Text

LER 93-010-07:on 940203,WNP-2 TS Identified as Part of TS Surveillance Improvement Project.Caused by Less than Adequate Procedures.Procedures,Ts & Design Changes implemented.W/940314 Ltr
	ML17290B029
Person / Time
Site:	Columbia
Issue date:	03/14/1994
From:	Mackaman C; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17290B032	List: ML17290B029; ML17290B033;
References
	GO2-94-061, GO2-94-61, LER-93-010, LER-93-10, NUDOCS 9403230287
	Download: ML17290B029 (74)
	v • d • e

LICENSEE EVEhOIEPORT (LER)

'GILITY NAME (1) DOCKET NUHB R (f) PAGE (3)

Washin ton Nuclear Plant - Unit 2 0 5 0 OI0 3 9 7 I OF 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS EVENT DATE (5) LER NUHBER 6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

HONTH DAY YEAR YEAR SEQUENTIAL EV I 5 ION HONTM DAY YEAR FACILITY NAHES CKE NUMB R (5)

NUMBER UMBER 0 5 000 0 2 0 3 9 4 9 3 0 I 0 0 7 0 3 1 4 9 4 000 PERATING HIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREHEHTS OF 10 CFR 5: (Check one or more of the following) (11 ODE (9) I OWER LEVEL 20.402(b) 20.405(C) 50.73(a)(2)(iv) 77.71(b)

(i0) 20.405(a)(1)(i) 0.36(c)(1) 50.73(a)(2)(v) 73.73(c) 0.405(a)(1)(ii) 0.36(c)(2) 50.73(a)(2)(vii) THER (Specify in Abstract 0.405(a)(l)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) elow and in Text, NRC 0.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) orm 366A) 0.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LER (12)

'TELEPHONE NUHBER C.D. Mackaman, Licensing Engineer REA CODE 5 0 9 7 7 - 4 4 5 1 COMPLETE ONE LIHE FOR EACH COHPONEHT FAILURE DESCRIBED IH THIS REPORT (13)

CAUSE SYSTEH COMPONENT MANUFACTURER EPORTABLE I',".$44 CAUSE SYSTEH COMPONENT HANUFACTURER EPDRTABLE 0 NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED SUBMISSION MONTH DAY YEAR ATE (15)

YES (If yes, complete EXPECTED SUBHISSION DATE) NO 05 12 94 TRACY TTIO On March 4, 1993, a condition of noncompliance with WNP-2 Technical Specifications was identified as part of a Technical Specification Surveillance Improvement Project (TSSIP). This two-year project was recommended by a Supply System Quality Action Team formed as a corrective action of LER 91-013-02.

The TSSIP revises and broadens the scope of the Surveillance Procedure Verification Program completed in May 1991.

A total of 23 reportable problems identified by this process are described in this LER. All 23 items relate i to failure of procedures to fully implement WNP-2 Technical Specification surveillance requirements.

This LER reports the initial findings of the TSSIP surveillance procedure review process. Based upon previous experience with the Surveillance Procedure Verification Program, it is likely that additional reportable items will be identified. A supplement to this LER will be submitted on an approximate monthly basis, or as necessary, to describe future reportable items. I Immediate and further corrective actions include, but are not limited to, entering Technical Specification Action Statements, additional testing, Plant Procedure changes, Technical Specification changes, and design changes.

9403230287 9403i4 PDR ADOCK 05000397 8 PDR

'LICENSEE EVENT REPORT (l)

TEXT CONTINUATION ACIL1TY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 2 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

~Ast~c (Cont'd)

The root causes for these events include less than adequate barriers and controls for program changes and less than adequate test procedures, directives/requirements, and design. The general root cause has been determined to be less than adequate management control of the Surveillance Test Program.

The safety significance of each item and the whole surveillance program was evaluated and it has been concluded that this event had potential safety significance.

Plant n ii n Power Level - 100%

Plant Mode - 1 (Power Operation)

Event D cri i n On March 4, 1993, a condition of noncompliance with WNP-2 Technical Specifications was identified as part of a Technical Specification Surveillance Improvement Project (TSSIP). This is a two year project recommended by a Supply System Quality Action Team formed as a corrective action of LER 91-013-02.

The TSSIP is staffed by Contract Engineers and Supply System employees, and revises and broadens the scope of the Surveillance Procedure Verification Program completed in May 1991.

The previous Surveillance Procedure Verification Program was a five week Technical Specification surveillance implementation review. This was a limited scope review that compared Technical Specification surveillance requirements with information obtainable from the Scheduled Maintenance System (SMS) data base. The surveillance procedures were reviewed for purpose, but not content or methodology. Approximately 145 discrepancies were identified during the review.

In contrast to the previous review, the TSSIP review is an in-depth technical review of the surveillance procedures to ensure they meet Technical Specification surveillance requirements. The review criteria includes proper test methodology, procedure consistency, technical accuracy, and reference bases for acceptance criteria. The goals of the project are to assure:

1. That related procedures required to be performed to satisfy Technical Specification surveillance

~ requirements are referenced (listed) and explained in the Purpose section of the procedure.

2. That prerequisites and special conditions required to assure Technical Specification compliance are stated in the procedure.

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TEXT CONTlNUATlON ACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 3 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

3. That procedure acceptance criteria satisfy the Technical Specification surveillance requirements and acceptance criteria have reference bases.

4. That procedure steps associated with assuring Technical Specification acceptance criteria are met and identified.

5. That numerical values, setpoints, tolerances, calculations, graphs, figures, and tables included or referenced in the procedure are consistent with values specified in Technical Specifications.

6. That the procedure tests the entire channel, including sensor, indicators, alarms, and trip functions as applicable.

7. That the procedure performance frequency meets Technical Specification requirements.

8. That the procedure satisfies the applicable Technical Specification surveillance requirements and meets the intent of the Technical Specification Bases.

Potential deficiencies will be evaluated for validity and necessary follow-up actions.

A total of 23 reportable problems identified by this process are described in this LER. All 23 items relate to failure of procedures to fully implement WNP-2 Technical Specification surveillance requirements. This LER reports the initial findings of the TSSIP surveillance procedure review process. The project was initiated November 1, 1992, and is scheduled to continue through April 1994. Based upon previous experience with the Surveillance Procedure Verification'rogram, it is likely that additional reportable items will be identified. A supplement to this LER will be submitted on an approximate monthly basis, or as necessary, to describe future reportable items.

This LER is written with each item discussed as a separately numbered paragraph under the major headings of Specific Event Description, Immediate Corrective Action, Further Evaluation, Specific Further Corrective Action, and Specific Safety Significance. A general discussion of all items is found under the major headings of General Event Description, above, and General Further Corrective Actions, General Safety Significance, and Similar Events, below.

ecific Event D cri on n- f- 1 R ir 1 inPm Tri Surveillance Requirement 4.3.4.2.3 requires the End-Of-Cycle (EOC) Recirculation Pump Trip

-(RPT) circuit breakers to be tested at least once per 60 months to demonstrate that arc suppression time is less than or equal to 83 milliseconds. Technical Specification Surveillance

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'LtCENSEE EVENT REPORT (ltd)

TEXT CONTINUATION ACILITY NAHE (1) OOCKET NUHBER (2) LER NOHBER (8) AGE 3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 4 F 51 TITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS (TSS) 7.4.3.4.2.3.3A, "EOC-RPT Breaker Arc Suppression Time RPT-3B/RPT-4A," and TSS 7.4.3.4.2.3.3B, "EOC-RPT Breaker Arc Suppression Time RPT-3A/RPT-4B," were used to perform this test. However, a review of these procedures discovered that they actuate Trip Coil 1 (TC-1) for EOC-RPT circuit breaker arc suppression response time testing, and not Trip Coil 2 (TC-2). TC-2 performs the actual EOC-RPT breaker trip safety function, whereas, TC-1 performs the normal and Anticipated Transient Without Scram (ATWS) RPT breaker trip functions. Since the electrical and mechanical characteristics of TC-2 could vary from that of TC-1, the test methodology is inadequate to assure the RPT breaker trip and arc suppression response time meets the surveillance requirement. Consequently, inadequate surveillance procedures caused the Plant to violate Technical Specification 4.0.3 by not satisfactorily completing the ACTION requirements within the allowed time. Technical Specifications 3.0.1 and 3.0.4 were violated when reactor power was increased to 30% without meeting the operational condition surveillance requirements, and by not entering Technical Specification Action Statement (TSAS) 3.3.4.2.e.

2. Trin v rVlv -F I r Surveillance Requirement 4.3.4.2.1 requires the EOC-RPT Turbine Governor Valve - Fast Closure system instrumentation to be demonstrated operable by the performance of a monthly Channel Functional Test (CFT) and a Channel Calibration (CC) every 18 months in accordance with Table 4.3.4.2.1-1.2. TSS 7.4.3.1.1.20, "RPS and EOC Recirc Pump Trip - TGV Fast Closure Channel A - CFT/CC," and TSS 7.4.3.1.1.78, "RPS and EOC Recirc Pump Trip -TGV Fast Closure Channel B - CFT/CC," were used to perform the CFT and CC. However, a review of these procedures discovered that they direct that certain safety-related function verification steps in the CFT not be performed, and marked "N/A" (Not Applicable), when reactor power is less than 30%. When these portions of the CFT were not completed, the CFT did not meet the surveillance requirements.

This also results in the CC not meeting the surveillance requirements because it takes credit for satisfactory completion of the CFT. WNP-2 Technical Specification definitions require a CC to include a CFT. Consequently, inadequate surveillance procedures caused the Plant to violate Technical Specification 4.0.3 by not satisfactorily completing the ACTION requirements within the allowed time. Technical Specifications 3.0.1 and 3.0.4 were violated when reactor power was increased to 30% without meeting the operational condition surveillance requirements, and by not entering TSAS 3.3.4.2.e.

