ML20099J476

Annual Operating Rept for 1984

ML20099J476
Person / Time
Site:	Columbia
Issue date:	02/28/1985
From:	Martin J WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8503190732
Download: ML20099J476 (22)

Text

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Washington Public Power Supply System 9${

. P.O. Box 968 3000Gt wgeWashingtonWay Richland, Washington 99352 (509)372-5000 1;35 Pj,R -It Mi !O: 'I' TimuF Docket No. 50-397 February 28, 1985 Mr. J.B. - Martin Regional Administrator Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596

Subject:

NUCLEAR PLANT NO. 2 1984 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) Supply System to NRC (JB Martin) Letter of 12/20/84 G02-84-654 The Reference 1) states that "the licensee shall maintain records of changes in - the facility-and of changes in procedures.made pursuant to this section, to the extent that such changes constitute changes in the facility as described in the safety analysis report or constitute changes in procedures as described in the safety analysis report. The licensee shall also maintain records of tests and experiments car-ried out pursuant to paragraph (a) of this section. ' These records shall include a written safety evaluation which provides the bases for the determination that the change, test or experiment does not involve an unreviewed safety question."

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, _, NUCLEAR PLANT N0. 2 - ANNUAL REPORT I -

Further it states that the licensee submit

. . .. annually or = at such shorter intervals as may be specified in the license, a report containing a brief description of such changes, tests, and experiments, including a summary of the safety evaluation of each."

l Pursuant to this reference, the Supply System has procedures in place

that require a written safety evaluation be performed for all procedure changes, all changes to the Final Safety Analysis Report, and all changes in the facility. These evaluations provide the bases for the determination that the change, test or experiment does not involve an unreviewed safety question.

As a result of discussions with representatives of your staff, the Supply System has detennined that only significant safety evaluations should be reported and all others be available for audit purposes as i necessary. Significant safety evaluations are thosa in which it is not 4

readily apparent that the change does not involve an unreviewed safety

question as defined in 50.59(a)(2). In other words, further analysis and evaluation is required to arrive at a conclusion that an unreviewed safety question does not exist. The attached report submits those safety evaluations recognized by the Supply System according to the j

above criteria.

4 This information was included in a December 20, 1984 report (Ref. 3) ,

that covered the initial year of Plant operation. The interval between

i. December 20, 1984 and December 31, 1984 is covered by Attachment A to i

this report. Future annual reports will cover the yearly period ending

, December 31 and will include the 50.59 summary report.

Reference 2) states that l " Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted' prior

. to March 1 of the year following initial criticality."

Further. it states that the annual report shall include 4

"a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiv-ing exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions * (e.g.,

reactor operations and surveillance, inservice inspection, i

i

l J.B. Ma'rtin Page 3 NUCLEAR PLANT NO. 2 - ANNUAL REPORT routine maintenance, special maintenance [ describe mainte-nance], waste processing, and refueling). The dose assign-ments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD)...or film badge measurements. Small exposures totaling less than 20%

of the individual total dose need not be accounted for. In this aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions; and t

b. Documentation of all challenges to main steam line safety /

relief valves."

These are included as Attachments B and C to this report.

A narrative summary of Plant operating experience is included as At-

., tachment D. This attachment also includes a summary of forced outages and addresses the lack of Emergency Core Cooling System outages or indications of fuel failure.

Should you have any further questions, please contact Mr. R.L. Koenigs, WNP-2 Compliance Engineer.

Very truly yours, b?$0%- f J. D. Martin Plant Manager JDM:RLK:m Attachments cc: R Auluck - NRC WS Chin - BPA AD Toth - NRC Site RC DeYoung - NRC D Sherman - ANI Document Control Desk - NRC (18 copies)

Attachment A Page 1 of 1 e

T_ESTS OR EXPERIMENTS During this period WNP-2 was involved in normal Plant connercial operations and no other tests and/or experiments were conducted during this period.

