Offsite Power Systems & Controls m EB AC Onsite Power Systems & Controls EG Other Electric Power Systems & Controls HA Turbine Generator & Controls HJ Other Features of Steam & Power Conversion Systems (not included elsewhere)
MS Main Steam System RB Reactivity Control Systems
O O O O O O .O O O. O O 2.C
SUMMARY
OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIONS (Continu:d)
COMPONENT CODE COMPONENT TYPE / CODE CORPONENT TYPE INCLUDES:
Circuit Closers / Interrupters Circuit Breakers (CKTBRK) Contactors Controllers Starters Switches (other than sensors)
Switchgear Control Rod Drive Mechanism Control Rod Drive Mechanism (CONR00)
Instruinentation and Controls Controllers (INSTRU) Sensors / Detectors / Elements Indicators m Di f ferentials i
Integrators (Totalizers) 3 Power Supplies Recorders Switches Transmitters Computation Modules
< Penetrations, Primary Containment Air Locks (PENETR) Personnel Access Fuel Handling Equipment Access Electrical Instrument Line Process Piping Pipes, Fittings Pipes (PIPEXX) Fittings Pumps Pumps (PUMPXX)
2.C Suff4ARY OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER REDUCTIO;is (Continued) l C0!@0NENT CODE (continued) i COMP 0NENT TYPE / CODE COMPONENT TYPE INCLUDES:
i Relays Switchgear j (RELAYXX) 4 Transforiners Transformers -
j (TRANSF)
Turbines Steam Turbines (TURBIN) Gas Turbines i Hydro Turbines Valves Valves i (VALVEX) Dampers N
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O O O O O O O O O O O.
2.D
SUMMARY
OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT EQUIPMENT REQUIRING liAINTENANCE SYSTEM PROBLEM - ACTION TAKEN CRD-V-186 Control Rod Drive Cables for valve have Replaced cables with same kind; shortened prematurely degraded due to surveillance time to monitor cable degradation.
heat generated by continuously-energized solenoid valve.
RPS-RLY-K10A Reactor Protection Current annunciator logic does Rewired annunciator logic, tested and returned thru K10H System not indicate subchannel trip to service.
indication for the MSIV and turbine throttle valve closures.
MSLC-M0-4 Main Steam Valve blows fuses when Replaced torque switch and fuses, functionally Leakage Control closing. tested and returned to service.
MSLC-M0-2D Main Steam Valve will not open on demand. Tightened limit switch terminations and Leakage Control checked rotors / switches in limit switch for m looseness. Valve returned to service.
e g DMA-TI S-31 Diesel Gen. HVAC flow indication downscale Repaired broken resistor lead on circuit Bldg. HVAC low-repair or replace, board and recalibrated.
RHR-M0-68 Residual Heat Valve blows fuses when Installed new torque switch, adjusted Removal closing. limit switch and cycled valve to assure operability.
DCW-TS-12B2 Diesel Cooling Temperature switches found Replaced temperature switches with qualified 1101 Water outside acceptable tolerances. spares, recalibrated and returned to service.
1182 DSA-PS-SA Diesel Starting Low pressure alarms sealed in During troubleshooting, found pressure switches 6A Air with pressure normal. to be defective. Replaced pressure switches, recalibrated and returned to service.
e
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2.0 SUMitARY OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING ACTION TAKEN SYSTEM PROBLEM MAINTENANCE FDR-V-3 Radioactive Floor Valves are leaking excessively. Removed rust and scale from seating surfaces, 4 Drains blued gate to seat, tested and returned to service.
LD-RLY-K01A,0 Leak Detection Presently installed relays Removed relays, installed new seismically-K02A,B not seismically qualified. qualified relays, timed tested and returned K03A,B,C,0 to service.
K04A,B Main Steam Leakage Valve will not complete full Installed new torque switch, performed leak MSLC-M0-3B Control closure. rate test and returned to service.
CRD-R00-4607 Control Rod Drive Position indication probe for Replaced position incore probe, verified CRD 4607 is inoperable. operability and returned to service.
Contact in level switch did Cleaned switch contacts and successfully MS-LIS-36C Main Steam
[ not make contact during reperformed channel functional test.
