ML20196G532
| ML20196G532 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/31/1987 |
| From: | Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 8803090226 | |
| Download: ML20196G532 (60) | |
Text
. --
G WXP-2 ANXUAL .
OPERATIXG REPORT 1987 WASHINGTON PUBLIC POWER SUPPLY SYSTEM
.N O
${,$3 $7iSSI NOUhl, f 400154
ANNUAL OPERATING REPORT OF WNP-2 FOR 1987 DOCKET NO. 50-397 I
FACILITY OPERATING LICENSE NO. NPF-21 i
- Washington Public Power Supply System l 3000 George Washington Way
! Richland, Washington 99352 l
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TABLE OF CONTENTS
1.0 INTRODUCTION
. . ................... 1-1 1.1 1987 Power History Graph for WNP-2 ....... 1-4 2.0 REPORTS ....................... 2-1 2.1 Annual Personnel Exposure and Monitoring Report .............. 2-2 2.2 Main Steam Line Safety / Relief valve Challenges . . . . . . . . . . . . . . . 2-3 2.3 Summary of Plant Operation ........... 2-8 2.4 Summary of Significant Maintenance Performed on Safety Related Equipment . . . . 2 - 19 2.5 Indications of Failed Fuel . . . . . . . . . . . 2 - 27 2.6 Plant Modifications . . . . . . . . . . . . . . . 2 - 30 2.6.1 Plant Design Changes . . . . . . . . . . . 2 - 31 2.6.2 Lifted Leads and Jumpers . . . . . . . . . 2 - 39 2.6.3 FSAR Amendment Evaluations . . . . . . . . 2 - 41 2.6.4 Other . . . . . . . . . . . . . . . . . . 2 - 43 2.7 Plant Tests and Experiments . . . . . . . . . . . 2 - 47 2.8 Plant Procedure Changes . . . . . . . . . . . . . 2 - 48 2.9 Reactor Coolant Activity Cumulative Iodine Levels . . . . . . . . . . . . . . . . 2 - 51 2.9.1 WNP-2 Dose Equivalent Iodine Graph . . . . 2 - 52 i
I
1.0 INTRODUCTION
The 1987 Annual Operai.ing Report of Washington Public Power Supply System Plant Number 2 (WNP-2) is provided as a supplement to the Monthly Operation Report.
This report is submitted in accordance with the requirements of Federal Regu-lations and Facility Operating License NPF-21. It should be noted that, for ease of reference and completeness, additional required reports are also included. WNP-2 is a 3323 MWt, BWR-5, which began commercial operation on December 13, 1984.
During the first part of 1987 (January-April), the plant continued single-loop operation due to the higher-than-acceptable vibrations experienced on Reactor Recirculation Pump "A". As a result of single-loop operations, power output was limited to 72 percent for that time frame.
On April 10, 1987, the plant was shut down for the annual maintenance and refueling outage. In late June-early July, af ter the outage, the plant experi-enced five unplanned automatic snutdowns (scrams) within a two-week period.
As a result of those scrams, Supply System management directed that the plant remain shut down pending a thorough review of plant operations. It was con-cluded from the review that the five scrams were caused by a series of unrelated mechanical and electrical f ailures that, for the most part, could not have been anticipated.
On July 26, 1987, the plant was restarted and operated for a record 133 con-secutive days until Cecember 6, when it was shut down for a scheduled three-day maintenance outage to repair an inoperable flow control valve in the Condensate Filter /Demineralizer System. The inoperable valve had made it necessary to reduce power to 85 percent every four to five days to backwash and precoat the remaining filter demineralizers. Since starting up on July 26, until the planned shutdown on December 6, the plant operated at an average of 91 perrent of its capacity. On December 10, the plant was restarted and ran at or near 100 percent capacity for the remainder of the year.
In November, indications of a probable pinhole leak in a single fuel pin were noted. The indications were noted during cycle 3 operation, approximately three days following completion of a control rod sequence exchange, when Of f-gas System post-treatment radiation monitor levels began rising. The increase was confi rmed by c.iemical analysis. During a subsequent sequence exchange, selected rod movements were performed in conjunction with on-line pretreatment sample gama spectroscopy analysis to attempt to isolate the location of the leak. Although iniilai results appear promising, further testing is planned in conjunction with the next sequence exchange (scheduled for mid April, 1988) to further isolate / confirm the location of the leak. The current plan is to perform fuel sipping during the upcoming refueling outage (R-3) to identify the leaker and remove it f rom the core. It should be noted that, to date, no further increases which would indicate more leaking fuel pins have been noted.
4 1-1
[
Ouring 1987, there were several examples of major accomplishments which (
~
required significant effort on the part of Supply System personnel to success-fully complete. The following is a summary of those efforts:
(a) The second refueling outage was successfully completed. Significant i activities included: l o Modification and reinstallation of the two reactor recirculation pumps. Higher-than-acceptable vibrations caused by an internal design defect required that the plant be operated at reduced power since late 1985. Stronger wear rings and bearing assemblies were installed in each pump to solve this problem.
o Removal of spent fuel assemblies and refueling the reactor. The <
refueling activity included replacing 148 fuel assemblies, using a fuel shuffle scheme.
o Removal, inspection, cleaning and reinstallation of two of the three {
low pressure main turbine rotors.
- Tube bundles were replaced and improvements made internally to both two-stage Moisture Separator Reheaters. ;
i
- A new moisture preseparator system was installed for excess moisture removal from steam leaving the high pressure turbine, o Installation of a new plant process computer. The new computer will enable Operations personnel to more rapidly identify trends that could lead to problems in plant systems. This state-of-the-art system processes information faster than the computer it replaced ,
and has added features such as color graphics display and the ability to provide operators with an historic analysis of an event.
(b) WNP-2 continued to have an excellent record for limiting worker radiation exposure. In 1987, total radiation exposure at the plant was 406 man-rem. Of that total, 98.7 man-rem was attributed to activities associated with repair of the two reactor recirculation pumps. To be rated in the -
top 25 percent of BWRs for 1987, the Institute of Nuclear Power Opera-tions (INPO) has established the limit of 543 man-rem. In addition, INPO r has set 460 Nn-run as the industry goal for 1990 for BWRs.
(c) WNP-2 was recognized by the National Academy for Nuclear Training with r the accreditation of seven nuclear training programs and the acceptance of the Supply System as a member of the academy. The Supply System is l one among 18 of 60 utilities with fully accredited programs. ;
(d) WNP-2 produced a record 3,315.340 megawatt-hours (gross) of electricity I during a record 133 days of continuous operation (July 26 - December 6),
at a capacity f actor of 91.3 percent. The previous record of a continu- :
ous operating period was 100 days (August 4 - November 13, 1985), when ,
1,918,970 megawatt-hours (gross) was produced. i 1-2 ,
1 1
In October, the plant's best monthly capacity f actor of 94.6 percent was achieved, with a record generation of 799,630 megawatt-hours (gross) of electricity.
(e) WNP-2 continued to have a positive trend in the reduction of Licensee Event Reports (LERs). A total of 33 LERs were written during 1987 as compared to 44 LERs in 1986. In addition, the NRC rated the overall average WNP-2 LER score at 8.8, as compared with the industry average of 8.4.
Also during the year, the NRC performed an in-depth Safety System Functional Inspection (SSFI) at WNP-2. The purpose of the inspection was to access the operational readiness of the AC and DC Electrical Distribution Systems, the Standby Service Water System, and the Automatic Depressurization System to function under all operational and analyzed accident conditions. At the con-clusion of the inspection, it was noted that no significant deficiencies were identified which would prevent the reviewed systems from performing their intended functions. However, it was concluded that improvemerts were needed in several areas. In the inspection report, the in::pectortAsidentified ten a result, the Notices of Violation (five at level IV and five at Level V).
Supply System has developed an action plan for strengthening the design modification process. The action plan addresses the broader programmatic issues which either were identified by the NRC SSFI team, or which resulted f rom Supply System consideration of underlying causes which may have contri-buted to the deficiencies noted by the NRC. The action plan includes both near and long-term initiatives to address the programmatic improvement needs as a result of the SSFI. In addition, we will continue to monitor our per-formance in this area and make improvements where necessary.
The 1987 actual and adjusted capacity factors, based upon net electrical energy output, are listed in the following table. For the period of t;me when the plant was operating with one recirculation pump, the adjusted capacity factor was based on a maximum power output of 71.7 percent rather than 100 percent.
Menth Capacity Factor Adjusted Capacity Factor January *** 69.6 97.1 February *** 55.0 76.6 March *** 63.4 88.3 April
- 20.8 29.0 May 0 0 June ** 1.8 1.8 July 15.3 15.3 August 89.2 89.2 September 93.1 93.1 October 94.6 94.6 November 91.9 91.9 December 79.8 79.8 Overall 56.2 63.0
- Started Maintenance / Refueling Outage
- Ended Maintenance / Refueling Outage
- Single Loop Operation 1-3
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0 FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC JAN l JAN 1987 MAXIMUM POWER OUTPUT LlhillTED TO APPROXIMATELY 72% BASED ON SINGLE LOOP OPERATION DATA BASED ON AVERAGE POWER GENERATED PER DAY. THEREFORE, RECOVERY FROM A SCRAM
.. w o f THAT OCCURRED WITHIN A 24 HOUR PERIOD WILL NOT INDICATE A ZERO PERCENT POWER LEVEL.
2.0 REPORTS The reports 'Jrovided in this section meet the requirements of Federal Regula-tions (10CFR50.59) and the WNP-2 Operating License. Complete data for the year 1987 has been included.
2-1
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2.2 MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES This section contains information concerning main steam line safety / relief valve challenges for calendar year 1987 i~ accordance with the requirements of NUREG 0737 Item II.K.3.3, and as required by WNP-2 Technical Specifications, Administrative Controls section, paragraph 6.9.1.5(b).
TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATION POWER ASSOCIATED DATE COMPONENT ID (CODE) (CODE) (CODE) LEVEL LER 03/22/87 MS-RV-5B B G E 0% 87-002 03/22/87 MS-RV-SC B G E 0% 87-002 03/22/87 MS-RV-30 8 G E 0% 87-002 03/22/87 MS-RV-40 8 G E 0% 87-002 The 03/22/87 actuations were in response to a plant trip.
n 04/10/87 MS-RV-1A B C 20% --
04/10/87 MS-RV-2A 8 0 C 20% --
04/10/87 MS-RV-3A B : C 20% --
04/10/87 MS-RV-4A B D C 20% --
04/10/87 MS-RV-18 8 0 C 20% --
04/10/87 MS-RV-28 8 0 C 20% --
04/10/87 MS-RV-3B B D C 20% --
04/10/87 MS-RV-48 8 0 C 20% --
1 04/10/87 MS-RV-58 8 0 C 20% --
04/10/87 MS-RV-1C 8 D C 20% --
04/10/87 MS-RV-2C 8 0 C 20% --
04/10/87 MS-RV-3C 8 0 C 20% --
04/10/87 MS-RV-4C B D C 20% --
04/10/87 MS-RV-5C B 0 C 20% --
04/10/87 MS-RV-10 8 3 C 20% --
04/10/87 MS-RV-20 8 0 C 20% --
04/10/87 MS-RV-30 8 0 C 20% --
04/10/87 MS-RV-40 8 0 C 20% --
The 04/10/87 actuations were in support of the acoustic monitoring system calibration procedure, Technical Specification requirement 3/4.4.2.
I 2-3 L
2,2 MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES (Continued)
TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATION POWER ASSOCIATED DATE COMPONENT 10 (CODE) (CODE) (CODE) LEVEL LER 06/18/87 MS-RV-3B B G C 0% --
06/18/87 MS-RV-38 8 G C 0% --
06/18/87 MS-RV-4A B G C 0% --
06/18/87 MS-RV-4A B G C 0% --
06/18/87 MS-RV-2A B G C 0% --
06/18/87 MS-RV-2A 8 G C 0% --
06/18/87 MS-RV-48 8 G C 0% --
06/18/87 MS-RV-48 8 G C 0% --
06/18/87 MS-RV-2C 8 G C 0% --
06/18/87 MS-RV-2C B G C 0% --
06/18/87 MS-RV-4C B G C 0% --
06/18/87 MS-RV-4C 8 G C 0% --
The 06/18/87 actuations verified S/RV operability from remote and alternate remote shutdown panels. This has since been incorporated into acoustic monitoring surveillances.
06/20/87 MS-RV-40 C + Hydroset C C 3% --
06/20/87 MS-RV-4A C + Hydroset C C 3% --
06/20/87 MS-RV-4A C + Hydroset C C 3% --
06/20/87 MS-RV-4A C + Hydroset C C 3% --
06/20/87 MS-RV-4A C + Hydroset C C 3% --
06/20/87 MS-RV-4A C + Hydroset C C 3% --
l 06/20/87 MS-RV-2A C + Hydroset C C 3% --
06/20/87 MS-RV-2A C + Hydroset C C 3% --
I 06/20/87 MS-RV-3A C + Hydroset C C 3% --
06/20/87 MS-RV-3A C + Hydroset C C 3% --
! 06/20/87 MS-RV-4A C + Hydroset C C 3% --
06/20/87 MS-RV-38 C + Hydroset C C 3% --
06/20/87 MS-RV-3B C + Hydroset C C 3% --
2-4 l
l 2.2 MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES (Continued)
TYPE OF PLANT REASON FOR REACTOR ACTUATION CONDITION ACTUATICN POWER ASSOCIATED (CODE) (CODE) (CODE) LEVEL LER DATE COMPONENT 10 06/20/87 MS-RV-5B C + Hydroset C C 3% --
06/20/87 MS-RV-5B C + Hydroset C C 3% --
06/20/87 MS-RV-40 C + Hydroset C C 3% --
The 06/20/87 actuations were part of the setpoint verification surveillance.
06/24/87 MS-RV-1A B C C 10% --
06/24/87 MS-RV-2A B C C 10%
06/24/87 MS-RV-4A B C C 10%
06/24/87 MS-RV-1B B C C 10%
06/24/87 MS-RV-38 8 C C 10%
06/24/87 MS-RV-10 B C C 10%
06/24/87 MS-RV-30 B C C 10%
06/24/87 MS-RV-4B B C C 10% --
06/24/87 MS-RV-5B B C C 10%
06/24/87 MS-RV-1C B C C 10%
06/24/87 MS-RV-4C B C C 10% --
06/24/87 MS-RV-50 B C C 10%
The 06/24/87 actuations were performed to reduce leakage through the valve seats.
07/02/87 MS-RV-18 A G A 0% 87-020 07/02/87 MS-RV-1B B G E 0% 87-020 07/02/87 MS-RV-1C A G A 0% 87-020 07/02/87 MS-RV-2C B G E 0% 87-020 07/02/87 MS-RV-3B A G A 0% 87-020 The 07/02/87 actuations were in response to a reactor trip.
2-5 1
2,2 MAIN STEAM LINE SAFETY / RELIEF VALVE CHAlt.ENGES (Continued)
TYPE OF PLANT REASON FOR REACTOR ACTUATION CON 0! TION ACTUATION POWER ASSOCIATED DATE COMPONENT 10 (CODE) (CODE) (CODE) LEVEL LER 11/26/87 MS-RV-2D A E D 97%
11/26/87 MS-RV-20 A E O 97%
The 11/26/87 actuations were caused by a plant technician incorrectly installing a set of jumpers.
12/11/87 MS-RV-20 8 E C 15%
The 12/11/87 actuation was initiated to verify a repair on the associated acoustic monitor.
2-6
2.2 MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES (Continued) ggEQ:
Tvoe of Actuation A. Automatic B. Remote Manual C. Spring Plant Condition A. Construction B. Startup or Power Ascension Tests in Progress C. Routine Startup D. Routine Shutdown E. Steady State Operation F. Load Changes During Routine Operation G. Shutdown (Hot or Cold)
H. Refueling Reason for Actuation A. Overpressure B. AOS or Other Safety System C. Test D. Inadvertent (Accidental / Spurious)
E. Manual Relief NOTES: 1) Remote manual actuations occurred in support of acoustic monitor position indication calibration testing required by Technical Specifications LCO 3/4.4.2.
- 2) Spring set testing was performed in accordance with ASME Sec-tion XI and Technical Specifications requirement in appli-cability paragraph 4.0.5.
2-7 l
2.3
SUMMARY
OF P(ANT OPERATION INCLUDING UNIT SHUIDOWNS/ POWER REDUCTIONS GtNERAIOR OUIAGE OFF-LINE CAU5t SHUIDOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 0 A S --
CB PUMPX Power output limited to 72% due to 11/10/86 F thru inoperability of the "B" Recirculation Pump.
4/10/87 2/20/81 F 122 A 1 HA 1URBIN The plant was shut down because of Icw turbine bearing oil pressure. An inspection revealed a broken check valve in the bearing oil supply header.
2/25/87 F 5.9 G 1 HA GENERA The turbine generator was removed f rom service due to high exciter tempera-ture which initiated the fire protec-tion system and admitted 002 inside the exciter housing. An inspection m revealed that the cooling water e
valving was misaligned. The exciter I oa was inspected for damage and the plant (
Was returned 10 service.
l l
2,3
SUMMARY
OF PLANT OPERATION INCLUDING UNIT SHUIDOWNS/ POWER REDUCTIONS (Continued)
GENERA 10R 00 TA6E OFF-LINE CAUSE SHulDOWN LER DAlf TYPE hours CODE METHOD NUM8ER SYSTEM COMPONENT CAUSE AND ACTION 10 P.* EVENT- RECURRENCE I 57.3 2 87-02 CH INSTRU 1he reactor was manually scrammed at 3/22/81 A 71% power due to a sudden reduction in RPV level caused by loss of both feed-water pumps on low suction pressure.
A failed fuse in the feedwater level l controller caused the event. Recovery efforts led to the flooding of the Main Steam Lines up to the closed MSIVs, due to the improper lineup of the RFW System startup level control valve. The plant was brought to cold shutdown. An engineering evaluation was performed to include extensive plant piping and support system m
inspections to verify no adverse e
effects.
e 4/10/87 5 1823 C 1 RC FUEL The plant was shut down as scheduled for the annual refueling outage.
thru 6/25/87 6/25/87 S 10.73 8 1 --
HA MECFUN 1he generator was removed from the grid to perform turbine overspeed testing.
I
2.3 $UMMARY OF PI ANI OPERATION INCLUDING UNIT SHUIDOWNS/ POWER REDUCTIONS (Continued)
GENERA 10R OUTAGE OFF-LINE CAUSE SHulDOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 6/26/87 F 17.63 A 3 87-18 E8 TRANSF The reactor automatically screamed due 6/27/87 F 26.32 A 3 87-18 E8 TRANSF to a Reactor Protection System (RPS) actuation from a Turbine Control Valve (ICV) fast closure. The TCV fast clo-sure was in response to a unit lockout signal, found later to be the result of a sudden pressure relay actuation on the Normal Auxiliary Power Trans-former, TR-NI. Following the 6/26 scram, the plant was restarted with TR-N1 out of service for further evaluation. The cause of the i 6/26 scram was thought to be a degraded
- TR-N1 transformer that had previously
"' sustained potential degrading tran-8 sients and showed signs of recent
! Es degradation from oil samples taken.
The same sequence of events occurred on 6/27, causing a similar reactor scram due to IR-N2.