CHANGES TO PROCEDilRES Procedures described in the WNP-2 FSAR are used by the Plant Operating Staff and by various offsite support organizations. -

that none of the changes involved unreviewed safety questio Changes na ture. to procedures were generally either administrative or technical in Administrative changes consisted of title, organizational and editorial changes, while technical changes were the result of system or i component modifications, or improvements in procedural processes. A safety evaluation was conducted for each change, in accordance with 10 CFR 50.59, and was reviewed and approved by the appropriate personnel.

The review concluded mal function were notthat the increned , probability of o::currence or consequences of an accide margins, and the possibility of an accidentthere was no reduction in any plant safety evalua ted was not increased. All safety evaluations or malfunction not previously performed have been reviewed and accepted by the Plant Operations Committee per the WNP-2 Techncial Specifications and are available for audit as necessary.

CHANGES IN THE FACILITY Inasmuch safety impact, as all changes design to non-safety related systems could potentially have a changes evaluated in accordance with 10 CFR 50.59regardless of safety classification are this safety evaluation is documented in each case and available for auditfor unrevie As a result of the design control process utilized at WNP-2, all technical and safety questions are evaluated and resolved during the design review process:

no changes were made in the plant that increased the probability of occurrence or consequences of an accident or equipment malfunction or reduced any plant safety margins, previously evaluated. or increased the possibility of an accident or malfunction not Since throughthe12/31time frame for which this report is being made is short (12 changes approv/84), very few changes were incorporated and in no caseswere /13/84 forward. No further anlaysis or evaluations were required.ed for which th i

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Attachment C Page 1 of 9 .

MAIN STEAM LINE SAFETY / RELIEF VALVE CHALLENGES During this period WNP-2 was involved in Plant Startup, the Power Ascension Test Program (PATP) and Commercial Operation. The following documents all challenges to WNP-2 main steal line safety / relief valves during the period.

Detailed Information:

S/R Valye Serial Number 63790-00-0048 63790-00-0054 63790-00-0055 63790-00-0059 63790-00-0045 Component ID (Location) MS-RV-1A MS-RY-2A MS-RV-3A MS-RY-4A- MS-RV-1B Date of Actuation (Mo/Da/Yr) 8/6/84 8/6/84 8/6/84 8/6/84 8/6/84 Time of Day (24 Hour Clock) 1606:19 1901:20 1712:18 1423:45 1831:43 Type of Actuation (Code) B B B B B Cause/ Reason for Actuation (Code) C C C C C Rx Operating Condition Prior to Lift (Code) B B B B B Rx Power Level Prior to Lift

(% Rated Thermal) 48.6 48.03 48.33 49.0 48.14 Tailpipe Temperature Prior to Lift (*F) Ambient Ambient Ambient Ambient Ambient Other. Instrumentation -

Type (Code) A A A A A i Other Instrumentation -

Number, Reading and Units Rx Pressure Prior to Actuation (PSIG) 919 918 919 920 918 Reseat Pressure at Valve Closure (PSIG) 917 915 915 917 915 Duration of This Actuation (Minutes: Seconds) 2 min 30 sec 1 min 24 sec 1 min 26 sec 1 min 23 sec 1 min 35 sec Failures, Reports (Code) -- -- -- -- --

. Attachment C -

Page 2 of 9 S/R Valve Serial Number 63790-00-0049 63790-00-0053 63790-00-0057 63790-00-0046 63790-00-0047 Component ID (Location) MS-RV-2B MS-RV-3B MS-RV-4B MS-RV-1C MS-RV-2C' Date of Actuation (Mo/Da/Yr) 8/6/84 8/6/84 8/6/84 8/6/84 8/6/84 ,,

Time of Day (24 Hour Clock) 1642:05 1319:07 1704:02 1632:37 1843:57 Type of Actuation (Code) B B B B B Cause/ Reason for Actuation (Code) C C C C C Rx Operating Condition Prior to Lift (Code) B B B B B Rx Power Level Prior to Lift

(% Rated Thermal) 48.66 49.34 48.29 48.42 48.1 i Tailpipe Temperature Prior to Lift (*F) Ambient Ambient Ambient Ambient Ambient Other Instrumentation -