+ performance of a Channel Functional Test.
Standby Liquid The square root extractor for Replaced transistors QS and Q7. Recalibrated SLC-SQRT-1 instrument and returned to service.
Control SLC is out of service.
Residual Heat Valve blows control power During troubleshooting, a loose connection RHR-M0-8 Removal when attempting to operate. was found between terminals at the motor control center. Reterminated connections and returned to service.
Standby Gas Valve will not cycle. Replaced fuses at motor control center SGT-V-381 and found motor leads reversed. Rewired Treatment correctly and returned to service.
g 2.0
SUMMARY
OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Ccntinued)
EQUIPMENT REQUIRING ACTION TAKEN MAINTENANCE SYSTEM PROBLEM WOA-EH0-528 Radwaste Bldg. Found 8 oz. of oil in end Replaced 0-rings, repaired cable, replaced HVAC cover of operator and one oil and returned to service.
of the leads for the dump valve damaged.
LPRM-DET-54 DAB Local Power Cables to these LPRMs were Replaced both the connectors and pins on all Range Monitoring broken during replacement of LPRMs listed. Ran an insulation integrity 4025A 2417A CRD mechanisms. and returned to service.
4017A CMS-PT-7 Containment Drywell pressure transmitters Perfonned a channel calibration and found a 8 Monitoring do not appear to be responding wiring problem. Properly wired the instrument, properly to changing system completed calibration and returned to service.
conditions. l MS-P0T-3A Main Steam Replace preamps in acoustical Replaced charge converters per manufacturer's 4A monitoring sensors to meet procedures, verified operation and returned 40 equipment qualification to service.
m requirements.
D0-DPS-12Al Diesel Oil Fuel filter high d/p alarm Installed snubbers in sensing line for 12A2 comes in when d/p is low, referenced pressure switches, leak checked 13Al caused by flow oscillations. and returned to service.
13A2 Main Steam Pressure recorder indication Cleaned, realigned gear shafts, performed MS-LR/PR-623B is drifting. surveillance procedure and returned to service.
j MS-V-38L Main Steam During the perfonnance of Removed worn 0-rings and installed new 0-rings.
38N a surveillance test, 0-rings Lubricated hinge shaft and bushings. Checked.
38J were found to be worn for proper opening pressure and returned to 37V excessively, service.
O O O O O O O O O O O 2.D ~SUlHARY OF SIGNIFICANT f%INTENANCE PERFCRMED ON SAFETY RELATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING MAINTENANCE SYSTEM PRDBLEM ACTION TAKEN CAC-RV-65A Containment Relief valve failed leakage Cleaned valve, machined nozzle, lapped disk 63B Atmospheric criteria and set pressure and bench tested satisfactorily. Returned to Control was found low. service.
CIA-V-33B Containment ADS accumulators are leaking Cut body to bonnet seal weld and found sand Instrument Air back through check valves. blast grit and metal particles in bottom of val ve. Cleaned out valve and lapped disk.
Reassembled and welded. Returned to service.
CRD-SPV-ll8/E427 Control Rod Drive Scram valve solenoid chatters Replaced pilot head assembly and verified no when energized. leaks. Tested valve operation and returned to service.
CMS-RR-27E Containment While performing surveillance ' Installed new recorder servo amplifier and m Monitoring procedure, recorder would not performed calibration. Completed surveillance
, respond. procedure and returned to service.
DIM-TIS-32/1 Diesel Bldg. HVAC Found temperature switch Replaced temperature switch with direct inoperable during performance replacement, calibrated and returned to of PM. service.
RCIC-T-3 Reactor Core Steam line drain trap is not Disassembled trap, cleaned and flushed.
Isolation Cooling passing adequate flow. Reassembled and returned to service.
FPC-FLY-FCV/1 Fuel Pool Cooling Replace and test existing Replaced existing relays with Agastat time V175/1 time delay relays with QC-I. delay relays, tested and returned to service.
(NRC commitment)
Dl%-AD-51 Diesel Bldg. HVAC Suction damper linkage is Replaced broken swivel, adjusted damper and broken, returned to service.