The cause of both scrans was deter-l mined to be the opening of a (poppet) valve used to test the relay trip function. Following the scras on 6/27, the poppet valves were removed
,and plugged to prevent recurrence.
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2.3
SUMMARY
OF PLANT OPERATION INCLUDING UNIT SHUTDOUNS/ POWER REDUCTION 3 (Continued)
GENERA 10R DUTAGE OFF-LINE CAUSE SHUIDDWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 29.13 3 87-19 EB CKIBKR The reactor scrammed at $4% power due 6/28/87 F A to a spurious loss of Motor Generator (MG) power to RPS Bus A, coincident with an existing RPS 1/2-trip from an inoperable / tripped Governor Valve fast closure pressure switch. The cause of the RPS Channel "A" breaker tripping appeared to be the improper alignment of the undervoltage restraint coll.
The circuit breaker and oil pressure switch were replaced and tested prior to plant restart. Subsequent analysis by General Electric suggested another probable cause of coil failure was
' thermal aging.
U 88.03 3 81-20 E8 GENERA 1he reactor scrammed from 80% power 7/2/87 F A due to a loss of power to both RPS buses. Initially a Motor Generator (MG) Set failed, causing the loss of RPS "A" power. When plant operators attempted to switch RPS "A" power to its alternate source, per procedure.
RPS "B" power was lost. The deenergi-zation of both RPS buses causes, by design, a reactor scram. 1he faulty MG Set motor was replaced and tested.
An inspection of the transfer switch, common to both RPS divisions of power, revealed a broken stop tab which was repaired and tested.
.. _- - _ - - - .. . . = - ._
2.3
SUMMARY
OF PI ANT OPERATION VNCLUDING UNIT SHUIDOLMS/ POWER REDUCTIONS (Continued)
GENERATOR DUTAGE Off-LINE CAUSE SHUIDOWN LER DAIE TYPE HOURS CODE _ METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVENT RECURRENCE 7/6/87 F .27 A 9 --
HA RELAYX Two minutes af ter synchronizing, the generator tripped due to insufficient load pickup by the DEH System. After evaluation, a higher load reference and rate was used and the generator was resynchronized su:cessfully.
7/6/87 F 223.23 A 3 87-22 EB CKTBKR The r ictor scrammed f rom 33% power on low reactor pressure vessel (RPV) water level. During the transfer of plant electrical loads from the Startup Power Supply to the Normal Power Supply, an electrical breaker failed, resulting in a loss of the inservice Reactor feedwater (RFW) pump and subsequent loss of feedwater to the RPV. The R' cause of the breaker failure was a bent finger in the "C" Phase Dis-connecting Contact Finger Cluster, which prevented the breaker from fully inserting. 1he breaker was repaired and extensive testing was performed prior to plant restart.
i 2.3
SUMMARY
OF PLANT OPERATION INCLUDING UNIT SHUTDOWNS / POWER RfDUCTIONS (Continued)
GENERATOR OUTAGE OFF-LINE CAUSE SHUIDDWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PREVEN 1 RECURRENCE 7/15/87 5 267.47 8 9 -- -- -- Supply System management directed that the plant remain shut down pending a thorough review of the events sur-rounding the five previous scrams. A subsequent Confirmatory Action Letter concurred with Supply System manage-ment's decision and outlined steps to be taken prior to reactor startup.
Areas evaluated include, but were not limited to 1) an evaluation of the effectiveness of the WNP-2 post-trip review and root cause assessment pro-grams 2) a reevaluation of the root cause analysis of problems encountered N during the recent startup program.
- 3) an assessment of major work items accomplished during the 1987 refueling Z
outage, and 4) discussions with mem-bers of plant staf f regarding the j implementation of the rcot cause analysis program. At the conclusion of this intensive review, progranraatic changes and corrective actions were discussed with the NRC and found to be acceptable, resulting in plant restart.
9/17/87 5 0 11 5 -- RB CCNROD Reactor pcwer was reduced, as required, to perform a scheduled control rod sequence exchange. .
2.3
SUMMARY
OF PLANT OPERATI C INCLUDING OIT SHUTDOWNS / POWER REDUCTIONS (Continued)
~
GENERATOR 001 AGE OFF-LINE CAUSE SHU1DOWN LER DATE TYPE HOURS CODE METHOD NUMBER SYSTEM COMPONENT CAUSE AND ACTION TO PRE'/ENT RECURRENCE 11/9/87 5 0 H $ --
RB CONR00 Reactor power was reduced, as required, to perfonn a scheduled control rod sequence exchange.
12/6/87 S 103.8 B 1 -- -- --
The plant was shut down in support of various inspection and repair activi-ties necessary to support sustained rated power operation. Activities inciuded inspection of condensate filter demineraltiers (CFD), repair of CFD outlet valves, replacement of resin strainers, repair of reactor building ventilation supply fans and corraction of secondary side steam
- 1eaks.-
% 12/13/87 5 4.1 B 1 -- -- --
The generator was removed from service to perform tests to evaluate a mechani-cal binding condition of turbine gover-r.or valve #4.
101AL GENERAT03 0FF-LINE CAtlSE CODE TOTAL FOR 1987 HOURS A 9 563.9 8 4 ~386.1 C 1 1823.0 0 0 0.0 F 0 0.0 G 1 5.9 H 2 0.0 101AL 2778.9
2.3
SUMMARY
OF PiANT OPERA 110N INCLUDING UNIT SilVTDOWNS/ POWER REDUCTIONS (Continued)
SUMMARY
OF CODES OUTAGE TYPE F- Forced 1
5- Scheduled l
l CAUSE CODE l
' A- Equipment failure B- Maintenance or Test l C- Refueling D- Regulatory Restriction
' E- External Cause i G F- Administration l
l G- Personnel Error l
II - Other
$41UT00WN METii0D 1- Manual 2- Manual Scram 3- Auto Scram 4- Continued 5- Reduced Load 9- Diner
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2.3 SIWW4ARY OF PLANT OPfRATION INCLUDING UNIT SHU10065fS/ POWER REDUCTIONS (Continued)
SYSTEM CODE STANDARD CODE SYSTEM DESCRIPTION CA Reactor Vessels & Appurtenances 08 Coolant Recirculation Systems & Controls CF Residual Heat Removal Systems & Controls CH Feedwater Systems & Controls IA Reactor Trip Systems
" Offsite Power Systems & Controls
[A s ,
5 E8 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 R8 Reactivity Control Systems I
I
2.3
SUMMARY
OF PLANT OPERATION INCLUDING UNIT SHUT 00lNIS/ POWER REDUCTIONS (Continued)
COMPONENT CODE COMPONENT TYPE / CODE COMPONENT TYPE INCLUCES:
Circuit Closers / Interrupters Circuit Breakers (CKlBkK) Contactors Controllers Starters
, Swit-hes (other than sensors)
Switchgear Control Rod Drive Mechanism Control Nod Drive Mechanism (CONROD)
Instrumentation and Controls Controllers (INSTRU) Sensors / Detectors / Elements
~ Indicators Offferentials
[ Integrators (Totallzers)
~ Power Supplies Recorders Switches Transmitters Computation Modules 1 Penetrations Primary Containment Air Locks (PENE1R) Personnel Access fuel Handling Equipment Access Electrica Instrument Line Process Piping Pipes. Fittings Pipes (PIPEXX) Fittings Pumps Pumps (PUNPXX)
?.3 SINetARY OF Pt ANT OPERATION INCLUDING UNIT SHUT 00bMIS/ POWER RE00CTIONS (Continued)
COMPONENT CODE (continued)
COMPONENT TYPE / CODE COMPONENT TYPE INCLUDES:
Relays Switchgear (RELAYXX) 1ransformers Transformers (TRANSF) ,
Turbines Steam Turbines (TUR81N) Gas Turbines Hydro Turbines Valves Valves (VALVEX) Dampers .
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2.4 SIGNIFICAN! T1AINifNANCE PERf0RMED ON SAFETY-RELATFD EQUIPMENT EQUIPMLNT Rf0 HIRING MAINTENANCE SYSTEM PROBtEM ACTION TAKEN CRD-IK-125/2203 Control Rod Drive Prior inspections have Removed water accumulators and CRD-IK-125/2247 indicated signs of cor- replaced with new spares. Per- -
CRD-1K-125/4243 roston on the internal formed a leak test on all CHD-Ir.-125/1051 walls of these water connections.
SW-V-90 Service Water Valve motor runs, but Disassembled valve and found a the valve stem does not breken shaft assembly. Replaced move. broxen part:;, greased, tested and returned to service.
CRD-DRVE-5431 Control Rod Drive Control Rod Driv.e:. have Removed and replaced position CHD-DRVL-1803 no position indication indicator probes per procedure at notch "46". and verified position indication
~
at all notclies.
e 2 REA-FT-7 Reactor Building Elevated release stack Replaced M.e associated flow RE A -SQR T -1 Exhaust Air flow instrument failed transmitter and square rooter and te respond during the calibrated the instruments.
performance of a sur- Reperformed applicable steps of veillance test. surveillance proceoure and
- returned to service.
Dl 0 -I S-34 B1 Diesel tube 011 The treperature switch During troubleshooting activiiles.
l for DG *B" Engine 011 the temperature switch was found j cooler outlet appears to be darag.-d. The switch wo:.
I to be giving false replaced'and recalibrated per indication. plant procedures.
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2.4 SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY-RELATED EQUIPMENT (Continued)
[\ '
EQUl *ENT ACTION TAKEN RE0ti! RING'Hi!!NTENANCE SYSTEM PROBLEM MSLC-PS-20 ^ 'Aaln Steam Leakage The operational trending Removed instrument from service w ~
1 "
_ Control program indicates that and replaced the Jwitch with a x MSLC-PS-20 switch #2 is new spare. Ikcalibrated the ' ^ '
resetting erratically. instrume @ per riant procedure ~
and returncJ tv service. '
I EA-SR --38 Turbine Exh3u;t ~
Particulate and lodine Secured sample rack and replaced .' .