Type (Code) A A A A A

- Other Instrumentation -

Number, Reading and Units Rx Pressure Prior to Actuation (PSIG) 91 9 919 918 918 9) 3 Reseat Pressure at Valve Closure (PSIG) 91 6 917 915 91 6 914 Duration of This Actuation (Minutes: Seconds) 1 min 49 sec 4 min 18 sec 1 min 23 sec 1 min 36 sec 1 min 48 sec Failures, Reports (Code) -- -- -- -- --

Attachment C

. Page 3 of 9 S/R Valve Serial Number 63790-00-0051 63790-00-0058 63790-00-0050 63790-00-0052 63790-00-0056 Component ID (Location) MS-RV-3C MS-RV-4C MS-RV-1D MS-RV-2D MS-RV-30 Date of Actuation (Mo/Da/Yr) 8/6/84 8/6/84 8/6/84 8/6/84 8/6/84 Time of Day (24 Hour Clock) 1539:13 1726:56 1738:35 1311:23 1650:25 Type of Actuation (Code) B B B B B Cause/ Reason for Actuation (Code) C C C C C Rx Operating Condition Prior to lift (Code) B B B B B Rx Power Level Prior to Lift

(% Rated Thermal) 48.8 48.27 48.31 48.06 48.2 ,

Tailpipe Temperature Prior to Lift (*F) Ambient Ambient Ambient Ambient Ambient Other Instrumentation -

l Type (Code)- A A A A A Other Instrumentation -

Number, Reading and Units Rx Pressure Prior to Actuation (PSIG) 919 918 91 9 918 920 -

Reseat Pressure at Valve Closure (PSIG) 917 915 915 915 915 Duration of This Actuation .

(Minutes: Seconds) 2 min 8 sec 1 min 38 sec 1 min 53 sec 2 min 2 sec 1 min 39 sec Failures, Reports (Code) -- -- -- -- --

Attachment C l Page 4 of 9 S/R Valve Serial Number 63790-00-0060 63790-00-0061 63790-00-0062 ,

Component ID (Location) MS-RV-4D MS-RV-5B MS-RV-SC

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Date of Actuation (Mo/Da/Yr) 8/6/84 8/6/84 8/6/84 l Time of Day (24 Hour Clock) 1403:32 1754:14 1415:14 Type of Actuation (Code) B B B Cause/ Reason for Actuation (Code) C C C Rx Operating Condition Prior to Lift (Code) B B B Rx Power Level Prior to Lift

(% Rated Thermal) 49.16 48.3 49.08 Tailpipe Temperature Prior to Lift (*F) Ambient Ambient Ambient Other Instrumentation -  !

Type (Code) A A A .)

Other Instrumentation -

Number, Reading and Units Rx Pressure Prior to l

Actuation (PSIG) 91 9 918 920 Reseat Pressure at Valve Closure (PSIG) 917 915 917 Duration of This Actuation (Minutes: Seconds) 3 min 11 sec 1 min 36 sec 2 min 24 sec Failures, Reports (Code) -- -- -- -- --

Attachment C Page 5 of 9 .

S/R Valve Serial Number 63790-00-0049 63790-00-0058 , ,

Component. ID (Location) MS-RV-4A MS-RV-4C Date of Actuation (Mo/Da/Yr) 10/28/84 10/28/84 Time of Day (24 Hour Clock) 0426 0551 Type of Actuation (Code) B B Cause/ Reason for Actuation (Code) E E Rx Operating Condition Prior to Lift (Code) G G Rx Power Level Prior to Lift *

(% Rated Thermal) 0 0 Tailpipe Temperature Prior to Lift (*F)

Other Instrumentation -

Type (Code) A' A Other Instrumentation -

Number, Reading and Units 100% open 100% open Rx Pressure Prior to Actuation (PSIG) 1000 980 Reseat Pressure at Valve Closure (PSIG) 949 (Tm 0434) 817 (Tm 0553)