0 0 0 0 0 0 0 0 0- O . O.
2.0
SUMMARY
OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continu!d)
EQUIPMENT REQUIRING MAINTENANCE SYSTEll PROBLEM ACTION TAKEN RCIC-V-69 Reactor Core Valve did not meet closing Inspected and cleaned stem and packing gland Isolation Cooling stroke time requirements, area. Cleaned contacts inside valve operator.
Retested within Technical Specification limits.
C I A-V-21 Containment Check valve leak in excess Disassembled cleaned and lapped valve disk.
Instrument Air of 20,000 SCCM. Unable to successfully repair valve. Cut out and replaced valve with a direct replacement.
Leak checked and returned to service.
DLO-P-3Al Diesel Lube 011 Pressure indicator reads low Found oil inside gauge; unable to calibr?te.
with system pressure normal. Replaced indicator with direct replacement, calibrated and returned to service.
DLO-M-P/2B2 Diesel Lube Oil Annual preventative maintenance Replaced motor per plant procedures, tested schedule calls for replacement and returned to service.
m of motor.
[ DSA-RV-15A,B Diesel Starting Relief valves are not qualified Replaced originally installed relief valves u 16A,B Air and adequate documentation is with valves which are seismically qualified.
17A,B unavailable.
18A,B SGT-DPIC-1A/1 Standby Gas Vortex damper does not allow Adjusted damper linkage to allow damper to Treatment 100% stroke - adjust as close 100%, tested and returned to service.
necessary.
DE-FLX-2Al Diesel Engine Flex installed for engine Replaced previously installed flex with the is the wrong size. correct size, torqued to 100 ft./lbs. and returned to service.
ROA-AO-V2 Reactor Bldg. HVAC Response time of damper is Inspected, cleaned and lubricated damper.
too slow. Retested satisfactorily.
g g g g g-- .
g 2.0 SUWiARY OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continusd)
EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN DSA-V-12A,B Diesel Starting Valves are installed upsidedown Removed valves and reinstalled in the proper 13A,B Air from manufacturer's recom- orientation. Tested and returned to service.
15A,8 mended positioning.
16A,8 i
i DSA-B3-ClA2 Diesel Starting Top of the battery is cracked Removed old battery and replaced with a Ai r between cells. new direct replacement.
DSA-PI-ll Diesel Starting Presently installed pressure Replaced instruments with new ones which
! 12 Air indicators are not environ- were environmentally qualified, calibrated mentally qualified. -
and returned to service.
DMA-TIS-31/1 Diesel Bldv. HVAC Local indication for outside Repaired a broken resistor on the circuit board 4 TIC-31/2 air into HPCS room is pegged for TIS-31/1 and installed a new controller for low. TIC-31/2. Recalibrated both instruments and returned to service, m
b LD-TS-6148 Leak Detection Temperature switch input wire Determed, relugged and relanded wire.
- broken. Performed a functional check and verified proper calibration.
RFW-M0-65B Reactor Feedwater Found full length vertical Blocked valve stem to hold valve closed, crack on the gearbox of motor Removed from steam tunnel and found pivot operator. fork broken, bearings damaged and upper
' bearing support flange broken. Replaced motor operator, set limit switches, torque switch and returned to service.
CRD-HCU-1023 Control Rod Drive When HCU was tagged out with Disassembled rod and detector display and.found 0 psig, there were no "ACCUM a broken wire on SCRAM ACCUM indicator.
TRBL lights on. Repaired by soldering. Reassembled rod and detector display.
1
Ci O O O O O O O O O O 2.D SUf41ARY OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN SGT-M-FN/lB1 Standby Gas Fan did not auto start on a While troubleshooting conkol circuit, a Treatment "Z" actuation while performing loose terminal at 4CR relay was discovered.
surveillance procedure. Tightened and checked all wires on relays.
Satisfactorily completed surveillance test.
CRD-DRVE-2227 Control Rod Drive Position indicators for drives Runoved and replaced Position indicator CRD-DRVE-0619 listed are not displaying probes with functionalsy verified spares.
CRD-DRVE-5419 current rod loc. tion at all Tested and returned to service.
CRD-DRVE-3443 positions.