Air flow control is flow control valve #2 with a
^ '~
malfunctiontr.3 spare. Replaced the diaphragm on ' _
flow control valve #1. Perforned restoration of sample rack per x
- plant procedure.
i DLfJ-P-6 Diesel Lube Oil HPCS Diesel "Iccalating The pump was taken out of service
.Y~ Oil Pump cou' Isling failed. to replace a broken coupling. l
^ '
- _ Upon completion of the work, she g y pump was returntd to service, DSA-LOC-482/1 Diesel Starting A lef t-harided air lubri- Removed reservoir from replace-Air cator was installed by a ment lubricator and cleaned with_
vendor on the right- . solvent. Isolated the air header hand air s line (i.e.,~ air and removed old lubricator uq1t.
flow 1 6ackwards). Replaced the installed lubricator Replac'e lubricators with the correct type. -
with the new lubricator and returned to tervjcp. 1 ,
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x 4 %
00-LS-12A - Ts~* A 011' The low-level switch on Removed the electrical cover and i day tank #3 is shorted switch housiD9 and cleaned them to the housing cover. as much as possible. Verified
'- that the switch was still opera-f tional. Repaired jamaged wiring per plant procpdure, recalibrated e
, N . instrument and'Esturned to service.
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2.4 SIGNIFICANT MAIN 1FNANCE PERFORMED ON SAFETY-RElATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN MS-AO-228 Pain Steam General Electric Valve operators were removed from MS-AO-22C requires a five-year valves and moved to a suitable Main Steam Isolation work area. A sample of hydraulic Valve operator overhaul. fluid was taken and all elastomers were logged for wear analysis and potential EQ life extension.
Leakage tests were perfcrmed and foun1 to be acceptable. The operators were reassembled and installed on the valves.
MS-SPV-22Al, 2, 3 Main Steam The equipment graalifi- Isolated electrical and air sys-MS-SPV-22Bl. 2, 3 cation program requires tems and disconnected leads from MS-SPV-22Cl, 2, 3 periodic maintenance of each solenoid. Replaced coils and MS-SPV-22DI, 2, 3 the main steam isola- valve seals with environmentally MS-SPV-28Al, 2, 3 tion valve solenoid qualified spares. Reassembled the MS-SPV-28Bl. 2, 3 valves. Replace coils valves and retLrned to service.
MS-SPV-28C1, 2, 3 and elastomers.
S MS-SPV-28DI, 2, 3 RPS-PS-SD Reactor Protection lhe lurbine Generator Removed pressure switch from serv-System Valve fast closura pres- ice and replaced with a new spare.
sure switch was leaking Recalibrated the instrument and DEH oil. performed a channel check, returned the instrument to service.
LPRM-del-08/25 Low Power Range Two channels appear to During troubleshooting activities Monitor have erroneous output - APRM channel C was found to have investigate and repair a defective power supply in the as necessary. ion chamber. Replaced the trans-former and tested satisfactorily.
The other channel was tested and appears to be operating normally.
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2.4 SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY-RELATED EQUIPMENT (Continued)
EQUIPMENI SYSTEM PR_0BLEM ACTION TAKEN REQUIRING MAINTENANCE LPCS-42-7868 Low Pressure Core The motor starter for Troubleshooting activities found Spray the LPCS water leg pump that the motor starter had some appears to be broken. auxiliary contacts broken off.
The magnetic starter was replaced, all connections were checked for tightness and the operational test was successfully completed.
RCC-V-6 Reactor Closed The valve closes and A loose wire was found on a termi-Cooling Water cannot be electrically nal block in the motor control opened. center. A broken torque switch was also found on the valve. The wire was tightened and the +crque switch replaced and the valve tested satisfactorily.
m RPS-LPA-3C Reactor Protection The reactor scrammed due Removed EPA breaker 3C and i System to the opening of RPS inspected the undervoltage N MG Set breaker. restraint coil. Found slight armature bending. Adjusted the coil to relieve the bending and exercised the trip mechanism to e
verify proper operation. Also inspected and verified operation of breakcrs 3A, B. D, E and F.
LEA-SR-38 Turbine Exhaust lhe sample rack for The isokinetic flow valve was Air Turbine Building stack cleaned, adjusted. The asso-flow shows poor linear ciated flow switch, transducer flow characteristics, and square rooter were replaced Remove, inspect and during overhaul of the sample repair as necessary. rack.
J
2.4 SIGNIFICANT MAINTENANCE PERFORMED ON SAFETY-RELATED EQUIPMENT (Continued)
EQUIPMENT REQUIRING MAINTENANCE SYSTEM PROBLEM ACTION TAKEN CMS-TI-43R Containnient Suppression poo' water Repaired a splice for the tempera-Monitoring System temperature insivament ture element in terminal box on the remote shutdown TBC-501. A wire in the outboard panel failed high. penetration box was also found damaged and was respliced in accordance with plant procedures.
i l REA-E/S-613C Reactor Building The Reactor Building During troubleshooting, techni-Exhaust Air exhaust plenum radiation clan found no 24-volt supply to monitor has failed the instrument. One bad fuse was downscale. found and replaced. A functional test was performed and the instrument was returned to service.
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2.4 OTHER SIGNIFICANT MAINTENANCE ITEMS CR0 Removal / Replacement Twenty (20) control rod drive mechanisms were removed from the reactor vessel and replaced with rebuilt spares. The drives were selected for maintenance based upon trended performance characteristics such as stall flows, high temperature, difficulty notching out of "00" or other operational problems.
Turbine Maintenance During the 1987 refueling outage, the main turbine was disassembled for manu-f acturers' required inspections. Low pressure (LP) turbines #1 and #2 were completely disassembled and inspected for signs of deterioration. A disc crack indication was found on LP #1. Due to this failure, the Supply System shortened the frequency of inspection from 47 months to 31 operating months for this component. No significant problems were found on LP #2. Two throt-tle valves were partially disassembled / inspected, two reheat stop valves and two intercept valves were also inspected with no significant problems identi-fied. All governor valves and bypass valves were completely disassembled and refurbished. New wear rings were installed on the governor valves and the bypass valves needed various internal comoonents. Eight new Moisture Separa-tor Reheater (MSR) tube bundles were installed due to the presence of rust, tube failure and support plate bowing on the originally installed tubt, bundles.
Reactor Recirculation Pumo Modification A significant portion of 1986 and 1987 was spent in single loop operation due to reactor recirculation pump problems. During the 1987 refueling outage, both recirculation pumps were overhauled and modified. Modifications to pump 1A included strengthening of the overall bearing design as well as a reinforcement of the wear ring surfaces. Pump 18 had the bearing design modi-fication implemented during 1986; therefore, only reinforcement to the wear ring was necessary during 1987. The repairs implemented during this refueling outage have successfully solved this problem and the unit has been able to set new generation records in double loop operation.
Limitoraue Motor Operator Torque Switch 8voajijt During investigation into the failure of a Main Steam Leakage Control (MSLC) motor operator, it was noted that a required torque switch jumper was not installed. Documentation was found indicating that jumpers had been correctly installed on the valve during the startup phase of construction. An engi-neering analysis was performed and indicated the potential for other errors with regard to missing torque switch jumpers. Based upon this analysis, the decision was made to perform field inspection of all MOVs having an "open" safety function. Of the 66 MOVs inspected,14 were found missing the required jumpers. All required jumpers were installed and top tier drawings were verified to be correct or were modified as necessary, 2 - 24
1 2.4 OTHER SIGNIFICANT MAINTENANCE ITEMS (Continued)
Motor Control Center Configuration Control As a result of the Limitorque MOV deficiencies, a fuse and motor overload relay inspection was conducted to verify that the plant configuration complies with design requirements. Of the more than 700 circuits inspected, 83 dis-crepancies were identified. Sixty-two (62) circuits could not be adequately inspected with the related equipment energized. Twenty-eight (28) of the 83 discrepancies have subsequently been corrected, and 40 of the 62 circuits previously unavailable for inspection have been successfully inspected. The remaining inspections and corrections will be completed either during or before the next refueling outage. Also, a procedure controlling fuse replacement has been draf ted for implementation in 1988 to enhance our ability to control plant configuration relative to fuses.
Installation of Viton Seals During 1987, all ITT General Controls hydromotor polyurethane seals installed on WNP-2 valve actuators were replaced with Viton seals. This modification was made because, with the originally installed seals, the actuators could not be environmentally qualified for the six-months post LOCA as required by the FSAR. Forty-seven (47) class 1 actuators, ten nonsafety-related and all spares located in the warehouse were upgraded. This modification will extend the service life on each of the actuators an average of eight years.
Safetv-Related MOV Inspection and Voorade Due to a number of equipment qualification group concerns, an inspection of safety-related motor operate 1 valves was performed. Items of interest during the inspection included: Oroperly insulated splices, installation of correct terminal blocks, valve rrientation relative to drain plugs, integrity of torque switches and correct use of space heaters. Various deficiencies were corrected and a large number of cables were respliced using Raychcm shrinkable materials. The addition of this insulation has resulted in a greatly increased effective service life of these valves.
Three-Year Diesel Engine Maintenance The Supply System performed the manuf acturers' recomended one-year and three-year preventive maintenance to all five of the WNP-2 diesel engines. This maintenance included, but was not limited to, the inspection of the turbo-charger and exhaust piping, replacement of air-start motors, rebuilding or replacement of starting air system solenoid valves, inspection of selected main and connecting rod bearings, replacement of all filter elements in the lube oil, fuel oil and intake air systems and engine one-revolution inspec-tion. All equipment was found to be f ree of excessive wear or degradation.