3. T r in Thr I Valve- I sure Surveillance Requirement 4.3.4.2.1 requires the EOC-RPT Turbine Throttle Valve - Closure system instrumentation to be demonstrated operable by the performance of a monthly CFT in accordance with Table 4.3.4.2.1-1.1. TSS 7.4.3.8.2.1, "Monthly Turbine Valve Tests," was used to perform

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TEXT CONTlNUATION AGILITY NAME (1) OOCKET NUMBER (2 LER NUMBER (8 AGE 3 ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 5 F 51 iTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS this test. However, a review of the procedure discovered that it allows that certain safety-related function verification steps not be performed, and marked "N/A," if either Reactor Recirculation (RRC) pump is not in 60 Hertz operation. The RRC pumps are normally in 15 Hertz operation at a reactor power level less than 30%. When these portions of the CFT were not completed, the CFT did not meet the surveillance requirement. Consequently, an inadequate surveillance procedure caused the Plant to violate Technical Specification 4.0.3 by not satisfactorily completing the ACTION requirements within the allowed time. Technical Specifications 3.0.1 and 3.0.4 were violated when reactor power was increased to 30% without meeting the operational condition surveillance requirements, and by not entering TSAS 3.3.4.2.e.

4. -RPT em In m tion Surveillance Requirement 4.3.4.2.1 requires the EOC-RPT Turbine Governor Valve - Fast Closure system instrumentation to be demonstrated operable by the performance of a CC every 18 months in accordance with Table 4.3.4.2.1-1.2. The system logic is dependent on the proper operation of pressure switches MS-PS-3A, 3B, 3C, and 3D, which sense main turbine first stage pressure and enable the EOC-RPT logic at reactor power levels greater than or equal to 30%. Although these pressure switches are part of the EOC-RPT system instrumentation, no procedures were developed to meet the CC surveillance requirements. The Preventive Maintenance (PM) Program includes these pressure switches and instrument calibrations were performed at approximately 18 month intervals. However, WNP-2 Technical Specification definitions require that a CC include a CFT.

There is n'o assurance that acceptable CFTs were performed following each calibration.

Consequently, the lack of adequate surveillance procedures caused the Plant to violate Technical Specification 4.0.3 by not satisfactorily completing the ACTION requirements within the allowed time. Technical Specifications 3.0.1 and 3.0.4 were violated when reactor power was increased to 30% without meeting the operational condition surveillance requirements, and by not entering TSAS 3.3.4.2.e.

5. IRMN iveV l eP wr 1 N T ed On April 14, 1993, Technical Specification Surveillance Review personnel determined that aH Intermediate Range Monitors (IRMs) were inoperable. Personnel attributed the inoperability to a lack of a Logic System Functional Test (LSFT) of the negative-voltage-low IRM inoperative trip function. This trip function is provided with each IRM channel. The Reactor Manual Control System (RMCS) uses IRM inoperative trip signals to generate rod blocks, and the Reactor Protection System (RPS) uses these same inoperative trip signals to generate scrams. Technical Specification 4.3.1.2 requires "LSFTs and simulated automatic operation of all channels shall be performed at least once per 18 months." An LSFT is defined as "a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc., of a logic circuit, from

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~NSEE EVENT REPORT (Ol)

TEXT CONTINUATION ACILITY KAME (1) DOCKET NUMBER 2 LER NUMBER (8 AGE 3 umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 i 0 6 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS sensor through and including the actuated device, to verify operability. The LSFT may be performed by any series of sequential, overlapping, or total system steps such that the entire logic system is tested."

6. r Rn Mni r R nl nR On May 7, 1993, 'during the annual Maintenance and Refueling Outage, Technical Specification Surveillance Review personnel identified that there was a high probability that Surveillance Requirement 4.9.2.c.1 for SRM channel count rate verification was not being met. No surveillance procedure existed to assure compliance. The surveillance requirement is applicable prior to control rod withdrawal in Operational Condition 5 (Mode 5), and requires that each SRM channel be demonstrated operable by verifying that the channel count rate is at least 0.7 cps, provided the signal-to-noise ratio is greater than or equal to 20. Otherwise, the count rate must be greater than or equal to 3 cps, provided the signal-to-noise ratio is greater than or equal to 2. Plant Operators have been trained that if no specific procedural requirements exist for an activity required by Technical Specifications, the activity may be documented in the Reactor Operator's Log for compliance. However, a review of typical Mode 5 Reactor Operator's Log entries for control rod withdrawals in fueled control cells found no SRM channel count rate entries prior to the rod withdrawals. Since no evidence of consistent compliance with the surveillance requirement was found, WNP-2 has violated Technical Specification 4.0.3 in the past by not satisfactorily completing the ACTION requirements within the allowed time.

7. M in I 1 inVlv IV I r Tri B On June 9, 1993, with the plant in Mode 4 (Cold Shutdown), TSSIP personnel discovered a problem involving Main Steam system pressure switches MS-PS-20A, B, C, and D which provide MSIV closure trip bypass signals to the RPS. Bypass logic requires reactor pressure to be less than 1037 psig as sensed by these pressure switches and the reactor MODE switch not to be in RUN.

Increasing pressure opens the switch contacts which removes the bypass; conversely, decreasing pressure closes the switch contacts, which completes the bypass logic when the reactor MODE switch is not in RUN. In accordance with Technical Specifications Table 3.3.1-1, the trip must not be bypassed at 1037 psig or greater. Contrary to Table 3.3.1-1, TSSIP personnel determined that Instrument System Test Procedures PPM 10.27.2 and PPM 10.27.25 directed cognizant personnel to verify that the pressure switches opened at 1037+/- 6 psig; thus, the switches have reclosed at a pressure greater than 1037 psig and bypassed the trip function when not permitted by the Technical Specifications.

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TEXT CONTINUATION ACILITY NAHE (i) DOCKET NUHBER (2) LER NUHBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 7 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

8. in onrlR mRm -In k R i i nMni r On June 12, 1993, with the plant in Mode 4 (Cold Shutdown), TSSIP personnel discovered a problem involving main control room remote-intake radiation monitors WOA-RIS-31A(B) &

32A(B). These monitors monitor for radiation in the two divisional remote-air intakes to the main control room. Upon detection of a preset value of radiation, the monitors alarm the condition in the main control room and alert control personnel to isolate the affected intake. TSSIP personnel determined that Health Physics/Chemistry Shift Channel Checks Procedure TSS 7.1.1 was not in compliance with Technical Specification Definition 1.6, CHANNEL CHECK, in that "comparison of channel indications and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter" were not being performed.

9. I I r Tri F n i Surveillance Requirement 4.3.1.1 requires the MSIV closure trip (scram) instrumentation to be demonstrated operable by the performance of a CFT quarterly in accordance with Table 4.3.1.1-1.5.

TSS 7.4.3.1.1.9, "MSIV Closure Scram Functional," was used to perform this test. However, on July 9, 1993, with the plant in Mode 1 (Power Operation), TSSIP personnel determined that the procedure did not comply with Technical Specification Definition 1.7.b, CHANNEL FUNCTIONALTEST, in'hat each channel was not being fully tested to "... verify OPERABILITY including alarm and/or trip functions."

Each MSIV closure trip instrumentation channel functions to initiate a reactor scram logic signal when the associated MSIV is not fully open (approximately 10% closed). TSS 7.4.3.1.1.9 tests this function by visually verifying that the MSIV closure trip logic relays (RPS-RLY-K3[A - H]) drop out when their associated MSIV is not fully open. This methodology does not positively (i.e.,

electrically) verify the relay contact status to assure the trip channel alarm and/or logic relays (RPS-RLY-K14[A - H]) function as required.

10. RP r in -Thr le Valve lo re On August 9, 1993, it was determined that Technical Specification requirement 4.3.1.1.9, Channel Functional Test (CFT) at a quarterly frequency of the RPS Turbine-Throttle Valve Closure reactor scram logic, was not being adequately met. Specifically, the procedure did not test, as required by the Technical Specification definition for a CFT, each of the relays and alarms that constitute the logic. The RPS-RLY-K10 series relays were visually verified to deenergize as a result of testing, but the relay contacts were not verified either electrically or visually to have opened and the associated alarms were not verified to have annunciated.

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ACILITY HAHE (1) OOCKET HUHBER (2) LER HUHBER (8) AGE (3) umber ev. Ho.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS in mI 1 i nVlv Leak e nrl mPr r In i n wich On September 21, 1993, it was determined that the Main Steam Isolation Valve Leakage Control System (MSLC) Pressure Indicating Switch MSLC-PIS-60 was not being tested as part of a CFT of the MSLC pressure instrumentation. Technical Specification 4.6.1.4.d.1 requires that a CFT of the MSLC pressure and temperature instrumentation be performed monthly. MSLC-PIS-60 is part of the control logic for the outboard MSLC train which takes suction from the Main Steam system downstream of the outboard Main Steam Isolation Valves (MSIVs) and routes it to the Standby Gas Treatment (SGT) system post accident. MSLC-PIS-60 senses Main Steam pressure downstream of the MSIVs and closes the MSLC outboard depressurization valves if sensed pressure is greater than 1.4 psig and the depressurization valves have been open for greater than 50 minutes.