CRA 7828 Containment Drywell head recirculation Replaced fan motor breaker, tested and
- Recirculation Air fans breaker would not close. returned to service.
CRD-PS-130/1447 Control Rod Drive Trouble alarm will not clear Found switch to be out of calibration on CRD-PS-130/5419 although local indication 14-47; recalibrated satisfactorily. Unable N is acceptable. to recalibrate 54-19; replaced with new switch
' and recalibrated.
CRD-DRVE-3827 Control Rod Drive Drives are either running hot, Removed drives from vessel and replaced with CRD-DRVE-3831 have high stall flows or other rebuilt spares. Overhauled drives that were CRD-DRVE-3451 operational problems. removed are to be used as spares for next .
CRD-DRVE-2651 outage.
i CRD-DRVE-2623 CRD-DRVE-2223 CRD-DRVE-2251 RilR-V-154A Residual Heat Valve is leaking 355 SCCM. Disassembled valve and found rust trapped on Removal seat. Lapped seat, machined disc, reassembled and repacked. Retested successfully.
MSLC-M0-3B Main Steam Valve will not close Installed new torque switch and adjusted to Leakage Control completely, previous torque setting.
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O O O O O' .O O O O O O 2.D SUt44ARY OF SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY RELATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING SYSTEtt PROBLEM ACTION TAKEN MAINTENANCE CRD-HCU-2647 Control Rod Drive Drive mechanisms stroke times Adjusted directional control valves on HCUs CRD-HCU-1039 are not within acceptance until time tests of drives were within CRD-IICU-3035 criteria, specified limits.
CRD-HCU-3855 CRD-l:CU-3451 CRD-HCU-2251 CRD-V-111/3803 Control Rod Drive N2 charging valve leaks Depressurized HCU, replaced valve and excessively. repressurized the accumulator.
MS-POT-3D Main Steam Replace preamps in acoustical Replaced originally installed preamps with MS-POT-4A monitoring sensors to meet environmentally qualified equipment.
11S-P0T-40 equipment qualification requirements.
FPC-RMS-P/lB1 Fuel Pool Cooling Pumg will start with the C/S During a wire check of the control circuit,
' in Start", but when the C/S a wire was found to be landed on the wrong E$ spring returns to auto the terminal. Correctly wired circuit and pump stops. verified operability.
- MS-SPV-SCA Main Steam Solenoid wires were reported Replaced connectors, verified operability separated from the connector. and returned to service.
! MS PLY-ADKSAA Main Steam While performing surveillance Relay failed and would not adjust to setpoint.
procedure, relay did not pickup Replaced relay, set correct timing, tested and within allowable time frame. returned to service.
LD-TS-601E Leak Detection Temperature switch will not Replaced defective switch with a qualified calibrate. spare, calibrated and returned to service.
RCIC-PI-601 Reactor Core Indicated pump speed is too During troubleshooting, the flow monitor was RCIC-FI-600/1 Isolation Cooling high for flow and pressure, found to be inoperable. Replaced and recalibrated.
DG-ENG-1Al,lA2 Standby Electrical 6 year scheduled maintenance Disassembled and inspected engines. Replaced DG-ENG-1Bl.182 due on all diesel generator seals, requalified internal components, DG-EHG-C llPCS Electrical engines, refurbished all cylinder heads, replaced rocker arm bushings, reassembled, tested and returned to service.
)
2.E INDICATIONS OF FAILED FUEL
) Plant operation in 1986 produced no evidence of any fuel failures. No indication of fuel failure was detected from main steam line or offgas radiation monitors during the course of plant operation. Also, no indica-tions of fuel failure were observed through normal P1 ant chemical or radio-logical analysis.
) In accordance with the WNP-2 FSAR, Section 4.2.4.3, a visual inspection of discharged fuel assemblies from Cycle 1 was performed and a special report was submitted to the NRC. Subsequent routine reports of this nature will be submitted to the NRC by means of the annual report as agreed to by the Supply System, and J. Bradfute, S. Phillips and U. Chang of NRR. In the event the inspection reveals significant degredation of fuel assemblies,
) the Supply System will notify the NRC by means of a special report or the appropriate regulatory reporting process (e.g., LER). This information is supplied in accordance with requirements as set forth in Regulatory Guide 1.16.