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2,4 OTHER SIGNIFICANT MAINTENANCE ITEMS (Continued)
Imolementation of MOVATS Program During 1987, the Supply System responded to IE Bulletin 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Set-tings," by implementing the MOVATS Program.
Site Engineering personnel determined which valves would be tested in accor-dance with 10CFR50.55a(g). A program was then developed to ensure that the selected valve operators were tested and subsequently properly maintained. As a result of testing, several deficiencies were noted, including torque switches r.ot bypassed in the opening direction and torque switches set too high. Other items inspected were overall valve integrity and correct Ilmit switch settings.
In addition to meeting the IE Bulletin requirements, the Supply System has decided to expand MOVATS testing to other safety-related valve operators.
All deficiencies were corrected and an engineering evaluation was performed to ensure that all valves were not overstressed due to high torque switch set-tings. All valves have been approved for use until the 1988 refueling outage and further determination for long-term action is anticipated at that time.
9 2 - 26
2.5 INDICATIONS OF FAILE0 FUEL In accordance with the WNP-2 FSAR, Section 4.2.4.3, a visual inspection of discharged fuel assemblies from Cycle 2 was performed. This information is supplied in accordance with requirements as set forth in Regulatory Guide 1.16.
Selection of Assemblies Eight assemblies were selected for inspection representing greater than 5 per-cent of the discharged fuel. The selected assemblies are all medium enriched (1.76 w/o U-235). Some characteristics of the selected assemblies are shown below.
CYCLE 2 DISCHARGED FUEL ASSEMBLIES SELECTED FOR EXAMINATION Haling Cycle 2 POWERPLEX Radial Core EOC 2 Fuel Assembly Power location Exposure Bundle Tvoe Assembly Group 1 LJT 157 1.108 37-40 14,266 GE 1.76 LJT 159 23-22 14,265 GE 1.76 LJT 163 37-22 14,266 GE 1.76 LJT 164 23-40 14,266 GE 1.76 Assembly Group 2 LJT 259 .968 21-10 14,338 GE 1.76 LJT 321 21-52 14,338 GE 1.76 LJT 332 39-52 14.339 GE 1.76 LJT 357 39-10 14.338 GE 1.76 Inspection Technique The poolside visual examination was performed with an underwater periscope system and the results of the fuel inspection recorded with a 35mm camera. In general, two sides of each fuel assembly were viewed. Thirty-five mm (35mm) photographs were taken of the points of interest. A total of 52 photographs of the examined fuel were taken. Of these, 45 were successful photographs.
The inspection proce'ere involved moving the selected fuel assembly in a vertical position past the fixed periscope. This was accomplished by raising the fuel assembly out of the spent fuel rack by means of the fuel-handling mast on the refuel bridge.
2 - 27
2.5 INDICATIONS OF FAILED FUEL (Continued)
Inspection Criteria Visual inspection of the selected assemblies was performed to determine the extent of the following phenomena:
o Proper rod seating in the laver tie plate, o Rod bow and spacing, o Spacer location and perpendicularity, o Relative rod growth, o Condition of tie rod hex .'uts and other structural components, o Nodular corrosion and crud formation, and o Rod fretting.
Results of the Examination The inspected fuel assemblies appeared to have good structural integrity. The upper tie plates were level, rod springs had ample compression space, tie rod nuts were snug, there was no apparent rod bow and all rods were properly seated in the lower tie plate. The spacers were perpendicular to the rods and properly located. Major damage to one set of finger springs was noted.
Small scratches attributed to dechanneling were observed on some peripheral rods and on spacer surfaces. There was no evidence of deb-is damage. There was also no evidence of fretting behavior. In general, the mechanical integ-rity of the fuel appears to be good.
The rate of crud and nodular buildup on the fuel appears to have increased f rom that observed af ter Cycle 1. Some of the nodules seem to have started to come together to exhibit a spallation type of appearance. On one bundle, the crud layer, which was quite heavy, was cleaned from one fuel rod by mechanical cleaning. Oxide nodules were observed on this fuel rod where the crud had been.
The grid spacers exhibit a very heavy nodular buildup. In addition, several of the spacers exhibit holes in the heat-af fected zone of the welds. These holes are not an isolated phenomena. One spacer had four holes on one side and three on another. There is some disagreement at present as to the origin of these holes, with some reviewers believing they are a corrosion-based phenomena and others believing they were caused at manufacture.
Surma ry The inspected fuel appears to be free of significant mechanical damage caused by plant operation. Mechanical damage appears to be limited to minor scratches except for damage to one set of finger springs which was felt to have occurred during handling af ter fuel discharge.
2 - 28
2,5 INDICATIONS OF FAILED FUEL (Continued)
Heavy nodular corrosion exists on virtually all spacers. Significant nodular corrosion was observed on many of the fuel rods and is present under the crud layer. Some fuel regions appear to be exhibiting a spallation-type growth of nodules. The overall growth of nodule deposit appears to be at least keeping pace with the fuel exposure.
Based on the observed data, the risk of fuel failure from nodular corrosion appears to be low for Cycle 3. However, the current inspection techniques do not allow for examination of the Gd 023 fuel rods which are usually the fuel rods at most risk to this type of fuel failure.
1
( 2 - 29
i l
2.6 PLANT MODIFICATIONS 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 Regulatory Commission (NRC) approval, unless the proposed change, test or experiment involves a change in the Technical Specifications incorporated in the license or an unreviewed safety question. In accordance with 10CFR50.59, summaries of the permanent design changes and temporary plant modifications completed in 1987 are provided. Included are sumaries of the safety evaluations.
2 - 30
L 2.6.1 PLANT DESIGN CHANGES The following plant design changes were completed in 1987 and reported in accordance with 10CFR50.59. These modifications were evaluated 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 previ-ously evaluated in the WNP-2 updated Final Safety Analysis Report (FSAR),
I (b) create the possibility of an accident or malfunction of a different type than previously evaluated in the FSAR, (c) reduce the margin of safety as defined in the basis for any WNP-2 Technical Specifications, or (d) require a change to the WNP-2 Technical Specifications.
PLANT DESIGN CHANGE 83-0047 Plant Design Change 83-0047 was initiated to accorrrnodate surface drainage during severe conditions, such as rainf all and rapid snowmelts, by install-ing a system of catch basins. This modification also included the final site paving and grading plan.
This plant design change provided direction for the installation of a system of catch basins t. provide adequate drainage during conditions of l severe water or snow accumulation. This system replaces the previously I used system of ditches anc culverts and completes the deferred construction item of final site paving and grading.
This modification did not result in a change to the WNP-2 Technical Speci-fications or involve an unreviewed safety question because the modification actually precludes the potential flooding of safety-related structures as evaluated in FSAR Section 3.4.1.4.1.1.
PLANT DESIGN CHANGE 84-1071 Plant Design Change 81-1071 was initiated to provide a dedicated back-up water supply for the Fire Protection System.
This modification installed a 400,000-gallon bladder tank which serves as the secondary source of water for the Fire Protection System. The system previously used was not a dedicated source and also had a smaller storage capacity.
This modification did not result in a change to the WNP-2 Technical Specifications or involve an unreviewed safety question because the margin of safety is increased by the additional water supply for fire protection.
2 - 31
2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 84-1144 Plant Design Change 84-1144 was initiated to design and install a decon-tamination facility within the Radwaste Building of the plant.
This modification package included design for modifications of support equipment such as the HVAC system, floor drains, chemical waste and electrical systems, as well as the placement of essential decontamination equipment within the decon room.
This modification did not result in a change to WNP-2 Technical Specifi-cations or involve an unreviewed safety question because the equipment being added is not safety-related and because structural, HVAC and electrical loads associated with this modification are within the design rating of equipment and plant structures.
PLANT DESIGN CHANGES 84-1249 AND 84-1322 Plant Design Changes 84-1249 and 84-1322 were initiated to remove the gase-ous chlorination portion of the plant make-up water treatment system.
These plant design changes removed and/or disconnected piping and valves associated with the operation of the gaseous chlorination system. This system was removed to increase personnel safety by eliminating any possi-bility of an accident involving gaseous chlorine.
This modification did not result in a change to the WNP-2 Technical Speci-fications. This change did not result in an unreviewed safety question because the removal of gaseous chlorine f rom the site increases overall personnel safety and does not decrease the capabilities of the plant make-up water treatment system.
PLANT DESIGN CHANGE 84-1570 Plant Design Change 84-1570 was initiated to procure and install a single-package heat pump for the sodium hypochlorite tank storage area.
This modification installed a roof-mounted heat pump unit to serve the sodium hypochlorite tank storage area. All duct, hanger supports and accessories were also installed via this design change package.
This modification did not result in a change to the WNP-2 Technical Speci-fications or involve an unreviewed safety question because the addition of this equipment cannot cause an unanalyzed accident.
2 - 32
2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 84-1623 Plant Design Change 84-1623 was initiated to install a higher capacity jockey pump for the Fire Protection System. This modification also added a still water casing to the pump to assure action of the water from adjacent pumps will not cause damage to the jockey pump.
This modification was initiated because the originally installed pump had two major failures and was determined to be inadequate.
This modification did not result in a change to WNP-2 Technical Specifi-cations or involve an unreviewed safety question because the upgraded equipment only increases the reliability of the system; no functional system changes have been made.
PLANT DESIGN CHANGE 85-0050 Plant Design Change 85-0050 was initiated to replace the originally installed Plant Process Computer with a state-of-the-art computer system to increase reliability and system capabilities.
The originally installed plant process computer was unreliable and dif fi-cult to maintain due to the unavailability of spare parts. This design change replaced the original process computer with a state-of-the-art system designed to increase system capabilities and overall reliability.
This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question because the overall system function has not been altered as a result of this modification.