12. I I inIn i nR neTim T tin i i 1 On October 1, 1993, it was determined that the response time testing of the containment isolation valve logic was not performed in accordance with the requirements of Technical Specification 4.3.2.3. The particular components not tested are the final electro-mechanical relays for a portion of Isolation Groups 3 and 4. The containment isolation valves in Isolation Groups 3 and 4 are listed in Technical Specification Table 3.6.3-1. The existing response time testing procedures measure the system response time from the sensed parameter through two (out of a total of nine in two channels and out of a total of ten in the other two channels) relays per channel at the appropriate level of the system logic per division (see Attachment 1). In each case, these two relays that are response time tested are in parallel with, and of the same manufacturer and model type as the untested relays in each channel. In addition, the containment isolation valve response times (stroke times) were verified through testing.

13. Av eP w rRan eM ni r i semF nci nalT in On October 7, 1993, it was determined that the Logic System Functional Test (LSFT) of the Average Power Range Monitor (APRM) Flow Biased Simulated Thermal Power - Upscale logic was inadequate to satisfy Technical Specification 4.3.1.1-2.b. Specifically, APRMs E and F each provide a trip signal to two separate RPS logic channels. The APRM trip logic design is based on six APRMs and eight trip channels (see Attachment 2). The testing performed included verification of actuation for one, but not both, logic channel functions for APRMs E and F.

14. Ave e P w r Ran e Moni r Flow Bia ed im 1 ed Therm I P w r- Hi h On October 7, 1993, it was determined that the Channel Check of the APRM Flow Biased Simulated Thermal Power - Upscale signals was not being performed in a manner that meets the requirements of Technical Specification 4.3.1.1-2.b. The testing did provide for a comparison of

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TEXT CONTINUATION AGILITY NANE (1) OOCKET NUNBER (2) LER NUNBER (8) AGE (3) eer ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 51 iTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE "IMPROVEMENT PROJECT IDENTIFICATION

-OF NONCONFORMING CONDITIONS the outputs of each of the APRMs. However, the Technical Specification contains a note which requires WNP-2 to "Measure and compare core flow to rated core flow." This comparison was not being performed.

15. echni IVc mPm Tri ndI I i nT in On October 26, 1993, it was determined that the Main Steam Line Radiation Monitor - High trip and isolation of the mechanical vacuum pumps had not been tested since initial startup. Technical Specification Table 3.3.2-1, note (c) associated with the MSLRMs stated "Also trips and isolates the mechanical vacuum pumps."

16. R c rB ildin o re ion h m rV c mBr ker Valv On November 4, 1993, it was determined that no surveillance procedures were available to perform a visual inspection of Reactor Building to Suppression Chamber Vacuum Breaker Valves CSP-V-S, 6 and 9 in accordance with the requirements of Technical Specification 4.6.4.2.b.2.b. A visual inspection of these valves is required to be performed every 18 months. Credit was being taken for an external inspection of the valve. It was concluded that, like Reactor Building to Suppression Chamber Check Valves CSP-V-7, 8, and 10, an internal inspection was required to satisfy the Technical Specification requirement.

17. I i nA i nIn mn i nR neTimeT in indin 2 On November 16, 1993, it was determined that limited portions of the containment isolation actuation instrumentation logic for Isolation Groups 1, 2, 5, 6, 7, 8, and 9 had not been response time tested in accordance with the requirements of Technical Specification 4.3.2.3. A verbal notification of this condition was made at 1344 hours0.0156 days 0.373 hours 0.00222 weeks 5.11392e-4 months on November 16, 1993. Specifically, the interval not response time tested was the time from relay coil deenergization to the associated relay contact operation. The existing response time testing procedures measured the system response time from the sensed parameter to the relay coil voltage drop off, and the time between opening of a hand triggered contact and completion of mobilization of the final actuated device (valve). The testing did not measure the interval from relay coil deenergization to coil contact operation.

18. nr 1R m Emer enc Fil i n ni HEPA Fil rIn ion During the period between October 18, 1983 and July 15, 1991, Surveillance Procedure PPM 7.4.7.2.2 was credited with the performance of Control Room Emergency Filtration Unit WMA-FU-54A and WMA-FU-54B HEPA Filter inspections in accordance with TSS Requirement 4.7.2.c.1. This procedure is the predecessor to PPMs 7.4.7.2.2A and 7.4.7.2.2B, which are the current procedures credited with performance of the TSS requirement. The surveillance requirement

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1 LICENSEE EVENT REPORT (IOI)

TEXT CONTINUAT)ON AGILITY KAME (1) OOCKET KUKBER (2) LER KUKBER (8) AGE (3) ear ev. Ko.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 0 7 10 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS includes an in-place visual inspection of the ESF atmosphere cleanup system and all associated components using the guidance in Regulatory Position C.5.a of Regulatory Guide 1.52, Revision 2.

This guidance endorses the visual inspection criteria of ANSI N510-1975, Section 5.

A review of all revisions to PPM 7.4.7.2.2 was conducted by the TSSIP Group. The review concluded that Revisions 0 and 1 to the procedure did not include'all aspects of the required visual inspection (see Attachment 3). Revision 0 was performed only once, on April 9, 1985, for WMA-FU-54A and'WMA-FU-54B operability and was the first performance of the procedure following initial plant commercial operation. The failure to include all required visual inspection criteria in Revisions 0 and 1 was recognized after this initial performance of Revision 0. The omitted visual inspection criteria that satisfied TSS Requirement 4.7.2,c.1 was added to Revision 2 (approved June 30, 1986) and Revision 2 was used for the next performance of the procedure on October 2, 1986 (Revision 1 was never performed).

During the approximate eighteen month period between April 9, 1985 and October 2, 1986, WNP-2 was in OPERATIONAL'ONDITIONS requiring the Control Room Emergency Filtration. System to be OPERABLE in accordance with Technical Specification 3.7.2. Contrary to the Technical Specification requirement, the system was inoperable due to the failure to satisfy TSS Requirement 4.7.2.c.1 and the Technical Specification 3.7.2 ACTION requirements.

19. Is 1 i nAc i nIns men ti nRe neTimeT tin indin On December 17, 1993, it was determined that certain portions of the containment isolation actuation instrumentation logic for Group 4 had not been response time tested in accordance with the Technical Specification requirements. A Technical Specification Amendment was issued on October 15, 1993 to allow continued operation until the next cold shutdown without performance of the required response time testing of the Isolation Group 4 logic. The portions of the Group 4 logic that were identified in this item are different than those identified in item 12. Specifically, the interval not response time tested was the time from sensor relay coil deenergization to final associated contact operation in the operating logic of the subject valves. This affected a portion of the Group 4 valves.

20. n Tim T in f Emer en re lin In m n i n in i On January 10, 1994, with the plant in Mode 1 (Power Operation), TSSIP personnel discovered a deficiency in the testing method used to satisfy Surveillance Requirement 4.3'.3.3. This testing did not adequately measure the total response time of two in-series relays in the logic string for the opening of the injection valve in the Low Pressure Core Spray (LPCS) and Residual Heat Removal (RHR) B and C low pressure ECCS loops, and three in-series relays in the logic string for the injection valve in the RHR A low pressure ECCS loop.

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TEXT CONTINUATION FACILITY NAME (I) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 11 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS On January 12, 1994, similar deficiencies were found in the total response time testing of the logic strings for the start of the associated pumps: one relay in the logic string to the LPCS and RHR B and C pumps, and two relays in the logic string for the RHR A pump were not included in response time testing.

21. R n TimeT in fE In mn i n in in 2 On January 12, 1994, with the plant in Mode 1 (Power Operation), TSSIP personnel discovered a deficiency in the testing method used to satisfy Surveillance Requirement 4.3.3.3. This testing did not adequately measure the total response time of the actuation circuitry for the High Pressure Core Spray (HPCS) pump and discharge valve since two relays in the logic circuitry were not response time tested.

22. ervice W er V 1v P ii n V rificai On February 3, 1994, a TSSIP review determined that not all Standby Service Water (SSW) and High Pressure Core Spray Service Water (HPCSSW) valves were being verified every 31 days for correct position in accordance with TSS Requirements 4.7.1.1.a and 4.7.1.2.a. PPMs 7.4.7.1.1.1, "STANDBY SERVICE WATER LOOP A VALVEPOSITION VERIFICATION," 7.4.7.1.1.2, "STANDBY SERVICE WATER LOOP B VALVEPOSITION VERIFICATION,"and 7.4.7.1.2A, "HPCS SERVICE WATER VALVEPOSITION VERIFICATION,"'are credited with satisfying these requirements. However, certain safety-related equipment room cooler isolation valves, the RHR-P-2C seal water cooler isolation valves, and the Containment Atmospheric Control (CAC) scrubber and aftercooler isolation valves were not included in the surveillances,

23. HP In ec i n V 1veIn men ion On February 16, 1994, a TSSIP review determined that plant procedures are not adequate to satisfy the LSFT requirements of TSS Requirement 4.3.3.2. PPM 7.4.3.3.2.27, "HPCS-LSFT," is credited with meeting this requirement, but the procedure was found not to be in full accordance with Technical Specification Definition 1.22, "LOGIC SYSTEM FUNCTIONALTEST." The Technical Specification defines an LSFT as a test of all logic components of a logic circuit, from sensor through an'd including the actuated device, to verify OPERABILITY. This is clarified by Technical Specification Interpretation 94-01, which states: "In order to satisfy Tech Spec OPERABILITY requirements, the LSFT shall include testing of all features for which credit is taken in the WNP-2 accident mitigation design and licensing basis. Contrary to this interpretation, PPM 7;4.3.3.2.27 does not include adequate testing requirements for the Reactor Pressure Vessel (RPV)

Injection Valve HPCS-V-4 accident mitigating features taken credit for in WNP-2 Loss of Coolant Accident (LOCA) analysis.