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10 2.F PLANT CHANGES AND TESTS
-O Federal Regulations (10CFR50.59) and the Facility Operating License (NPF-21) allow changes to be made to the facility as described in the safety analysis report and tests or experiments to be conducted which are not described in the safety analysis report, without prior Nuclear Regula-tory Commission (NRC) approval, unless the proposed change, test, or exper-iment involves a change in the Technical Specifications incorporated in the
.O license or an unreviewed safety question. In accordance with 10CFR50.59, sumaries of the changes performed and tests or experiments conducted in 1986 are provided. Included are summaries of the safety evaluations.
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O 2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES The following plant modifications and design changes were completed in 1986 O and reported in accordance with 10CFR50.59. These modifications were eval-uated and it was determined that they did not (a) increase the probability of occurrence of an accident or malfunction of the equipment important to safety, as previously evaluated in the WNP-2 updated Final Safety Analysis Report (FSAR), (b) create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR, (c) reduce the mar-O gin of safety as defined in the basis for any WNP-2 Technical Specifica-tions, or (d) require a change to the WNP-2 Technical Specifications.
PLANT DESIGN CHANGE 83-0016
- O Plant Design Change 83-0016 was initiated to satisfy License Condition Number 20, Section 9.5.6, SSER #4, which installs equipment designed to remove moisture from the Diesel Starting Air Systems prior to startup following the first refueling outage.
This plant design change added air dryers to each of the three Diesel
~O Starting Air Systems. Each dryer system consists of an air cooler, air dryer, drain trap and post filter. All components including the dryers are static and require no electrical input.
This modification did not result in a change to the WNP-2 Technical Speci-fications. An unreviewed safety question was not involved because the additi n f the dryers assures the dewpoint of the air put into the system O is low, preventing condensate from forming downstream of the dryers. This enhancement provides assurance that moisture or corrosion products will not reach the air start motors, thereby increasing overall reliability of the Diesel Starting Air System. Moisture accumulation was addressed up to R-1 by a daily receiver manual moisture removal process.
'O PLANT DESIGN CHANGE 83-0050-0A l
Plant Design Change 83-0050 was initiated to satisfy License Condition 12 which states that " prior to startup following the first refueling outage, n the licensee shall install, test, and have operable the alternate remote
'V shutdown system."
This plant design change provided for the installation of an independent alternate shutdown system using only Division I components. The design included safe shutdown controls for Residual Heat Removal, Service Water, three Safety Relief Valves and applicable process instrumentation.
This modification did not result in a change to WNP-2 Technical Specifica-tions or involve an unreviewed safety question because the modification was performed in accordance with ADC 19 as presented in the WNP-2 SER.
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,O' 2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 83-0111 Plant Design Change 83-0111 was initiated to increase the efficiency and reliability of the air start logic for the No.1 and 2 Diesel Generators.
This plant design change replaced circuit logic that transferred between air start motors until the engines reached 150 RPM with an independtnt and
.O constant air supply to all air motors during engine startup.
This modification did not result in a change to WNP-2 Technical Specifica-tions. Although the starting logic was significantly modified, it was done to make the logic more reliable and, therefore, dees not involve an unreviewed safety question.
O .
PLANT DESIGN CHANGE 84-1162 Plant Design Change 84-1162 was initiated to satisfy Regulatory Guide 1.97, Revision 2, which states that "the licensee shall implement (installation 10 or upgrade) requirements of R.G.1.97, Rev. 2, with the exception of flux monitoring prior to startup following the first refueling outage." This modification also is related to an original license issue, dated December 20, 1983, Attachment 2, Item 3.
This plant design change provided direction for the installation of an
- O environmentally qualified H 02 2 sampling system, including sample probes in containment and sample lines to two divisionally separate sample racks containing pumps, analyzers, indicators and controls. This modification also provides operators with the capability to remotely calibrate monitored parameters from the main control room.
A Technical Specification change was made in regards to the concentration
- O of the samples used to calibrate the hydrogen and oxygen monitors, but it did not address the hardware modification of the system. The plant modi-fication was within the boundaries of the original safety analysis and, thereby, does not involve an unreviewed safety question.