Duration of This Actuation (Minutes: Seconds) 8 3 min Failures, Reports (Code) --

C C

Attachment C Page 6 of 9 ,

S/R Valve Serial Number 63790-00-0061 63790-00-0056 63790-00-0062 63790-00-0060 63790-00-0057 Component ID (Location) MS-RV-5B MS-RV-3D MS-RV-5C MS-RV-4D MS-RV-4B Date of Actuation (Mo/Da/Yr) 10/28/84 10/28/84 10/28/84 10/28/84 10/28/84 '

Time of Day (24 Hour Clock) 0405 0408 0411 1415 0420 Type of Actuation (Code) B B B B B

'Cause/ Reason for Actuation (Code) E E E E E Rx Operating Condition Prior i to Lift (Code) G G G G G Rx Power Level Prior to Lift .

(% Rated Thermal) 0 0 0 0 0 Tailpipe Temperature Prior to Lift (*F)

Other Instrumentation -

Type (Code) A A A A A Other Instrumentation - ,

Number, Reading and Units 100% open 100% open 100% open 100% open 100% open Rx Pressure Prior to Actuation (PSIG) 1000 1000 1000 1000 1000 ,

Reseat Pressure at Yalve

Closure (PSIG) 930 920 910 900 910 Duration of This Actuation ,

(Minutes: Seconds) Not Required Not Required Not Required Not Required Not Required Failures, Reports (Code) C C C C C

Attichment C Page 7 of 9 .

S/R Valve Serial Number 63790-00-0062 ,

Component ID (Location) MS-RV-5C Date of Actuation (Mo/Da/Yr) 11/10/84 Time of Day (24 Hour Clock) 1154:30 Type of Actuation (Code) B' Cause/ Reason for Actuation (Code) E Rx Operating Condition Prior to Lift (Code) G  ;

Rx Power Level Prior to Lift

(% Rated Thermal) -

Tailpipe Temperature Prior to Lift (*F) 211*F Other Instrumentation -

Type (Code) A Other Instrumentation - Acoustical Number, Reading and Units 100%

Rx Pressure Prior to Actuation (PSIG) 880 Reseat Pressure at Valve

Closure (PSIG) 640 Duration of This Actuation (Minutes

Seconds) 5 min.

Failures, Reports (Code) C

Attachment C Page 8 of 9 ,

S/R Valve Serial Number '63790-00-0053 63790-00-0059 63790-00-0050 63790-00-0061 63790-00-0056-t Component ID (Location) MS-RV-38 MS-RV-4A MS-RV-4D MS-RV-5B MS-RV-3D Date of Actuation (Mo/Da/Yr) 11/10/84 11/10/84 11/10/84 11/10/94 11/10/84 Time of Day (24 Hour Clock) 1135 1135 1135- 1140 1149-Type of Actuation (Code) A A A A A Cause/ Reason for Actuation (Code) A,C A,C A,C E E Rx Operating Condition Prior ,

to Lift.(Code) E E E G G .

. Rx Power Level Prior to Lift -

(% Rated Thermal) 91% 97% 97%- 97% --

Tailpipe Temperature Prior to Lift (*F) 212*F 129'F 211*F 210*F 155*F

- Other Instrumentation -

Type (Code) A A A A A

! Other Instrumentation - Acoustical Acoustical Acoustical Acoustical Acoustical .

3 Number, Reading and Units 'l00% 100% 100% 100% 100%

, Rx Pressure Prior to

Actuation (PSIG) 1080 1080 1080 1040 985 Reseat Pressure at Valve Closure (PSIG) 1000 1000 1000 875 855 Duration of This Actuation (Minutes

Seconds) 20 sec 20 sec 20 sec- 3 min 3 min Failures, Reports (Code) C C C C C J

J

Attachment C Page 9 of 9 Codes:

Type of Actuation 4

A. Automatic B. Remote Manual C. Spring Cause/ Reason for Actuation . . _

A. Overpressure B.- ' ADS or Other Safety C. Test D. Inadvertent (Accidental, Spurious)

E. Manual- Relief Reactor Operating Condition Prior to Lift (LER Codes)

, A. Construction B. Preoperational Startup or Power Ascension Tests in Progress C. Routine Startup D. Routine Shutdown E. Steady State Operation F. Load Changes During Routine Operation i G. Shutdown (Hot or Cold) Except Refueling H. Refueling Other Instrument-Type i

t A. Acoustic Monitor l B. Pressure Sensor C. Other Failures-Reports -

A. Failure of Electrical or Other Components Not Considered Part of Valve Assembly - No SRVS Failure Report is Required ,

B. Failure of Any Part of Valve Assembly - SRYS Failure Report Will be Filed '

C. No Failures Occurred - No SRVS Report Required 4 D. LER Submitted - Give LER Number in Item 316 E. NPRDS Will be Submitted

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Attachment D Page 1 of 8

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WNP-2 OPERATING

SUMMARY

WNP-2 was conducting initial fuel completion of fuel loading, the loading activities as 1984 began. At the Plant obtained its first criticality on January 19, 1984 and conducted Low Power Physics Testing until the Power Ascension Test Program (PATP) began on April 10, 1984. PATP testing continued through December 12, 1984 on December 13, 1984 and the Plant was declared in Comercial Operation WNP-2 first synchronized trical system and producedwith the Bonneville Power Administration (BPA) elec-electrical power on May 27, 1984 WNP-2 gross electrical energy production for 1984 was 1,804,110 MWH.

Plant tion ofoperation fuel in 1984 produced no evidence of any fuel failures. No indica-or samples. failure was obtained from either offgas normal operating levels Plant chemical or radiological analysis.Also, no indications of fuel failure were ob WNP-2 committed to reporting Emergency Core Cooling System outages for a five year period to provide data for availability analysis.

period covers only commercial operation and is very short (12/13/84Since 12/31/84), throughthis reporting no data is included in this report.

mitted with the 1985 ECCS outage data as part of the 1985 annual report.The 1984 Following is a summary of Plant outages and forced power reductions.

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Attachment D' Page 2 of 8 DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE 05/27/84 Forced 0.1 Equipment Auto Upon initially loading the generator, a turbine Failure Scram trip occurred due to antimotoring. The prot 1em!

was determined to be in the DEH control system.

Note that the reactor did not scram.

05/28/84 Forced 13.2 Equipment Auto Auto scram on low RPV level due to loss of con-Failure Scram densate booster pump and feed pump while placing the condensate demins in service. See LER 84-051 05/29/84 Scheduled 14.0 Testing M n1 The generator was unloaded for turbine valve testing. As a result of that testing, a reactor scram was initiated due to rapid closure of the bypass valve and the subsequent high pressure spikes.

05/30/84 Scheduled 2.0 Testing Manual The generator was unloaded from the grid to per-form turbine overspeed testing.

05/30/84 Forced 10.5 Maintenance Manual The generator was rernoved from service and reactor power was reduced in order to install fuses in the RCIC and condensate systems. See LER 84-048, 6/25/84.

06/01/84 Forced 27.3 Equipment Auto Automatic scram occurred on high reactor pressure Failure Scram as a result of the closure of all four main tur-

. bine bypass valves due to a DEH malfunction. A replacement logic card was installed in the DEH System. See LER 84-056, 6/28/84.

06/03/84 Forced 234.7 Equipment Manual A plant shutdown was completed on 06/03/84 as a Failure result of increasing conductivity on the primary system caused by leaking main condenser tubes.

Repairs were made to the condenser to correct problem.

CNOTE: No releases or radiation exposure occurred which were associated with these outages and which accounted for more than 10 percent of the allowable annual values. ,

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Attachment D*

Paga 3 of 8 DATE TYPE . OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECllRRENCE I '06/13/84 Forced 0.5 Equipment -Auto The Generator tripped on high reactor level while Failure Trip transferring feedwater. control. The turbine was

relatched and changes to the feedwater control-r logic.were subsequently made.