PLANT DESIGN CHANGE 85-0080 Plant Design Change 85-0080 was initiated to provide a replacement micro-computer for the Rod Worth Minimizer System.
The plant process computer was replaced with an updated computer system that did not include the software for the Rod Worth Minimizer System. This design change package installed a microcomputer to interf ace with the Rod Position Instrumentation System for display and control of operator rod movements when less than the low power setpoint.
This design change did not involve a change to WNP-2 Technical Specifi-cations or involve an unreviewed safety question because replacement system design is capable of meeting the existing Technical Specification LCO and surveillance testing requirements.
2 - 33
2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0155 Plant Design Change 85-0155 was initiated because control air lines on originally installed dampers for the Turbine Building Exhaust Air System were freezing during winter months.
This design change replaced air-operated dampers with automatic backdraf t dampers. This modification was designed for more reliable operation by installing mechanically operated dampers rather than relying on a sensing /
control system.
This modification did not involve a change to WNP-2 Technical Specifi-cations or involve an unreviewed safety question because there is no change to the function or intent of the system design.
PLANT DESIGN CHANGE 85-0273 Plant Design Change 85-0273 was initiated because the Main Steam Line pressure transmitting detectors are temperature sensitive and were influ-enced by ambient condition changes in the Turbine Building.
This plant modification relocated the Main Steam Line pressure transmitting detectors to an area within the Turbine Building where ambient conditions are more stable. Following the instrument relocation, leak tests, vacuum tests and visual inspections were performed in accordance with the ASME Section XI plan.
This modification did not result in a change to the WNP-2 Technical Speci-fications or involve an unreviewed safety question because the function of the system has not changed and does not create the possibility of an unanalyzed event.
l 2 - 34 l
O 2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0330 Plant Design Change 85-0330 was initiated in response to deficiencies found during the Supply System's review of the WNP-2 Appendix "R" Safe Shutdown analysis.
The majority of work associated with this design change involved the installation of fire remote transfer switches on equipment required for safe shutdown of the plant. This change ensures control power is available from either the Remote Shutdown Panel or the Alternate Remote Shutdown Panel in the event of a fire in the main control room. The other major area of work involved covering cable tray and individual cables with thermolag insulation.
This modification did not involve a change to WNP-2 Technical Specifica-tions or involve an unreviewed safety question because the additional pro-tection reduces the consequences of a design bases fire in the control room, thereby maintaining the plant original design bases and compliance to the intent of Appendix R.
PLANT DESIGN CHANGE 85-0519 Plant Design Change 85-0519 was initiated to provide a method for reducing erosion in the High Pressure Turbine cross-under lines.
This modification installed moisture preseparator units to the high pres-sure turbine extraction lines. Storage tanks with automatic level controls that discharge to the feedwater heaters were also added. Various electri-cal and instrumentation support was required to provide automatic control of these drain tanks.
This modification did not result in a change to the WNP-2 Technical Specifications or involve an unreviewed safety question because the design change increases the reliability of the piping and because this equipment is not considered to be safety-related.
2 - 35
?.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 85-0676 Plant Design Change 85-0676 was initiated as a result of ongoing effects by the Supply System to mitigate ef fects of Intergranular Stress Corrosion Cracking (IGSCC) and in consideration of NUREG-0313 Rev.1.
This modification replaced piping f rom the drain lines on loops A and 8 of the Reactor Recirculation Pumps with material that is not susceptible to IGSCC. This design change also removed four Pipe Whip Restraints (PWS) af ter an engineering analysis and walkdowns were performed verifying that no safety-related targets were within range of the affected piping.
This modification did not result in a change to the WNP-2 Technical Specifications or result in an unreviewed safety condition because pipe breaks at the locations of the removed pipe whip restraints have no safety-related ta rgets . The other portion of this design change is a material upgrade that will enhance system reliability that has no safety concerns associated with it.
PLANT DESIGN CHANGE 86-0005 Plant Design Change 86-0005 was initiated to satisfy Regulatory Guide 1.97 requirements relative to the environmental qualifications of the wetwell level monitors.
This modification removed originally installed wetwell wide-range level monitors and replaced them with environmentally qualified instruments. The existing level recorders, as well as most interconnecting cable, were used with no modifications required.
This modification did not involve a change to the WNP-2 Technical Specifi-cations or involve an unreviewed safety question because the upgrade of the equipment increases system reliability, thereby potentially increasing the margin of safety.
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2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 86-0052 Plant Design Change 86-0052 was written in response to a General Electric recommendation to inspect certain sample probes installed at WNP-2 for signs of crevice corrosion.
This modification replaced an originally installed sample probe, RFW-SP-3, with a shorter type because when the inspection for crevice cracking was performed, it was found that the sample probe had broken of f. No evidence of crevice corrosion was found on the inspected weld area.
This modification did not involve a change to the WNP-2 Technical Specifi-cations or an unreviewed safety question because the function of the sample probe is unchanged. The stress corrosion resistance has been improved, thereby increasing overall component performance.
PLANT DESIGN CHANGE 86-0324 Plant Design Change 86-0324 was initiated because the Standby Service Water System was experiencing water hammer problems during svstem startup.
This modification added a keep-full pump with associated piping and con-trols to maintain the Standby Service Water System full under normal oper-ating conditions. This design package also replaced originally installed system isolation valves with new valves which have better throttling and sealing characteristics.
The changes that were made reduce the potential for hydraulic transients in the piping, thereby reducing the probability of an accident associated with the system. No modification to WNP-2 Technical Specifications was made 6:
a result of this design change.
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2.6.1 PLANT DESIGN CHANGES (Continued)
PLANT DESIGN CHANGE 86-0569 Plant Design Change 86-0569 was initiated to address the problem of under-sized circulating water pump feeders in underground duct banks.
This plant design change installed four 500-MCM cables per phase in the underground section of the circulating water pump feeder to lower all cables to within their rated temperatures. These motors previously had two 500-MCM cables per phase and experienced a f ailure on the feeder cable to pump 10.
This modification did not result in a change to the WNP-2 Technical Specifications. This design change increases the conductor size for some cables in underground duct banks, thus providing an additional margin of safety by lowering the normal operating temperatures within the duct banks.
As a result of this modification, an engineering analysis was performed to review the cable ampacity criteria used at WNP-2.
i 2 - 38
l 2.6.2 LIFTED LEADS AND JUMPERS The following are summaries of noteworthy modifications made to the plant by the use of lifted leads and jumpers during 1987. Each modification was evalu-ated and determined not to represent an unreviewed safety question or require a change to the WNP-2 Technical Specifications.
Reactor Recirculation Mechanical Jumper A mechanical jumper was installed between the extend and retract lines for the recirculation valve flow actuator. This jumper was installed to circulate hydraulic fluid to facilitate flushing and leak testing following refurbish-ment of the actuator. Upon completion of flushing and testing, the actuator was reinstalled on the flow control valve in its original configuration.
This modification did not involve an unreviewed safety question or reduce any margin of safety as defined in WNP-2 Technical Specifications because the plant was in Mode 5 and the recirculation system was not required to be in service for the time the jumper was installed.
Bypass of Low Level Trio Switch on FOR-P-21 The low level trip relay for Radioactive Floor Drains Pump 21 (FOR-P-21) was bypassed to allow a complete blowdown of a hold tank to the river. The con-tents of the hold tank had been processed through the condensate demineral-izers, but had a high total organic carbon (TOC) level, which is not suitable for high-purity reactor water. Prior to releasing water to the river, an analysis was performed to ensure the tank content was within limits stated in the WNP-2 NPDES permit and 10CFR20, Appendix B.
The installation of this jumper did not involve an unreviewed safety question or reduce any margin of safety as defined in WNP-2 Technical Specifications.
The bypassed switch function is to protect the pump from loss of suction due to a tank low level condition. The function of this switch is not related to reactor safety.
BvDass of HVAC Roll Filters Auto Advance Function The auto advance function of the intake air filters for the Control Room, Rad-waste, Turbine and Reactor Buildings was bypassed due to the variations in air flow through the buildings being greater than anticipated. The wide varia-tions in air flow caused the dP switches to actuate the auto filter advance function unnecessarily, thereby wasting roll filters. A design change is in process to replace the dP switches with automatic timers. Until the design change is implemented, the Operations staf f will continue to manually advance the filters on a routine basis.
The installation of these jumpers did not involve an unreviewed safety ques-tion or reduce any margin of safety as defined in the WNP-2 Technical Specifi-cations because the overall function of the system has not changed f rom the original desigri and adequate compensatory measures were implemented. Periodic area temperature monitoring also verifies adequate area cooling.
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2.6.2 LIFTED LEADS AND JUMPERS (Continued)
Lifted Lead on Turbine Governor Valve #4 Vibration levels on a governor valve (GV) actuator has resulted in adminis-trative limits being placed on the percent open at which GV #4 is allowed to operate. The limit was established f rom an empirical vibration evaluation.
During normal operation, the GV is placed in the DEH "test" mode to limit valve position. Other governor valve positions and overspeed functions are l
not affected by this method of operation. During the monthly bypass valve testing, required by Technical Specifications, a lifted lead is relied upon to close the valve while the "test" mode of DEH is utilized to cycle all of the turbine control valves. The lead is lifted only in support of this testing.
Maintenance to alleviate the GV vibration problem is scheduled for the 1988 Refueling Outage.
2 - 40 i
2.6.3 FSAR AMEN 0 MENT EVALUATIONS The following are summaries of changes made to the FSAR which were not initi-ated as a result of a plant modification. Prior to submitting an FSAR change, an analysis is performed in accordance with 10CFR50.59 to ensure the proposed modification does not involve an unreviewed safety question. The following summaries represent changes in system operation, clarification and/or updates of system descriptions, clarification of Supply System positions and, in some cases, changes to connitments previously made in the FSAR.