PLANT DESIGN CHANGES 84-1210, 84-1211, 84-1213, 84-1214 and 85-0008
! Plant Design Changes 84-1210, 84-1211, 84-1213, 84-1214 and 85-0008 were initiated to design and install the Fuel Pool Cooling modification and upgrade.
O These plant design changes added motor operated valves; instrumentation to
, monitor level, temperature and pressure; annunciation in the main control l room; and associated electrical components. These modifications were a result of the evaluation done by the Supply System of the Secondary Con-tainment Pressurization events and the consequences. Following the analy-sis, it was concluded that the Fuel Pool Cooling System needed to be O upgraded to Seismic Class I and Quality Class I.
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2.F.1 PLANT 'iODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGES 84-1210, 84-1211, 84-1213. 84-1214 and 85-0008 iO (Continued)
These modifications did not involve an unreviewed safety question because the upgrade of the system lessens the possibility of an accident or mal-function. A change to WNP-? Technical Specifications was made which reflected pertinent fleets of the system upgrade.
O PLANT DESIGN CHANGE 84-1424 Plant Design Change 84-1424 was written to reduce the blowdown time in the event of a Reactor Core Isolation Cooling (RCIC) line break.
This plant design change provided direction to change the time delay relays from three seconds to two seconds and provided testing directions to assure that inadvertent system isolation does not occur.
During the review process, it nas determined that the associated relays would be added to WNP-2 Technical Specifications. The modification does O not involve an unreviewed safety question, because the margin of safety is increased by reducing the steam line isolation delay time.
PLANT DESIGN CHANGE 84-1486 O Plant Design Change 84-1486 was initiated because operation of hotwell sample pumps 17B and 17C affected the hotwell level indications, causing major system perturbations.
Originally, three sample pumps were installed to provide three separate sample points. Through analysis of the samples, it was found that the
.O samples taken from the three different points were the same and that only one sample pump was necessary. This plant design change provided for the renoval and storage of pumps 178 and 17C, which are currently designated as spares.
This modification did not involve a change to WHP-2 Technical Specifications
- O or involve an unreviewed safety question, because the equipment that was spared was redundant and non-safety related.
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0-2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 84-1623 O'
Plant Design Change 84-1623 was initiated because the fire protection jockey pump FP-P-3 had failed twice with the cause considered a design error.
This plant design change increased the size of the fire protection jockey O pump from a design capacity of 50 gpm to 220 gpm, which provided better overall system reliability and performance.
This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question because the modification was within the confines of the previous safety analysis boundaries.
O PLANT DESIGN CHANGE 85-0131 Plant Design Change 85-0131 was written to ensure the required primary con-tainment temperature is met by supplying cooler water to the Reactor Closed
.O Cooling (RCC) Heat Exchangers during the hot summer months.
This plant design change added piping and wier boxes directing Tower Makeup (TMU) water directly to the suction of the Plant Service Water (TSW) pumps, which are the source of water for the RCC heat exchangers. This modifica-tion reduces the water temperature at the pump suction by approximately O 20'F during the summer.
This modification did not involve a change to the WNP-2 Technical Specifi-cations and does not involve an unreviewed safety question, because TMU and TSW are not safety related systems and do not directly affect the operation of any safety system.
O PLANT DESIGN CHANGE 85-0144 Plant Design Change 85-0144 was initiated to permit storage of combustible material on the 487' level of the Radwaste Building by extending the fire protection sprinkler system to this area.
!O
- This plant design change provided for installation of piping, hangers and j sprinkler fittings required to allow storage of combustible material at
! that location.
I This modification did not involve a change to the WNP-2 Technical Specifi-
! O cations and does not involve an unreviewed safety question, because the
! system will provide fire protection equal to or greater than what is cur-l rently installed, thereby, increasing the margin of safety. ;
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L 2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0155 Plant Design Change 85-0155 was initiated to delete automatic control of Turbine Building Exhaust Air (TEA) fan exhaust volume dampers.