"LlOENSEE EVENT REPORT (QIl)

TEXT CONTINUATION AGILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) ev. NO.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 07 12 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS mmedi t rr i A i Immediate corrective actions were initiated for each item discovered during the TSSIP procedure reviews.

They are enumerated below in paragraphs corresponding to the event description above:

n- f- I R ircl inPm Tri EOC-RPT System Channels A and B were declared inoperable and TSAS 3.3.4.2.e was entered at 1932 hours0.0224 days 0.537 hours 0.00319 weeks 7.35126e-4 months on March 4, 1993. Reactor power was reduced to 92% and the Minimum Critical Power Ratio (MCPR) was demonstrated to be less than the MCPR Limit at 2008 hours0.0232 days 0.558 hours 0.00332 weeks 7.64044e-4 months . Continued power operation was thereby authorized by the TSAS.

2. Trine vrn rValv -F I re No immediate corrective action was required as Turbine Governor Valve EOC-RPT System Channels A and B were in compliance with Surveillance Requirement 4.3.4.2.1 at the time of event discovery on March 9, 1993. TSS 7.4.3.1.1.20 and TSS 7.4.3.1.1.78 were satisfactorily completed at a reactor power level greater than 30% on February 19, 1993, and February 20, 1993, respectively.

3. Turbine Thr tie Valve - I re No immediate corrective action was required as Turbine Throttle Valve EOC-RPT System Channels A and B were in compliance with Surveillance Requirement 4.3.4.2.1 at the time of event discovery on March 9, 1993. TSS 7.4.3.8.2.1 was satisfactorily completed at a reactor power level greater than 30%, with both RRC pumps in 60 Hertz operation, on March 6, 1993.

4. -RPT emIn m n i

'I No immediate corrective action was required as Turbine Governor Valve EOC-RPT System Channels A and B were in compliance with Surveillance Requirement 4.3.4.2.1 at the time'of event discovery on March 9, 1993. Pressure switches MS-PS-3A, 3B, 3C, and 3D were all found to have been calibrated within the last 18 months. TSS 7.4.3.1.1.20 and TSS 7.4.3.1.1.78 meet the CFT requirements when performed at a reactor power level greater than or equal to 30%. As previously stated, they were satisfactorily completed on February 19, 1993, and February 20, 1993, respectively.

"LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATION ACILITY NANE (1) DOCKET NUMBER (2) LER NUNBER (8) AGE (3 ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 13 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

5. IRM i I P wr l T No immediate corrective action was required, because the IRMs were already deemed inoperable at the time Technical Specification Surveillance Review personnel discovered the IRM inoperability problem. The IRMs are normally declared inoperable in Mode 1, as associated CFT surveillances cannot be performed during this mode of operation.

6. RM h nnel n Ra Procedure deviations were prepared and incorporated into Fuel Handling Procedure PPM 6.3.2, "Fuel Shuffling and/or Offloading and Reloading," and Surveillance Procedure TSS 7.4.9.1, "Refuel Interlocks," to specify requirements to demonstrate adequate SRM channel count rate and signal-to-noise ratio prior to control rod withdrawal. These procedures govern activities that are imminent during the ongoing Refueling Outage, and that may require control rod withdrawal.

7. M IV I reTri B No immediate corrective action was required for the MSIV closure trip bypass problem because the reactor MODE switch was in the SHUTDOWN position and reactor pressure was below 1037 psig.

8. Main onr 1R mRem -In k Radi i nM ni r No immediate corrective action was required for the main control room remote-intake monitor problem, because this problem was discovered during Mode 4 of operation, and during this mode, the remote-intake monitors are not required to be operable.

M IV l sur Tri Funci n No immediate corrective action was required as all four MSIV closure trip function channels were in compliance with Surveillance Requirement 4.3.1.1-1.5 at the time of event discovery on July 9, 1993. TSS 7.4.3.1.2.1, the 18 month "Reactor Protection System" LSFT, satisfies the quarterly CFT requirements when taken in conjunction with TSS 7.4.3.1.1.9. TSS 7.4.3.1.2.1 verifies that the MSIV closure trip logic relays actuate the associated annunciators. TSS 7.4.3.1.1.9 verifies that when the MSIVs are not fully open, the MSIV closure trip logic relays actuate.

Together, these procedures meet the CFT requirements by the "sequential, overlapping" methodology allowed in Technical Specification Definition 1.'7. Both procedures were satisfactorily completed within the last quarter, TSS 7.4.3.1.2.1 on June 18, 1993, and TSS 7.4.3.1.1.9 on June 19, 1993.

LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATION FACIL1TY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE 3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 14 F 51 TiTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

10. RP Turbine-Throttle V Iv Iosure Procedure PPM 7.4.3.1.1.9A was written to include those portions of the RPS Turbine-Throttle Valve Closure CFT testing that were not covered by other procedures. This procedure was then performed satisfactorily on August 12, 1993, as part of the plant startup.

11. Min mI I inVIv k nr I m Pr r Indi i n witch The channel calibration procedure for MSLC-PIS-60, which also accomplishes a CFT of the switch, was successfully performed on September 22, 1993.

12. I I ionA i nInst m n tionR neTim Te in indin I A verbal request for discretionary enforcement was made by the Supply System on October 1, 1993, and discretionary enforcement was granted by the Staff. A written request for discretionary enforcement and a request for a Technical Specification amendment under emergency circumstances were submitted on October 2, 1993. A Technical Specification amendment was issued by the Staff on October 15, 1993, to allow continued operation until the Spring 1994 Refueling Outage without performance of the required response time testing of the Isolation Groups 3 and 4 logic.

13. Ave eP werR n eM ni r i temF n i naIT in A review of the computer records from the last performance of the CFT for APRMs E and F was made to verify that the appropriate relays deenergized when APRMs E and F tripped upscale.

PPM 7.4.3.1.1.47 (APRM E) and PPM 7.4.3.1.1.48 (APRM F) were changed to include verification of the necessary LSFT requirements. These procedures are used to perform both the LSFT and CFT testing of these APRMs. These procedures were satisfactorily performed on October 7, 1993.

14. Av ra eP werR n M ni rFI wBi im I Th rm IPow r-Hi h Plant procedure PPM 7.4.4.1.2 was revised to include a daily comparison of the expected drive flow signals to the drive flow input signal to each of the six APRMs. This procedure was then successfully performed on October 8, 1993.

15. h ni IV mP m Tri andI I tion T in Plant procedures PPM 7.4.3.1.1.11 and 7.4.3.1.1.56 were changed to include testing of the mechanical vacuum pump trip and isolation. These procedures were satisfactorily performed on October 27, 1993.

LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATION AGILITY KAHE (1) DOCKET KUHBER (2) LER KUHBER 8) AGE (3) umber ev. Ko.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 15 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

16. R rB ildin r i n m r mB kr 1v No immediate corrective action was required. Valves CSP-V-5, 6 and 9 were disassembled during the last refueling outage in May 1993. The Work Requests associated with that work were reviewed and determined to have satisfied the requirement for a visual inspection.

17. I 1 i nA i nInst mn tionRe neTimeT in in in 2 The response time testing of the untested portions of the isolation actuation instrumentation logic was performed using special test procedures. Technical Specification 4.0.3, which provides a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months grace period for the performance of past due surveillance procedures, was entered. Testing of Isolation Group 1 was completed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time period allowed by Technical Specification 4.0.3. Testing of Isolation Group 6 was completed in the time period allowed by the associated Action Statement.

On November 17, 1993, the Supply System requested and received Discretionary Enforcement for a four day period to allow performance of the required response time testing for Isolation Groups 2, 5, 7, 8, and 9. Special test procedures were written and performed to satisfy the Technical Specification response time testing requirements. This testing was successfully completed within the four day Discretionary Enforcement time period.

18. ontrol R m Emer enc Filtrati n nit HEPA Fil er In i n No immediate corrective action was required as the current WMA-FU-54A and WMA-FU-54B HEPA Filter inspections were in compliance with existing Technical Specification requirements at the time of event discovery on December 9, 1993. Revision 1 of PPM 7.4.7.2.2A and PPM 7.4.7.2.2B were reviewed by the TSSIP Group and found to include the visual inspection steps necessary to satisfy TSS Requirement 4.7.2.c.l. PPM 7.4.7.2.2A, Revision 1, was last completed on February 11, 1993 and PPM 7.4.7.2.2B, Revision 1, was last completed on February 20, 1993.

The required interval for the surveillance requirement is eighteen months, with the next scheduled due date for both procedures listed as June 22, 1994. Hence, the surveillances are current and no immediate operability concern exists.

19. I 1 i n Actuati n Inst men ti n R n e Tim T in indin As stated above, a Technical Specification Amendment was issued on October 15, 1993 to allow continued operation until the next cold shutdown without performance of the required response time testing of the Isolation Group 4 logic. The Supply System verbally notified the NRC Staff of the details of this condition on December 17, 1993. A letter detailing this condition was docketed on December 20, 1993. In order to response time test the maximum portion of the Group 4 logic that

'L'ICENSEE EVENT REPORT (Ol)

TEG CONTINUATlON FACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER 8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 07 16 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS could reasonably be performed at power, special test procedures were written to test the logic identified by this item. This response time testing was successfully completed on December 18, 1993.

20. R n Tim in f In mn in i i 1 A verbal request for Discretionary Enforcement was approved on January 11, 1994, allowing continued plant operation with the ECCS instrumentation functional, but not in strict compliance with Technical Specification testing requirements.