Appendix F. Control of Combustibles Modification - This revision to the FSAR defines all types of compressed gases stored onsite and eliminates the location of cylinder storage. All compressed gases are stored in accordance with Supply System requirements and comply with requirements as set forth in NFPA 5 and NFPA 50A.
Basis for Change - Due to an NRC inspection, modification to the Supply System program for storage of noncombustible compressed gases was made to include physical restraints at the top and bottom of gas bottles. Storage of ill bot-ties was either modified to comply with the requirements or verified to be in compliance with the program. As a result of this ef fort, the FSAR was modi-fied to reflect a list of all compressed gases stored onsite and to define the methods of storage.
Chapter 13. Conduct of Operations. ComDosition of Fire Brigade Modification - The following statement was changed, "The Fire Brigade shall consist of the following personnel:" to "The Fire Brigade shall normally con-sist of the following personnel; however, any combination of qualified person-nel meeting 10CFR50 Appendix R requirements is acceptable."
Basis for Chance - This modification was made to allow greater flexibility in utilizing plant personnel for the composition of the Fire Brigade. A require-ment for being a Fire Brigade Leader is to have successfully completed a j training course specifically developed for that position. The limitation in the previous FSAR section, that the Fire Brigade Leader be a Shif t Support
! Supervisor, was unnecessarily restrictive.
2 - 41
2.6.3 FSAR AMEN 0 MENT EVALUATIONS (Continued)
Appendix F. Safe Shutdown Eauipment Modification - Removes the commitment that during normal plant operation, power is removed from RHR-V-8 (suction line isolation valve).
Basis for Chance - To correct an error made in a previous FSAR amendment r This section discusses the Supply System's position on resolving the high/ low pressure interface system issue. This issue is currently being addressed with the NRC and an alternate solution has been proposed by the Supply System. To date, the Supply System has not de-energized RHR-V-8 or RHR-V-9 during normal operation Chapter 5. Residual Heat Removal (Shutdown Coolina Mode)
- Modification - The paragraph describing the flow path for prewarming RHR Loop A was changed to delete RHR-V-71A from the description and to add RHR-V-70. The basic change is that rather than releasing water to radwaste from a point at pump suction (RHR-V-71 A) and a point downstream of pump dis-charge (RHR-V-72A and RHR-V-70), only the downstream drain will be used.
Basis for Chance - This modification does not reduce the amount of piping prewarmed.
Chapter 7. Enaineered Safety Feature Systems Modification - A new paragraph was written to describe the actions that must be taken to initiate the Standby Gas Treatment (SGT) System. The FSAR section deals with manual actuation capabilities of ESF Systems and SGT was not fully addressed. The clarification states that in addition to actuating the system start control switch, a control switch for a suction valve must also be actu-ated. This change does not involve a new method of system operation or a plant modification; it simply clarifies what actions must be taken to manually initiate SGT.
Basis for Chance - To clarify what actions must be taken to manually initiate the Standby Gas Treatment System.
1 2 - 42
m 2.6.4 OTHER Included in the Plant Nonconformance Reporting (NCR) process at WNP-2 is the requirement to perform a 10CFR50.59 Evaluation for those NCRs which are dis-positioned as "Use-As-Is," "Repair," or "Conditional Release." The specific purpose of the 10CFR50.59 Evaluation is to recognize these categories of NCRs .
as implementing a change to the facility, thus requiring a 10CFR50.59 Evalua-tion. When a 10CFR50.59 Safety Evaluation is performed, the NCR is reviewed by the Plant Operating Committee and approved by the Plant Manager prior to the equipment being declared operable.
The following is a discussion of plant changes which were made by means of the NCR process during 1987:
NCR 287-198 (RHR Flow Control Valve Stroke Times) o Problem Description The two-year Valve Position Indication (VPI) procedures for RHR-FCV-64A, B and C were revised by means of a procedure deviation to require the valves to be operated until they were full open. Based on stam travel, the valves must be open at least 90 percent to meet the two-year VPI requirements. However, to meet this requirement, tne oper. limit switch required adjustment to allow the valve to stroke open further to meet the requirements of the two-year VPI. As a result, this caused an increase in the stroke time.
o Corrective Action The NCR immediate disposition ("Use-As-Is") was to adjust the open limit switch to meet the two-year VPI requirements, measure the stroke time and revise the procedures that perform the ASME stroke time testing require-ments. It should be noted that valve stroke time is not required of Technical Specification Table 3.6.3-1 because the valves do not perform a containment isolation function.
An engineering evaluation was performed and it was determined that 1) the valves travel at the same rate and are within the guidelines (four inches per minute), 2) the valves will be open to the same position at the same time as measured during Plant Startup Testing, and 3) the further opening will have a negligible effect because flow / pressure is diminished by three restriction orifices downstream.
The 10CFR50.59 Safety Evaluation and concluded that the change does not decrease the capabilities of the RHR System or af fect the minimum flow protective function.
An FSAR Change Request was prepared to revise FSAR Table 6.2-16 to agree with the Technical Specifications.
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2.6.4 OTHER (Continued)
NCR 287-340 (Reactor Recirculation-RRC-Valve Hydraulic Control Unit) o Problem Oescription Due to unacceptable oscillation in reactor recirculation flow caused by flow control valve RRC-V-60A position feedback signal problems, the hydraulic unit for the valve was shut down, rendering the valve station-ary, thus preventing the recirculation loop flow runback function given a Reactor Feedwater (RFW) pump trip and Level-4 actuation, o Corrective Action Annunciator Response Procedure 4.603.A8-3.7 was revised by means of a procedure deviation to account for the locked-in-position condition of RRC-V-60A. The deviation provided direction to transfer both Reactor Recirculation Pumps to slow speed (15 Hz) in the event of a loss of an RFW pump and the RRC flow runback in the operable loop. The deviation presented compensatory measures evaluated as acceptable by the Supply System and General Electric.
A 10CFR50.59 Safety Evaluation was performed and concluded that the operational condition of RRC-V-60A locked in position, preventing an FCV runbeck in the event of an RFW pump trip and Level-4 .ondition, is bounded by the t' SAP. (Chapter 15) analysis and does not represent an Unreviewed Safety Question (USQ). The directbn provided by the proce-dure deviation to trip both RRC pumps to 15 Hz operation compensates for the runback function in an attempt to prevent a scram. The direction merely replaces a plant reliability-based function with suitable operator actions. The flow control runback function does not perform a safety function.
NCR 298-299 (Division 1, 24-Volt Battery Replacement) 7 o Problem Description During a f.ield walkdown with a vendor representative, it was noted that
' the Division I, 24-Volt battery was different from the Division II battery. Further investigation revealed that the Division I battery had been replaced prior to Startup ano had not been seismically qualified prior to installation.
o Corrective Action An Engineering analysis was performed to determine required corrective actions with the following results:
- Based upon vendor-supplied information, it was determined that the battery would meet or exceed all capabilities of the originally installed battery.
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2.6.4 .0THER (Continued)
- Based upon Sandia Lab seismic test results, it was determined that due to inherent ruggedness, the installed, unqualified battery was acceptable for interim use.
- Concurrent with the engineering evaluation, a 10CFR50.59 Safety Evaluation was performed and concluded that 1) the battery meets or exceeds the caoabilities of the originally installed battery and, therefore, is justified for continued use, 2) the differences between this battery and the originally qualified b3ttery are not i significant and do not affect performance, and 3) to date, the battery has performed and been tested successfully to the surveillance requirements listed in the WNP-2 Technical Specifications.
- This battery will be replaced with a seismically qualified battery durirg the 1988 refueling outage.
NCR 287-794 (Missing Hangers on Diesel Starting Air Drain Lines) o Problem Description During review of a Design Change Packsge (DCP), it was noticed that the work package documentation was incomplete. A field walkdown revealed that two hangers for the air receiver drain lines had not been installed, although the work package showed the work as being complete.
o Corrective Aqtj,on A 10CFR50.59 Safety Evaluation was performed and concluded that the pro-jected stress levelt were not suf ficient to cause a failure of the drain line; therefore, there was not a reduction in the margin of safety as previously evaluated.
Results of an Engineering Analysis allowed interim operation until the hangers could be installed. The analysis showed that the piping stresses would exceed the ASME Code requirements, but would not exceed the yield stresses of the pipe material. Therefore, although Code allowables for seismic loading were exceeded, no actual failure of the line would occur had a seismic event occurred. The hangers were f abricated and installed, restoring the system to its intended design configuration.
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1 2.6.4 OTHER (Continued)
NCR 287-352 and 287-353 (Exceeded Thrust Loads on Safety-Related MOVs) o Problem Description As a result of MOVATS testing, the as-found thrust load on a number of valve operators exceeded their rated loads. The affected valves are HPCS-V-4, HPCS-V-12, RCIC-V-10, RCIC-V-13, RCIC-V-45 and RCIC-V-59. The motor operator manufacturer allows a nominal 10 percent overload with a maximum one-time overload of 250 percent. The MOVATS dau taken during testing indicated the measured thrust values exceeded the motor operator design thrust limits by 20 percent to 80 percent.
4 o Corrective Action A review of the as-found and' asdi ef t thrust loadings has been initiated with a consultant knowledgeable in the analysis of motor operators.