• l '06/13/84 Forced . 16.3 Equipment Auto Auto scram occurred on low level due to a loss Failure Scram of feedwater following condensate booster and L

feedwater pumps tripping on low suction pres-sure. The low suction pressure was due to the ,

condenaste cleanup flow control valve failing to open with only two condensate filter demins. -in service.' The cause for valve failure was deter-

. mined and the valve was replaced. See LER 84-060 06/19/84 Scheduled - 0.4 Te= ting Manual The Generator was tripped as a part of the Power Trip Ascension Test Program. Following the test, the Generator was placed back on line.

06/19/84 Forced -302 Equipment . Manual Plant shutdown to repair a turbine bypass valve ,

Failure which stuck open following a planned trip of the Generator. The valve was repaired, however the i ge was extended due to repairs on RHR Pump 07/10/84 Forced 553.5 Equipment Manual During monthly surveillance testing of Standby Failure Diesel Generator 1B (DG1B) and slip ring end bearing turned on the shaft insulation, thus

, destroying the insulation and allowing the shaft to drop slightly and rub on bearing housing.

Modifications, to improve reliability, were made-by mounting the bearings directly to shaft and insulating the bearing housing. See LER 84-075.

08/03/84 Forced 15.1 Equipment Manual Plant was shutdown to repair a severe steam leak Failure from a handhole on the Moisture Separator Heater Drain tank. The leak was repaired by seal weld-ing the handhole to prevent recurrence.

Attachment D*

Paga 4 of 8 DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE 08/07/84 Scheduled 80.9 Testing Auto Loss of Power test conducted as part of the Power Scram Ascension Test Program. .

08/12/84 Forced 66.3 Equipment Manual Plant was shutdown due to high conductivity Failure caused by a condenser tube leak. One failed condenser tube was found in the center of a tube bundle. The tube was plugged and a chemistry guidance letter issued to aid in early conduc-tivity excursion assessment. See LER 84-083.

08/16/84 Forced 30.1 Testing Auto MSL - Hi RAD DIV II surveillance caused 1/2 scram Scram SCRAM and C.R. Block flow comparator surveil-lance caused second 1/2 SCRAM for DIY I result-ing in a Reactor trip. See LER 84-089 C08/17/84 Forced 5.3 Equipment Manual Could not attain sufficient vacuum after startup Failure due to 6A and B High Pressure FW Heater ruptured discs, downstream of RY's, being ruptured from previous SCRAM. The cause was determined and the ruptured discs replaced.

08/18/84 Forced 265.8 Equipment Manual Plant was shutdown due to high conductivity Failure caused by condenser tube leaks. It was deter-mined there were nine (9) relatively minor tube leaks and one (1) major leak. The cause of 1eakage was determined, the leaking tubes were plugged and the condenser returned to service.

Attachment D-Paga 5 of. 8 '

DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE 09/10/84 Forced 180.8 Testing Auto A test switch which was intended for a test trip Scram on both RRC pumps was inadvertently connected .to -

the RPS logic. When closed in, it resulted' in the failure of power fuses to all four RPS chan-nels and a resultant Scram. The cause of the problem was determined while the plant was pro-ceeding to cold shutdown. An inspection and analysis of the incident revealed no harmful

. effects. The fuses were replaced and the proce-dure subsequently - performed. Since the proce-dure was a one time test on further corrective action is required. See LER 84-095 09/27/84 Forced 18.1 Equipment Manual Plant was shutdown due to failure of the Linear Failure Variable Differential Transformer (LVDT) for RRC-V-60B. This failure was a result of the differential transformer core becoming detached from the actuating rod. The LVDT was replaced with another of improved design.

10/01/84 Scheduled -132.5 Testing Manual Test trip of turbine-generator at 75% power as part of Power Ascension Test Program. The tur-bine bypass valves fast opening response time

, did .70t meet test criteria. Testing and trouble shooting to correct problems.

l 10/07/84 - Scheduled 0.4 Testing Manual Tripped turbine-generator at 24% power for by-

! pass valve (BPV) response time testing. Test did not meet cri teria. Continued testing and

, troubleshooting.