21. Res n e Time T tin of E Instrumen ion indin 2 The HPCS system was declared inoperable on January 12, 1994 pending performance of response time testing of the circuitry not previously tested.

22. Servi Waer V Ive Po iti n Verifi i n WNP-2 entered Limiting Conditions for Operation (LCO) 4.0.3 at 1315 hours0.0152 days 0.365 hours 0.00217 weeks 5.003575e-4 months on February 3, 1994 for failure to satisfactorily perform TSS Requirements 4.7.1.1.a and 4.7.1.2.a, The SSW and HPCSSW valves identified as requiring position verification were verified for proper position by using documentation for valve line-ups performed within the last 31 days or by physical inspection.

LCO 4.0.3 was exited at 2211 hours0.0256 days 0.614 hours 0.00366 weeks 8.412855e-4 months on February 3, 1994.

f

23. HP In ec i n lv In m ntation HPCS Level 8 instrumentation was declared inoperable at 1348 hours0.0156 days 0.374 hours 0.00223 weeks 5.12914e-4 months on February 16, 1994.

PPM 8.3.320, "ISOLATION - HPCS-V-4 LOGIC SYSTEM TEMPORARY TEST," was written and satisfactorily performed on February 17, 1994 to restore HPCS-V-4 instrumentation logic operability.

F herE 1 in d rr iv'A in Further Eval ati n These events are reportable under 10CFR50.73(a)(2)(i)(B) as "Any operation or condition prohibited by the plant's Technical Specifications...," and under 10CFR50.73(a)(2)(vii)(D) as "Any event where a single cause or condition caused... two independent trains or channels to become inoperable in a single system designed to... Mitigate the consequences of an accident."

"LICENSEE EVENT REPORT (&I)

TEXT CONTINUATlON FACILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (B) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 0 3 9 0 5 0 7 I 0 17 F iTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS There were no structures, components, or systems that were inoperable before the start of these events that contributed to the events.

Further evaluations were performed on each of the items discovered during the TSSIP procedure reviews.

They are enumerated below in paragraphs corresponding to the event description above:

1. End- f- cleR ir 1 ti n P m Tri In accordance with 10CFR50.72(b)(1)(ii)(B), this item was reported to the NRC Operations Center via the Emergency Notification System (ENS) at 2026 hours0.0234 days 0.563 hours 0.00335 weeks 7.70893e-4 months on March 4, 1993, as "Any event or condition during operation that... results in the nuclear power plant being... In a condition that is outside the design basis of the plant...." TSS 7.4.3.4.2.3.3A and TSS 7.4.3.4.2.3.3B were developed and approved on February 19, 1992, as a corrective action of LER 91-013-02. The previous surveillance procedure did not include the RPT-4A and RPT-4B circuit breakers in EOC-RPT breaker arc suppression response time surveillance testing.

The Surveillance Procedure Verification Program reviews did not identify the need to perform the response time testing using TC-2. Consequently, the LER did not include it as a corrective action.

2. rin vrn rVlve-F I r An investigation of TSS 7.4.3.1.1.20 and TSS 7.4.3.1.1.78 found that they were originally only the 18 month CC procedures. The monthly CFTs were conducted using TSS 7.4.3.1.1.19 and TSS 7.4.3.1.1.71. The CFT procedures met Surveillance Requirement 4.3.4.2.1 until they were revised on December 7, 1984. This revision added directions to mark certain status light and annunciator verification steps "N/A" when reactor power was less than 30%. The conditional steps were added in response to comments from the field, because the steps could not be performed as written.

They were being marked "N/A" by the field performers, with an explanation in the Comments section of the procedures. It was apparently not realized that the steps being marked "N/A" in the field, and now being made conditional, were required to verify RPS relay contact functional status.

They were, therefore, critical to the satisfactory completion of the CFT surveillance requirements.

When the CFT and CC were incorporated into Revision 5 of TSS 7.4.3.1.1.20 and TSS 7.4.3.1.1.78 on January 27, 1988, these conditional steps were carried over.

'iOE1VSEE EVENT REPORT +I)

TEXT CONTINUATION AGILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 18 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

3. r ine Thr l Valv - losure An investigation of TSS 7.4.3.8.2.1 found that the Note, allowing certain throttle valve position status light verification steps to be marked "N/A," was first added to Revision 5 of the procedure on April 15, 1987. Before this time, the procedure met Surveillance Requirement 4.3.4.2.1. The reason for the revision was given that 15 Hertz RRC pump operation causes an abnormal light configuration. The Revision 10 Note further clarifies this by stating that "Ifeither RRC pump is not in 60 Hertz operation, the... [turbine throttle valve position]... indicating lights will be extremely dim and monitoring of their status is difficult." However, based upon a review of previous procedure performances, there was no indication that the field performers had difficulty determining the light status. Apparently, the indicating lights are difficult, but possible, to use for throttle valve position status during 15 Hertz RRC pump operation. It was apparently not realized that the steps being made conditional were required to verify RPS relay contact functional status, and therefore, critical to the satisfactory completion of the CFT surveillance requirement.

4. E -RPT em In rumentation A review of the SMS data base for pressure switches MS-PS-3A, 3B, 3C, and 3D found they were being calibrated at approximately 18 month intervals under the Preventive Maintenance (PM)

Program. The pressure switch PM cards were recently revised to perform the calibrations in accordance with Plant Procedures PPM 10.27.53, "Main Turbine First Stage Pressure Switch Calibration Div 1," and PPM 10.27.54, "Main Turbine First Stage Pressure Switch Calibration Div 2." These procedures were developed and approved on March 18, 1993, to perform the pressure switch CCs every 24 months. They do not, however, reference Surveillance Requirement 4.3.4.2.1, nor do they meet the 18 Month CC surveillance interval requirement of Table 4.3.4.2.1-1.2. It is assumed that the failure to develop CC surveillance procedures for these pressure switches was due to an oversight during the initial procedure preparation process.

5. IRMNe iv VltaeP wr I N T General Electric Service Information Letter (GE SIL) 445, dated September 10, 1986, identified a blown fuse event at Monticello in which all positive and negative IRM fuses connected to the associated negative-voltage bus were blown by a power surge. After replacing the positive fuses, the IRMs appeared to be operating normally. But, because the negative-side fuses were not replaced, continued loss of the negative power supply prevented the IRMs from processing flux signals, and thus generating related IRM scram functions.

By design, the loss of the IRM's negative voltage supply was not annunciated, so the loss of the power supply, as well as the inability for the IRMs to generate scram functions remained undetected. The blown, negative-side fuses were detected later during IRM surveillance testing.

LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATION AClLITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3)

II ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 19 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS In response to this design error, the Supply System modified the IRM and SRM systems in June of 1987 to include a voltage sensing relay to detect the loss of the negative voltage supply, and upon loss of the negative voltage supply, generate IRM inoperative rod block and scram signals.

On April 14, 1993, TSSIP personnel discovered that related IRM LSFT requirements were considered, but deemed not necessary, during the design modification process. Further investigation revealed that the negative-voltage-low inoperative trips added to the SRM drawers had not been LSFT d since their installation, either. However, these SRM inoperative trips are not required to be LSFT'd by Technical Specifications.

6. RM hannl n R e According to Surveillance Requirement 4.9.2.c, SRM channel count rate verification must be performed prior to control rod withdrawal while in Mode 5. However, the surveillance requirement was never included in any WNP-2 fuel handling and refueling activity procedures to assure compliance. This failure to include the requirement in appropriate procedures was due to an oversight during the initial procedure preparation process.

Investigation of this event also identified related issues that should be addressed, they are described below:

a. Surveillance Requirements 4.3.7.6.c and 4.9.2.c both specify the channel count rate requirements for SRM channel operability. However, the requirements are not consistent.

Technical Specification Amendment No.'102 was issued on April 10, 1992, to change the SRM count rates and associated signal-to-noise ratios of Surveillance Requirement 4.9.2.c to the more conservative values recommended by GE in SIL 478. The applicability of the SIL to Surveillance Requirement 4.3.7.6.c was apparently overlooked during the Supply System's internal review of the amendment request.

Although Surveillance Requirement 4.9.2.c does not specifically establish a requirement for surveillance of signal-to-noise ratio, Surveillance Procedure TSS 7.4.9.2, "SRM Signal-To-Noise Ratio," was issued on May 15, 1993, to verify the signal-to-noise ratio at least once per seven days while in Mode 5. This is the SRM CFT frequency as specified in Surveillance Requirement 4.9.2.b.

C. Currently, the CC requirement of Surveillance Requirement 4.9.2.a.1 is being satisfied by Surveillance Procedure TSS 7.0.2, "Shift and Daily Instrument Checks (Mode - 5)." The procedure simply verifies that each SRM channel meets the count rate requirements of Surveillance Requirement 4.9.2.c. However, as defined by WNP-2 Technical Specifications, a CC should include a comparison of channel indications. To accomplish this, each channel

LICENSEE EVENT REPORT IIII)

TEXT CONTINUATION AGILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (B) AGE (3) umber, ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 20 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS count rate indication should be read, recorded, and compared against the acceptance criteria, the other channel indications and previous readings. This methodology would provide information and trendable data that could be a valuable aid in the early detection of increases in count rates, reduced signal-to-noise ratios, instrument errors, and channel failures.

7. M IV I r Tri B Further evaluation of the MSIV closure trip bypass problem determined that associated Instrument Master Data Sheets, as well as related Instrument System Test Procedures were not in compliance with Technical Specification Table 3.3.1-1.