This plant design change removes pressure transmitters, controllers and switches and provides direction for the caping of tubing lines. The auto-
) matic control of the Turbine Building supply fans have not been in use for over a year, because the supply and exhaust controllers were incompatible, thereby, causing unnecessary system oscillations.
This modification did not involve a change to the WNP-2 Technical Specifi-cations. This change does not involve an.unreviewed safety question,
) because modification of on/off control logic of TEA fan exhaust dampers is within the scope of the analysis as described in the FSAR.
PLANT DESIGN CHANGE 85-0181
) Plant Design Change 85-0181 was initiated to increase the longevity of the High Pressure Core Spray (HPCS) battery and to reduce the time required to restore the battery to a fully charged condition.
This plant design change reduced the number of battery cells on HPCS-B1-DG3 from 60 to 58. The modification enables equalization of the battery at a
) voltage recomended by the manufacturer without exceeding system rated voltage, thereby, increasing the life expectancy of the battery.
This modification did not result in a change to WNP-2 Technical Specifica-tions and does not involve an unreviewed safety question, because the reduction in the number of cells will not reduce battery capacity below
) design requirements.
PLANT DESIGN CHANGE 85-0278 Plant Design Change 85-0278 was written to improve the operation of the Plant Service Water (TSW) oumps and allow remoti starting capability of the
) pumps without local operator action.
This plant design change added automatic air vents to the TSW pump dis-charge piping to permit venting during pump startup. The modification also provided an alternate source of TSW lube water to ensure a continuous water source is available to the pumps at all times.
)
This modification did not involve a change to the WNP-2 Technical Specifi-cations. This change does not involve an unreviewed safety question, because the modification does not change the basic operation of the system and the plant service water system is not a safety related system.
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2.F.1 PLANT '4001FICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0446
,v Plant Design Change 85-0446 was written because analysis of reactor trips indicated the necessity to delay the automatic transfer of feedwater level control from three elements to a single element in order to maintain reactor level below the high level trip point.
O This modification added a relay to delay the automatic transfer of the reactor level control system for a period of 40 seconds to prevent high reactor water level caused by a flow mismatch during trip conditions.
This modification did not involve an unreviewed safety question because the feedwater system is not considered an Emergency Safeguard System and is not O relied upon for reactor water inventory as desc"ibed in the FSAR. Thi s modification increased reliability of the feedwater system, thereby, improving overall plant safety. A change to the WNP-2 Technical Specifica-tions was not necessary as a result of this modification.
O PLANT DESIGN CHANGE 85-0393 AND 85-0465 Plant Design Changes 85-0393 and 85-0465 were initiated to relocate seismic monitors which were originally installed on the HPCS injection line piping and on a Standby Service Water pipe. Both monitors were subjected to system operation related vibration which had caused damage to the instruments.
O Both plant design changes provided for the relocation of seismic monitors which were originally mounted on the HPCS injection line piping and on a Standby Service Water pipe. These monitors were moved to more stationary locations, each in the same general area of the original installation to reduce the amount of movement to which the monitors were subjected. Plant O Design Change 85-0393 also added protective covers to the monitors to pre-vent damage to the instruments from water and provide general protection.
These covers were installed on all monitors located throughout the site.
These modifications did not involve a change to the WNP-2 Technical Speci-fications or involve unreviewed safety questions, because the modifications enhanced the design function of the system and are within the scope of O
previous safety analysis.
PLANT DESIGN CHANGE 85-0666 O
Plant Design Change 85-0666 was initiated to reduce the probability of vibration induced failure of the reactor recirculation flow control hydraulic lines by replacing them with metal flexible hoses.
This plant design change was initiated because the reactor recirculation system flow control lines were subjected to vibration and have failed.
This modification replaced the originally installed piping with flexible O hoses from the terminal end support to the flow control valve hydraulic actuators on both the "A" and "B" recirculation lines.
t 2 - 28 O
l 2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0666 (Continued)
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' This modification did not result in a change to the WNP-2 Technical Speci-fications or involve an unreviewed safety question, because the modifica-tion reduces the probability of a failure, thereby, increasing the margin of safety. Additionally, this modification does not introduce the possi-l bility of an accident not previously analyzed.