10/08/84 Scheduled 0.4 Testing Manual' Tripped turbine-generator again at 24% power for BPV response time testing. Test did not meet criteria. Continued testing and troubleshooting.

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Attachment D Page 6 of.8 DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE 10/08/84 Forced 53.3 Maintenance Manual Plant shutdown due to failure' to meet test cri-teria. Continued testing and troubleshooting' to correct problems.

10/11/84 Scheduled 0.2 Testing Manual Tripped turbine-generator at 24% power ' for another test of BPV response time. Test failed i to meet test criteria. Resumed testing and troubleshooting.

10/11/84 Scheduled 0.3 Testing Manual Tripped turbine-generator at 24% power for another BPV response time test. Response time was satisfactory and turbine-generator was returned to service.

10/13/84' Forced 20.3 Equipment Manual Plant shutdown due to cycling of turbine Failure Scram governor & bypass valves. It was determined after shutdown that cycling was caused by radio frequency interference due to keying of hand held radio transmitters in the vicinity of elec-trosyn pressure transmitters.

All susceptable instrument locations were ident-ified and warning signs were placed in the area of DEH pressure transmitters. Action is also being taken to evaluate the possibility of mod-ifying electrosyn transmitters to reduce sensi-tivity to radio frequency interference. See LER 84-109 10/20/84 Forced 95.2 . Equipment- Auto Reactor. scrammed on low steam pressure due to an Failure Scram erroneous setpoint being initiated while making a pressure change. A plant modification (PMR)

, has been initiated to improve visibility of set-points displays. See LER 84-112 7

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Attachment D  !

Page 7'of 8 ,

. DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE

.10/28/84 Forced 32.5 . Equipment Auto Reactor scrammed from 92% power on low level due-Failure Scram to loss of condensate booster pumps from L low suction pressure. Low suction pressure was caused by the steam seal condenser bypass valve failing closed while troubleshooting valve pro-blems. The valve was repaired and plant returned to service. See LER 84-114 11/10/84 Scheduled . 249.7 Testing Auto Plant was shutdown by Reactor SCRAM, initiated Scram by main steam isolation valve closure test, as part of the test and ascension program. The test was successful _ and the plant remained down for a scheduled maintenance outage.

11/21/84 ' Scheduled 0.5 Testing Manual Tripped turbine generator at 24% power to test i bypass valves capacity .as part of the power i ascension test program. Test was satisfactory j and turbine generator was returned to service.

J 11/27/84 Forced 70.1 Equipment- Auto Reactor SCRAM at 40% power level due to 1ow Failure Scram condenser vacuum caused by a FW heater tube leak

, and leaking FW heater shell side relief valve.

The tube leak was repaired and modifications

were made to relief valve and plant was returned to service. See LER 84-125 12/02/84 Scheduled 25.5 Testing Auto Initiated a generator load reject trip at 100%

Scram power as part of the Test and Ascension Program.

12/03/84- Forced 12.0 Operator Auto Reactor scram at 25% power on low water level Error Scram caused by loss of feedwater flow while trans-ferring feedwater control from - fl ow control valve to speed control. See LER 84-124 i

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• 4 Attachment D Page 8 of 8 DATE TYPE OUTAGE (Hrs) REASON SHUTDOWN *CAUSE & CORRECTIVE ACTION TO PREVENT RECURRENCE 12/28/84 Forced 41.7 Equipment Auto A reactor SCRAM occurred from 100% powei due t.o Failure Scram a turbine trip, caused by actuation of a gener-ator protection relay from a drop in the auto stop oil header pressure. The fluctuation in pressure occurred when the air side seal oil pump was removed from service. Investigation revealed the setpoint of auto-stop pressure switch was high and the : auto stop oil. pressure was abnormally low. The switch setpoint was calibrated to its proper value and the auto stop oil orifices and relief valves were cleaned to restore ' normal oil header pressure. See LER 84-129