M in nr IR m Rem e-In keR diation M ni r With respect to the main control room remote-intake radiation monitors, it was subsequently determined that three Health Physics/Chemistry Shift Channel Check procedures needed revision or clarification to be in agreement with Technical Specification Definition 1.6.

9. M IV Io r Tri F ncti n The Supply System committed to the test methodology established in Institute of Electrical and Electronic Engineers gEEE) 338-1975, "Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Class 1E Power and Protection Systems," Section 6.4, "Test Methods,"

for the surveillance testing program. The IEEE standard's methods for "positive and direct" relay actuation verification were not incorporated into the original version (Revision 0) of TSS 7.4.3.1.1.9, approved on February 9, 1984, due to an apparent misinterpretation of the requirements. This procedural deficiency during the initial surveillance procedure preparation process caused the failure to comply with the Technical Specification Definition 1.7.b requirement for verifying associated alarm and/or trip functions.

The 18 month LSFT procedure (TSS 7.4.3.1.2.1) and the existing CFT procedure (TSS 7.4.3.1.1.9) combine to meet the quarterly CFT requirement of Surveillance Requirement 4.3.1.1-1.5 only intermittently. This is generally only during the first quarter following each annual refueling outage based on performance of both tests near the end of each outage. Consequently, WNP-2 has not consistently met the surveillance requirement since initial plant startup.

10. RP r in -Throt le Valve losure Through a review of previous revisions of PPM 7.4.3.8.2.1 it was determined that this procedure has been inadequate to satisfy the Technical Specification requirements for RPS Turbine-Throttle Valve Closure relay and alarm testing since initial plant startup.

l LlCENSEE EVENT REPORT (Ol)

TEXT CONTINUATION ACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 21 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

11. Min mI I in alv k e nrl mPr r In i in wi It was determined through a review of past procedure revisions that a CFT of MSLC-PIS-60 has never been performed on a monthly basis as required by the Technical Specifications. This condition has existed since initial plant startup.

12. I I inA inln rumen inR n Tim T in ml A review of past revisions of the response time testing procedures showed that adequate response time testing of the Isolation Groups 3 and 4 logic was not being performed. This condition has existed since initial plant startup.

13. Avera e P wer Ran e Moni r ic em Func ional T in A review of past revisions of the APRM procedures showed that the LSFT/CFT procedure did not contain adequate verification of logic system response to satisfy the Technical Specification requirements. This condition has existed since initial plant startup.

14. v eP wrRan eM ni rF1 wBi im I er 1P wr-Hi h A review of past plant procedure revisions showed'that the daily comparison of the expected drive flow signals to the drive flow input signal to each of the six APRMs was not included. This condition has existed since initial plant startup.

15. Mechani I Vacuum Pum Tri and I I i n T in A review of plant procedures showed that this function had, since initial plant startup, not been tested periodically in accordance with the Technical Specification requirements.

16. c rBilin t r in hmer Br ker V lv A review of plant procedures showed that the visual inspection had not been performed in accordance with the Technical Specification requirements since initial plant startup.

17. I I tion Ac ation In men i n R n e Time T in indin 2 Based on a review of past revisions of plant procedures it was determined that the subject portions of the isolation actuation instrumentation for Isolation Groups 1, 2, 5, 6, 7, 8, and 9 were not

J I

-'+a 4

'L'ICIVSEE EVENT REPORT {)

TEXT CONT1NUATION ACILITY NAME (1) OOCKET NUMBER (2) LER NUMSER (8) AGE (3) umber ev., No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 22 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS included in response time tests since initial plant startup. These deficiencies were the result of not verifying proper overlap of testing performed in separate procedures. Use of separate procedures is acceptable so long as overlap of the logic, and thus response time testing of the logic, is performed.

18. nrolR mEmer nc Fil i n ni HEPAFi1 rIn i n As discussed above, a review of the revisions to PPM 7.4.7.2.2 determined that the procedure failed to include all TSS Requirement 4.7.2.c. 1 visual inspection criteria during the period between initial plant startup and approval of Revision 2 to the procedure on June 30, 1986. The deficiency was recognized sometime between the initial performance of the procedure on April 9, 1985 and the approval of Revision 2. There is insufficient information available to determine the exact cause of the procedure deficiency or exactly when the deficiency was discovered. However, the Supply System believes the cause to be an oversight during the initial procedure preparation process.

19. 1 ionA 'In men i nR n TimeT in in in Based on a review of the past revisions of plant procedures it was determined that the identified response time testing of Isolation Group 4 components had not been performed since initial plant startup. These deficiencies were the result of not verifying proper overlap of testing performed in separate procedures, Use of separate procedures is acceptable so long as overlap of the logic, and thus response time testing of the logic, is'performed.

20. Re neTimeTe in fE Ins m n tion in in 1 A written request for Discretionary Enforcement was made that described the additional testing deficiencies that had been discovered on January 12, 1994. An emergency Technical Specification change was requested and verbally approved by the NRC on January 13, 1994, to permit deferral of complete response time testing until the startup following the next cold shutdown, but no later than the startup following the Spring 1994 Refueling Outage. Formal written NRC approval of this change was received on January 31, 1994.

21. R n TimeTe in fE In men i n indin 2 Response time testing was performed using a temporary test procedure on January 14, 1994, within the time allowed by the associated Action Statement. The actual HPCS system response time was 22.4 seconds, compared with the Technical Specification limit of less than or equal to 27 seconds.

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LICENSEE EVENT REPORT k}

TEXT CONTINUATION AGILITY HAHE (1) DOCKET NUHBER (2) LER HUHBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 I 0 23 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

22. rvi W erV v P iionVrifi i n TSS Requirement 4.7.1.1.a states (Surveillance Requirement 4.7.1.2.a contains a similar requirement for HPCSSW):

"At least the above required standby service water system subsystem(s) [serves Division 1 and Division 2 diesel generators] shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position."

TSS Requirements 4.7.1.1.a and 4.7.1.2.a were originally understood by the Supply System to include only valves that would render the service water systems inoperable. Valves that branched off the main supply and return headers were not necessarily included in PPMs 7.4.7.1.1.1, 7.4.7.1.1.2, and 7.4.7.1.2A. This was based on the language of the service water system surveillance requirements and plant practice.

The surveillance requirements are intended to assure proper SSW PIPCSSVFJ flow paths servicing safety-related systems or components. To provide this assurance, valves in the service water main supply and return flow paths, as well as those in branch fiow paths to and from safety-related equipment, must be locked, sealed, or otherwise secured, or be verified in their correct position at least once per 31 days, The equipment room cooler and CAC scrubber and aftercooler isolation valves have not been included in surveillance procedures to verify correct position at least once per 31 days since initial plant startup.

RHR-P-2C cannot be utilized for shutdown cooling, which is the only mode of operation where the pump seals are exposed to water hot enough to damage them. Thus, the seal water cooler is not required for this pump and is spared in place. The seal water cooler isolation valves were removed from PPM 7.4.1.1.2 in 1988 and 1991.

23. HP In i n Valve In men i n In accordance with 10CFR50.72(b)(2)(iii)(D), this item was reported to the NRC Operations Center via the ENS at 1548 hours0.0179 days 0.43 hours 0.00256 weeks 5.89014e-4 months on February 17, 1994, as "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:... (D)

Mitigate the consequences of an accident." This notification became necessary when HPCS was declared inoperable in accordance with TSAS 3.5.1.c due to expiration of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time limitation of TSAS 3.3.3.b.

J LICENSEE EVENT REPORT I%I)

TEXT CONTINUATION AGILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) eat ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 I 0 07 24 F 51 TITLE (4)

TECHNICAL SPECIFICATION'SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS HPCS-V-4 instrumentation logic is designed to initially auto-open the injection valve to commence injection flow into the reactor vessel following a HPCS initiation on either low reactor water level (Level 2, i.e., -50 inches) or high containment drywell pressure (1.65 psig). The injection valve auto-closes to secure RPV injection flow when high reactor water level (Level 8, i.e., +54.5 inches) is reached. Ifwater level later drops to Level 2, the injection valve will automatically reopen to reestablish reactor vessel injection.

PPM 7.4.3.3.2.27, Section 7.5.3 is performed when reactor water level is below +54.5 inches.

This section directs the performer to have a Control Room Operator open HPCS-V-4 and then verifies that it auto-closes upon receipt of a high reactor water level (Level 8) signal. This section is considered to be in accordance with the Technical Specification definition of a LSFT.

Section 7.5.2 of the procedure is performed when reactor water level is at or above +54.5 inches.

This section does not verify that HPCS-V-4 auto-closes upon receipt of a high reactor water level (Level 8) signal. Instead, the relay (HPCS-RLY-K13) that initiates the RPV injection valve auto-closure is only verified to energize. This is performed by verifying that the relay contacts. close to illuminate the high reactor water level (Level 8) seal-in indicator light. However, the indicator light contacts are not the same contacts that initiate the RPV injection valve auto-closure. Consequently, there is no "positive and direct" assurance that the RPV injection valve will auto-close. As previously discussed, the Supply System committed to the test methodology established in IEEE 338-1975 for the surveillance testing program. The IEEE standard's methods for "positive and direct" relay actuation verification were included in Revision 0 of PPM 7.4.3.3.2.27. These methods were not incorporated into Section 7.5.2, which was added to Revision 1 of the procedure in 1985 prior to initial performance of the procedure (Revision 0 was never performed). It was not recognized that the procedure change required a change to Technical Specifications.

As previously described, HPCS-V-4 instrumentation logic is designed to automatically reopen the injection valve to reestablish reactor vessel injection if reactor water level drops to the low level (Level 2) actuation limit. This logic is separate from that which initially opens the RPV injection valve on a HPCS actuation, but is not verified in any WNP-2 surveillance procedures. This condition has existed since initial plant startup.