)
PLANT DESIGN CHANGE 85-0706 Plant Design Change 85-0706 was initiated in response to IEN 85-47, "Envi-ronmental Qualification Tests of Electrical Terminal Blocks." This notice
)> was published because high humidity caused current leakage at terminal strips and low voltage / current signals could be affected enough to provide erroneous readings.
This modification affected terminal strips for Class 1E analog signals and penetration terminations which were inside containment. These terminal strips were replaced with spliced connections and covered with qualified heat shrink tubing.
This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question. Replacement of the term-inal blocks with spliced connections and heat shrink tubing improves the reliability of the system and does not alter designed system function.
)
PLANT DESIGN CHANGE 86-0002 Plant Design Change 86-0002 was a modification of the Reactor Recirculation pump stuffing box and bearing in response to a unit failure analysis and a
) joint design review completed by General Electric, Bingham Willamette and Supply Systems Engineering.
The plant design change provided for installation of a new pump bearing which modified load capabilities to increase overall unit strength within the pump bearing assembly. This modification also included installation of
) a replacement impeller and shaft purchased from the Black Fox unit.
This modification does not involve an unreviewed safety issue since no mod-ification was made to the unit which exceeded the previously addressed design criteria or analyses resolved within the FSAR. No modification was necessary to WNP-2 Technical Specifications as a result of this
) modification.
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2.F.1 PLANT MODIFICATIONS AND DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 86-0029 Plant Design Change 86-0029 was initiated to provide direction for the replacement of the single scram reset switch which was near the end of its
! service life. The originally supplied switch was no longer available from its vendor and no other manufacturers supplied switches that met the design requirements.
This plant design change replaced the single three position reset switch with a pair of two position switches. This modification split the original function of a single switch into two separate switches with no functional change to the system. The devices were purchased and installed in a fully l qualified manner to meet all seismic and quality class requirements.
) This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question, because no change was made to the system that altered its function as described in the FSAR.
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2.F.2 PLANT PROCEDURE CHANGES Procedures described in the WNP-2 Final Safety Analysis Report (FSAR) are used by the Plant Operating Staff and by various offsite support organiza-
) tions. In 1986, the Plant Staff made changes to procedures in accordance with 10CFR50.59 and concluded that none of the changes involved unreviewed safety questions.
Changes to procedures were generally either administrative or technical in nature. Administrative changes consisted of title, organizational and edi-
) torial changes, while technical changes were the result of system or com-ponent modifications or improvements in procedural processes. A safety evaluation was conducted for each change in accordance with 10CFR50.59 was reviewed and approved by the Plant Operations Committee and are available for audit as necessary. The review concluded that the probability of g occurrence or consequences of an accident or equipment malfunction were not
( increased, there was no reduction in any plant safety margins, and the pos-sibility of an accident or malfunction not previously evaluated was not
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increased.
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O 2.F.3 PLANT TESTS Tests performed at WNP-2 for the calendar year 1986 did not (1) place the unit in an unanalyzed configuration or condition not bounded by design O basis, or (2) perform an operation not described in the FSAR which could have an adverse effect on safety-related equipment or systems. A safety evaluation was conducted for each test in accordance with 10CFR50.59 was reviewed and approved by the P1 ant Operations Comittee and are avaliable for audit as necessary. The review of these tests concluded that the prob-n ability of occurrence or consequences of an accident or equioment malfunc-tion were not increased, there was no reduction in any plant safety mar-gins, and the possibility of an accident or malfunction not previously evaluated was not increased. This information is provided in accordance with 10CFR50.59.
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l C) 2.G REACTOR COOLANT ACTIVITY CUMULATIVE IODINE LEVELS This section contains information relative to reactor coolant cumulative C) iodine levels and iodine spikes. The specific activity of the primary
, coolant was significantly less than the Ifmits of (a) less than or equal to 0.2 microcurie per gram dose equivalent I-131, and (b) less than or equal to 100/E microcuries per gram as set forth in WNP-2 Technical Specifica-tions. A graph showing cumulative iodine dose equivalent for the calendar year 1986 is provided for reference and completeness. This information is
)C) provided in accordance with WNP-2 Technical Specifications.
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