Based on this initial review and evaluation of the valve operating history, the operators were approved for continued use until the 1988 refueling outage. Long-term corrective action is still being evaluated. Further analysis is being performed to formulate specific reconsnendations for each valve operator. }
A 10CFR50.59 Safety Evaluation was performed and concluded that 1) motor operator loads expected between now and the 1988 refueling outage and the past operating history loads are within the predicted f atigue limits for these operators; therefore, the operators will perform as previously evaluated in the FSAR, 2) no new failure mechanism has been introduced and the motor operators will perform their safety function per their original design, and 3) the fatigue usage remains within the design basis acceptance Yimits for these operators and the margin of safety for. these operators has not been reduced, i
e 2 - 46
k 2.7 PLANT TESTS AND EXPERIMENTS
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Federal Regulations (10CFR50.59) and the Facility Operating License (NPF-21) allow tests or experiments to be conducted which are not cescribed in the
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Safety Analysis Report without prior Nuclear Regulatory Commission (NRC) approval unless the proposed change, test or experiment involves a change in the Technical Specifications incorporated in the license or an unreviewed '
safety question.
Prior to performing any test or experiment, a safety evaluation was performed in accordance with 10CFR50.59. All such evaluations were reviewed and approved by the Plant Operations Committee prior to the performance of the tests. It was concluded from the reviews that the tests performed in 1987 did not (1) place the unit in an unanalyzed configuration or condition not bounded by design basis, or (2) perform an operation not described in the FSAR which could have an adverse affect on safety-related equipment or systems. The fol-lowing are summaries of tests performed in a mode of operation not described in the FSAR. It should be noted, however, that the abnormal mode of operation did not place the unit in an unanalyzed condition.
' PPM 8.3.64 Standby ticuid Control ATWS Preoperational Test A plant modification was made to comply with requirements set forth in 10CFR50.62(c)(4). This test was performed to ensure that the Standby Liquid Control (SLC) system meets the niinimum flow requirement of 86 gpm with an equivalent boron concentration of 13 percent. It also verified system operability following the relocation of the injection pat 5 and control circuit modifications. All acceptance criteria was met and verified to comply with the above-stated requirements. s i
PPM 9.5.4 Fuel Channel Measurement Acceptance 'Jst This procedure was prep 3 red to implement all ASEA-ATOM acceptance criteria and
tests, as well as all additional testing proposed by Supply System personnel on the new fuel char.nel measurement device purchased from ASEA-ATOM.
' Channel def ormation occurs due to neutron exposure and could potentially deform in such a manner to constrict the free movement of the control rods.
Deformation could also impact the LPRM and result in apparent flux tilts and erroneous flux profile indications. With the fuel channel measurement device, this deformation can be , measured and acceptance criteria established to elimi-nate the grossly deformed channels and re-use the acceptable :hannels.
This procedure verificd the corrett installation and function of the fuel channel measurement device fed was determined to be acceptable for use.
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2.8 PLANT PROCEDURE CHANGES Procedures des.ribed in the WNP-2 Final Safety Analysis Report (FSAR) re ,,
developed and used by the Plant Operating Staf f and various offsite suppon t1 -
organ 4ations. In 1987, the Plant Staf f made changes to procedures in accor-) i dance 4th 10CFR50.59 and concluded that none of the changes invgived unre- '
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viewe' ( d fety questions.
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Ohanges to procedures were generally either administrative or technical in /
nature. Administrative revisions consisted of title, organizational and
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editorial changes, while t4chnical changes were the result of system or com-ponent modif* cations or iidrovements in the procedural process. In all instances, a safety evaluation was conducted for each change in accordance with 10CFR50.59. All such evaluations were reviewed and approved by,the Plant Operating Comittee and are available for audit. It was conckded from the reviews that the probability of occurrence or consequences of an accident cr equipment malfunction was not increased, there was no reduction in any plant safety margins, and the possibyt,y of an accident or malfunction not Previ-ously evaluated was not increaseda 6 -
l Ouring September, 'a formal 10CFR50.59 Evaluation procedure was issued and, in accordance wi+h that process, the following procedural revision was identified as being a ;hange to a procedural commitment as described in the FSAR:
1 o ,PJf_12 3.:5. "Fire Protection Program Training" The purpose of this procedure is to establish the bdic' guidelines for Fire Brigade, Fire Watch and Fire Fighting which are ccins f stent with regulatory comitments, recognized industry, practices and Supply System programs. The procedure applies to the trtining '
of ' personnel for these programs. ,
The trorad w e was revised to provide (a) consistency with 10CFR50, Appendix R. :equirements and (b) flexibility regarding assigt.ments made to the F1 rte Brigade. *
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As a result of a 10CFR50.55 evaluation,. it was determineo that the revi-sion constituted a change to a procedural conrtitmer.t as described in the FSAR. Accordingly, an SAR Change Notice (SCti87-061 ) was prepared and will be sent to the NRC.
The following 's a discussion of significant plant procedure changes and development d.rics 1987:
- 1. PPM 1.3.5 "Reactor Trio and Recovery' The purpose of this procedure is to define the process for determining the causes of a reactor trip, provide guidelines for reacpor restart and document the plant response to the trip. ,
s 2 - 48 , f
2.8 PLANT PROCEDURE CHANGES (Continued)
The procedure was revised as a result of a thorough review of plant operations following the five unplanned scrams discussed previously in the Introduction Section of this report.
The following improvements were included in the revision:
The follow-up review comittee will be assembled immediately follow-o ing any reactor trip unless otherwise directed by the Plant Manager.
o When personnel performance is a contributing f actor to an event, a peer review committee will be formed to evaluate the decision process, o Predetermined experts will be relied upon for cause investigation in support of the restart decision process. It should be noted that this aspect is currently ef fectively performed and the change is to document who the resources are and facilitate management of the group.
o If the cause(s) of a reactor trip are indeterminate, a surveillance method will be developed, if possible, to monitor the component /
l l system response following restart.
o Independent verification will be required of significant conclusions relied upon in the restart decision process and root cause assessment.
- 2. PPM 1.3.43. "10CFR50.59 Evaluation Process" The purpose of this procedure is to provide guidance for performing i 10CFR50.59 evaluation and identify the plant procedures subject to the 10CFR50.59 evaluation process.
The procedure was developed to formalize the 10CFRM.59 evaiuation proc-ess at WNP-2. This procedure identifies the f ra:;;ional parts of the evaluation process and establishes the method for evaluating changes to structures, systems, componeni,s, procedures and proposed tests or experi-ments according to the requirements of 10CFR50.53. It is not the intent of the procedure to, in any way, limit required safety evaluations to only those proposed changes described in 10CFR50.59.
2'- 49 a
L8 PLANT PROCEDURE CHANGES (Continued)
- 3. Instrument Proaram Procedures o PPM 1.4.3. "Revision of Master Data Sheets and Setooints' The purpose of this procedure is to establish the responsibilities for the control of all plant setpoints and to provide a procedure for the preparation, review and approval of setpoint changes or other general Master Data Sheet changes. This also includes the control of calibration tolerances and Administrative Limits for instruments addressed in the Technical Specifications, o PPM 1.4.4. "Plant Instrument Desian Documentation" The purpose of this procedure is to provide instructions for the preparation and control cf Instrument Master Data Sheets.
o PPM 1.4.12. "Instrument SetDoint Calculations" The purpose of this procedure is to provide instructions for the preparation or review of instrument setpoint calculations.
These three procedures, which govern the Instrument Program at WNP-2, were revised to assure that setpoints are set equal or conservative to
' the Technical Specification Limits. The revisions were the result of an NRC Notice of Violation (NOV87-26) where it was identified that Main Steam Line Tunnel Temperature-High setpoints for channels "A", "B', 'C' and 'O' were set to a value (156*F) higher than allowed (150'F) by the Technical Specifications.
The Supply System acknowledged the validity of the violation in that the Main Steam Tunnel Leak Detection instrumentation was calibrated to a value higher than the Technical Specification Trip Setpoint without prior NRC approval. However, it should be noted that the decision to calibrate the MSL Tunnel instruments to the allowable limit was based on our inter-pretation of the Technical Specification Bases (Section 3/4.3.2). It is stated in the Bases that, "Operation with a trip set less conservative than its trip setpoint but within the specified allowable value is aJceptable on the basis that the difference between each trip setpoint and the allowable value is equal to or less than the drif t allowance assumed for each trip in the safety analysis."
This interpretation was considered to be a programatic concern and, as a result, the procedures were revised accordingly.
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2.9 REACTOR COOLANT ACTIVITY CUMULATIVE IODINE LEVELS This section contains information relative to reactor coolant cumulative iodine levels and iodine spikes. The specific activity of the primary coolant was significantly less than the limits of (a) less than or equal to 0.2 microcuries per gram dose equivalent I-131, and (b) less than or equal to 100/E micro-curies per gram as set forth in WNP-2 Technical Specifications.
Following a scheduled reactor shutdown, water sample analysis confirmed pre-vious indications of fuel failure. The iodine spike noted on December 6, reflects increased iodine concentrations due to the fuel pin failure.
A graph showing cumulative iodine dose equivalent for the calendar year 1987 is provided for reference and completeness. This inforri;ation is provided in accordance with WNP-2 Technical Specifications.
L 2 - 51 l
( WNP-2 DOSE EQUIVALENT IODINE 10 0 ; ,
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-7 Iiiiinl iil Ieaieill l iil eiiiIiieiil iiiiIiieiiIiei JAN 1 FEB 12 MAR 26 MAY 7 JUN 18 JUL 30 SEP 10 OCT 22 DEC 3 1
1987 12DEO PRI 8
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WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968
- 3000 George Washington Way
- Richland. Washington 99332
. Docket No. 50-397
-February 25, 1988 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:
Subject:
NUCLEAP. PLANT N0. 2 ANNUAL REPORT
Reference:
- 1) Title '0, 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 s'#,mi t: the- Annual Report for calendar year 1987. Should you have any questions or comments please contact M. R. Wuestefeld, WNP-2 Plant Engineer-ing Supervisor, Reactor Systems.
Very truly yours, h
C.M. Powers Plant Manager CMP:MRW:TRW Attachments ,,q 7 cc: Dottie Sherman, ANI ii iDocument Control. Desk.:NRC; '
P 723 058 %I
. _ . . . . . . -