Based on the above evaluation, PPM 7.4.3.3.2.27 does not test all logic components of the logic circuit, from sensor through and including the actuated device, to verify HPCS-V-4 related RPV level instrumentation OPERABILITY. WNP-2 has been in OPERATIONAL CONDITIONS since initial plant startup that required HPCS system RPV level instrumentation to be OPERABLE in accordance with Technical Specification 3.3.3. Contrary to this requirement, HPCS-V-4 related RPV level instrumentation has been technically inoperable due to the failure of PPM 7.4.3.3.2.27 to satisfy TSS Requirement 4.3.3.2 and Table 3.3.3-1.C.1.c ACTION requirements.

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LICENSEE EVENT REPORT TEXT CONTINUATION

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ACILITY NAHE (I) DOCKET NUMBER (2) LER NUNBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 25 F 51 TITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS eneral R Five general root causes were identified by the Surveillance Procedure Verification Program in 1991, and remain valid for this review. They are described below:

Pr r Th n TA - Surveillance procedures developed during the startup period that do not fully implement the requirements.

2. h n e M n em n LTA - Procedure revisions, procedure deviations or plant changes that introduced errors into the Technical Specification Surveillance Program.

Dir iv ir LTA - Technical Specifications were accepted at the time of startup that

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3. R men c'ould not be complied with because of hardware restraints. These issues were recognized at the time, but were not adequately documented or resolved.

4. T fl Mflfl nf ~i flfl fly fl fll design.

5. Pr mm i n rol LTA - Plant Procedures do not provide adequate control of the Surveillance Testing Program.

S ecifi R a e Root causes were determined for each item discovered during the TSSIP procedure re'views. They are enumerated below in paragraphs corresponding to the event description above:

1. En - f- le R ircul ion P im Tri r

The root cause for the failure to properly test the EOC-RPT circuit breaker trip response time was Procedures LTA.

2. rin vrn r lv -Fa I ur The root cause for the failure of the CFT and CC to meet the surveillance requirements was Change Management LTA.

3. T r ineThr tleVIv - I ur The root cause for the failure of the CFT to meet the surveillance requirement was Change Management LTA.

1 LICENSEE EVENT REPORT TEXT CONTINUATION

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FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (8) AGE 3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 0 7 26 51 I'ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

4. -RPT mIn m n i The root cause for the lack of CFT and CC surveillance procedures for the EOC-RPT related main turbine pressure switches was Procedures LTA.

5. IRM N a iv V lta e P er 1 The root cause for the IRM and SRM negative-voltage-low inoperative trip functions not being LSFT'd was Change Management LTA; during the design change process, cognizant personnel considered surveillance testing of the IRM's negative-voltage-low inoperative trips, but deemed the testing unnecessary. Additionally, applicable revisions to the FSAR were not identified during the design change process.

6. RM h nnel nt R The root cause for the lack of procedural requirements to meet Surveillance Requirement 4.9.2.c.l was Procedures LTA.

7. I r Tri B The root cause for the MSIV closure trip bypass problem was Procedures LTA.

8. Main nr 1R mRem e-In k M ni r The root cause for the main control room remote-intake radiation monitor problem was Procedures LTA.

9. I r Tri Fn in The root cause for the failure to consistently meet Surveillance Requirement 4.3.1.1-1.5 was Procedures LTA.

10. P Tur ine-Thr le V lve 1 re

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The root cause for the inadequate CFT of the RPS Turbine-Throttle Valve was Procedures LTA in that plant procedures did not include testing of the necessary relays and alarms.

LICENSEE EVENT REPORT I%I)

TEXT CONTINUATION ACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) AGE 3 ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 27 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

11. Main mI I i nV IveLeaka e nr I mPre r Indi i n wi h The root cause for the failure to perform a monthly CFT on MSLC-PIS-60 was Procedures LTA in that no procedure was developed and scheduled to satisfy this Technical Specification requirement.

12. I I i n A i In mn i R Tim T in i in The root cause for the failure to adequately response time test the Isolation Group 3 and 4 logic was Procedures LTA in that the response time testing procedures did not include testing of the necessary components.

13. Av P wrRn eM ni r i F i I T in The root cause for the inadequate LSFT of the APRMs was Procedures LTA in that the procedures did not provide verification of the function of the necessary components.

14. Avera Power Ran e M ni r FI w Bia ed im I Thermal Power - Hi h The root cause for the inadequate APRM testing relative to comparing core flow to rated core flow was Procedures LTA in that this comparison was not included.

15. Mech ni I V c mPum Tri ndI I i T in The root cause for not testing the mechanical vacuum pump trip and isolation on MSLRM - High was Procedures LTA in that no procedures were written to satisfy the requirements of this note in the Technical Specifications.

16. React r B ildin u re ion h m er V um Breaker V Ives The root cause for not performing a visual inspection of valves CSP-V-5, 6, and 9 was Procedures LTA in that no procedures were written to satisfy this Technical Specification requirement.

17. I i A inIn mn inR neTim T in inin 2 The root cause for not response time testing limited portions of the logic for Isolation Groups 1, 2, 5, 6, 7, 8, and 9 was Procedures LTA in that the available procedures did not provide overlap testing of components.

LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATlON ACIL1TY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE 3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 28 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

18. nr1R mEm r en F'1 i n ni HEPAF'1 rIn i The root cause for the failure of PPM 7.4.7.2.2 to include all TSS Requirement 4.7.2.c.1 visual inspection criteria was Procedures LTA in that required inspection criteria was omitted due to an oversight during the initial procedure preparation process.

19. I 1 i nAc i nInst n ionR neTim Te in in in The root cause for not response time testing limited portions of the logic for Isolation Group 4 was Procedures LTA in that the available procedures did not provide overlap testing of components.

20. R n Tim T in fE In men i n in in 1 The root cause for the failure to test the ECCS instrumentation response times properly was Procedures LTA.

21. R n e Tim T in fE In rumentai n in in 2 The root cause for the failure to test the ECCS instrumentation response times properly was Procedures LTA.

22. ervice Waer V 1ve P i i n Verifi i n The root cause for not including certain safety-related equipment room cooler and CAC scrubber and aftercooler isolation valves in surveillance procedures that verify correct position at least once per 31 days was a misunderstanding that resulted in Procedures LTA.

1 The root cause for removing the RHR-P-2C seal water cooler isolation valves from PPM 7.4.1.1.2 without a Technical Specification change or the valves being locked, sealed, or otherwise secured was a misunderstanding that resulted in Change Management LTA.

23. HP In'eci n V 1veIn men i n The root cause for the incorporation of a procedure change which does not meet Technical Specification requirements was Change Management LTA.

The root cause for not including required HPCS-V-4 instrumentation logic functions in surveillance procedures was Procedures LTA.

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LICENSEE EVENT REPORT (%)

TEXT CONTINUATION FAClLITY NANE (1) DOCKET NUMBER (2) LER NURSER (8) AGE (3) ear ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 29 F 51 iTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS eneral F her rr ive A i Following the completion of the Surveillance Procedure Verification Program in 1991, the Supply System recognized that the high number of specific items of Technical Specification noncompliance was indicative of a broader programmatic issue. The five general root causes were reviewed to determine Technical Specification Surveillance Testing Program corrective actions. The results of the review are as follows:

For the Procedures LTA and Change Management LTA root causes, the following two actions were taken:

1. PPM 1.2.6, "PPM Evaluation Program," was revised on September 9, 1992, to strengthen the Technical Specification surveillance procedure verification process.

2. PPM 10.1.5, "Scheduled Maintenance System (SMS)," was revised on January 11, 1993, to include specific signoffs for SMS changes to Technical Specification surveillance requirements.

3. Revision of appropriate plant procedures was completed on July 12, 1993, to assign central "ownership" of the Surveillance Testing Program within the Technical Staff Department. Future surveillance procedures, and noneditorial changes and revisions to the existing surveillance procedures will receive a Technical Specification compliance review by the TSSIP staff.

The TSSIP is already underway to methodically review surveillance procedures by applicable Technical Specification. Procedures received prior to their scheduled review date will be screened for significant problems, but will not receive a detailed review until scheduled by the TSSIP staff.

For the Programmatic Controls LTA root cause, the WNP-2 Technical Specification Surveillance Testing Program was reviewed by a Quality Action Team (QAT), the Supply System formal problem solving process. The QAT completed their review and presented their findings and recommendations to Plant Management on April 17, 1992. The TSSIP, which discovered the items reported in this LER, is one of the QAT recommended actions being implemented.

There were no programmatic corrective actions applicable to the Directives/Requirements LTA and Design LTA root causes since the problems occurred before Plant startup, while under administrative controls that are no longer in affect. These root causes will be addressed on an individual basis by specific corrective actions.

LICENSEE EVENT REPORT h)

TEXT CONTINUATION FACILITY NAME (I) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 30 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS S ecifi F her rr iv Acti n

1. End- f- le Recircul ti P m Tri TSS 7.4.3.4.2.3.3A and TSS 7.4.3.4.2.3.3B have been revised to test the RPT-3A, 3B, 4A, and 4B breaker trip time response using TC-2.

2. Tur in yern r Valve- F I re TSS 7.4.3.1.1.20 and TSS 7.4.3.1.1.78 have been revised to meet the CFT and CC surveillance requirements of Table 4.3.4.2.1-1.2 when reactor power is less than 30%, as well as, greater than or equal to 30%.

3. Tur ine Thr le Valve- I sure TSS 7.4.3.8.2.1 has been revised to meet the CFT surveillance requirement of Table 4.3.4.2.1-1.1 when reactor power is less than 30%, as well as, greater than or equal to 30%.

4. EO -RPT tern Instrumentation Procedures have been revised or developed to meet the CFT and CC surveillance requirements of Table 4.3.4.2.1-1.2 for pressure switches MS-PS-3A, 3B, 3C, and 3D.

5. MNeaiveV I eP wr I N T t

a. On May 2, 1993, RPS Surveillance Procedure TSS 7.4.3.1.2.1 was changed to LSFT the voltage sensing relay that initiates the negative-voltage-low IRM inoperative trip. The relay functioned as designed.

b. The applicable surveillances have been revised to LSFT the negative-voltage-low SRM inoperative trip. This was completed before the RPS Shorting Links were removed.

c. An FSAR change notice was prepared on June 30, 1993, to reflect the negative-voltage-low inoperative trip as being part of the IRM and SRM trip circuitry.

d. The generic implications of inadequate change management are addressed through performance of the TSSIP review and current programmatic controls on Technical

'pecification surveillance revisions.

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LICENSEE EVENT REPORT TEXT CONTINUATION

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ACILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 31 F 51 1TLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVENENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

6. RM h n n R ao Surveillance Procedure TSS 7.4.9.2, "SRM Signal-To-Noise Ratio," was issued on May 15, 1993, to verify the signal-to-noise ratio at least once per 7 days while in Mode 5.

b. A Technical Specification Change Request was initiated on September 2, 1992, to make Surveillance Requirement 4.3.7.6.c consistent with GE SIL 478.

C. A change to the Technical Specification Bases for 3/4.3.7.6 and 3/4.9.2 was initiated on July 23, 1993, documenting a signal-to-noise ratio measurement frequency that satisfies SRM surveillance requirements.

d. A Mode 5 SRM Channel Check surveillance procedure was developed on July 30, 1993, that records and compares SRM channel indications in accordance with the requirements defined in Technical Specifications. Also, consistent procedural compliance methodology was verified for Modes 1, 2, 3 and 4.

e. A review of applicable plant operating and surveillance procedures was completed on July 30, 1993, to assure adequate procedural compliance with Surveillance Requirement 4.3.7.6.c in Modes 2, 3, and 4.

7. M IV lo ure Tri B at On June 14, 1993, an instrument setpoint change request was approved to change the MSIV closure trip bypass setpoint to comply with Technical Specification Table 3.3.1-1.

b. Instrument System Test Procedures PPM 10.27.2 and PPM 10.27.25 were deviated to achieve compliance with Table 3.3.1-1 on June 15, 1993.

C. Maintenance Work Request AP4166 was performed to recalibrate the pressure switches on June 15, 1993.

8. M in nrlR mRm e-In eR I i nM i r ao On June 14, 1993, the Chemistry Supervisor issued Standing Order 080 which directs cognizant personnel to "compare the readings from WOA-RIS-31A to WOA-RIS-31B and the readings from WOA-RIS-32A to WOA-RIS-32B." Results of these readings are being documented on Health Physics/Chemistry Shift Channel Check Procedure TSS 7.1.1.

LICENSEE EVENT REPORT TEXT CONTlNUATION

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ACILITY HAHE (I) DOCKET HUHBER (2) LER HUHBER (8) AGE (3) ear ev. Ho.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 32 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

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b. Health Physics/Chemistry Shift Channel Check Procedure TSS 7.1.1 was changed to incorporate Standing Order ¹80 on July 14, 1993.

9. M IV l r Tri F nci n TSS 7.4.3.1.1.9 was revised, and performed on September 9, 1993, to comply with Technical Specification Definition 1.7.b and the quarterly testing frequency of Surveillance Requirement 4.3.1.1-1.5.

10. P r in -Thr le V lve I re As stated in the immediate corrective action section above, procedure PPM 7.4.3.1.1.9A was written to include those portions of the CFT testing that were not covered by other procedures.

11. Min mI I inVlv k nrl mPr r In i i n wic A new procedure, PPM 7.4.6.1.4.18, was written to support both the CFT and Channel Calibration testing of MSLC-PIS-60. This procedure was approved by the Plant Operations Committee November 3, 1993. Periodic performance of this procedure has been scheduled through the Scheduled Maintenance System (SMS).

12. I I inA u inIn men inR neTimeTe in inin 1 The response time testing of the Isolation Groups 3 and 4 logic will be performed at the first Cold Shutdown condition no later than startup from the Spring 1994 Refueling Outage.

13. Ave e P wer Ran e Monit r Lo ic tern F n i nal Testin As stated in the immediate corrective action section above, the CFT procedures for APRMs E and F were changed to include the required testing.

14. Ave e P w rR n e M ni r Fl w Biased imul ed Therm 1P wer- Hi h As stated in the immediate corrective action section above, PPM 7.4.4.1.2 was changed to include the required Channel Check comparison of the expected drive flow signals to the drive flow input signal to each of the six APRMs.

LICENSEE EVENT REPORT R)

TEXT CONTINUATION AGILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) ear ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 07 33 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

15. M hni 1Vcu mPum Tri andIol i nTe in As stated in the immediate corrective action section above, procedures were changed to test the mechanical vacuum pump trip and isolation.

16. R rB' r i n Surveillance procedures will be developed to satisfy the visual inspection requirements for valves CSP-V-5, 6, and 9. These procedures will be available for use during the Spring 1994 Refueling Outage to satisfy the Technical Specification visual inspection requirement.

17. I I i nA ai nIn rumen ti nRes neTim T in in i 2 The required overlap testing will be included in the appropriate plant surveillance procedures prior to their use for the next required surveillance response time testing of the subject Isolation Groups 1, 2, 5, 6, 7, 8, and 9 logic.

18. r 1R m Emer enc Fil ra i n ni HEPA Filer In i As discussed in the event description section above, PPM 7.4.7.2.2 was revised on June 30, 1986 to include all TSS 4.7.2.c. 1 visual inspection criteria.

19. Isolation Ac a ion In rumen i n Res n e Time Te in indin The required overlap testing will be included in the appropriate plant surveillance procedures prior to their use for the next required surveillance response time testing of the subject Isolation Group 4 logic.

20. R n e Time Te in fE In men ion indin 1 The procedures used to perform the LPCS and RHR instrumentation response time testing have been revised to meet the Technical Specification surveillance requirements. Complete response time testing of this instrumentation will be performed before startup from the next cold shutdown.

21. R n e Time T tin of E S In rumen tion indin 2 il The procedures used to perform the HPCS instrumentation response time testing will be revised to meet the Technical Specification surveillance requirements before the next test, scheduled for the Spring 1994 Refueling Outage.

LICENSEE EVENT REPORT (Ol)

TEXT CONTINUATION AGILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) ear ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 07 34 F 51 TITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

22. rvi W er V lveP iti n Verifi i n PPMs 7.4.7.1.1.1, 7.4.7.1.1.2, and 7.4.7.1.2A have been changed to satisfy TSS Requirements 4.7.1.1.a and 4.7.1.2.a.

23. HP In'n lv I n i PPM 7.4.3.3.2.27 will be changed by May 25, 1994 to satisfy TSS Requirement 4.3.3.2 and Table 3.3.3-1.C.l.c.

eneral S fe i nifi n The Supply System regards the programmatic aspects of these items as an important issue that had potential safety significance. The General Corrective Actions listed above are defined to prevent recurrence of Technical Specification noncompliance problems in the future.

ecific Safe i nificance The Safety Significance was determined for each of the items discovered during the TSSIP procedure reviews. They are enumerated below in paragraphs corresponding to the event description above:

En - f- leRecir I i n P m Tri A review of circuit breaker test procedures found that EOC-RPT breaker testing is inadequate to assure the RPT breaker trip and arc suppression response time meets. the surveillance requirement.

Breaker testing is performed by actuating TC-1. No procedures were found in the SMS data base that verify the characteristics of TC-2, which performs the EOC-RPT breaker trip safety function.

The characteristics of TC-2 are assumed to be similar to TC-1 based upon previous operation of the EOC-RPT breaker trips during actual events. However, the breaker arc suppression response times using TC-2 have not been accurately measured to ensure they are within the plant design basis.

Consequently, this event was determined to have had potential safety significance since a delayed response time could have resulted in a delayed power reduction. Both EOC-RPT system channels were declared inoperable and the Plant remained in an LCO until corrective actions for this item were completed. See "Specific Further Corrective Actions" section for completed actions.

2. Tur in overn r Valve- F s I ur IJ The EOC-RPT Turbine Governor Valve - Fast Closure system instrumentation CFTs are performed monthly and satisfy Surveillance Requirement 4.3.4.2.1 when at a reactor power level greater than or equal to 30%. The EOC-RPT safety function is automatically bypassed at a reactor power level

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TEXT CONTINUATION AGILITY NAME (1) OOCKET NURSER (2) LER NUNBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 07 35 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS of less than 30%. Worst case, the longest period of operation in a noncompliance condition was 30 days. This fact, combined with the testing that was performed and the redundancy of the associated instrumentation, provides a high degree of confidence that the system could perform its safety function. Accordingly, this event was determined to have had no safety significance.

3. r in Thr t I lv - I re The EOC-RPT Turbine Throttle Valve - Closure system instrumentation CFT is performed monthly and satisfies Surveillance Requirement 4.3.4.2.1 when both RRC pumps are in 60 Hertz operation.

The RRC pumps are normally in 60 Hertz operation at a reactor power level greater than or equal to 30%. The EOC-RPT safety function is automatically bypassed at a reactor power level of less than 30%. Worst case, the longest period of operation in a noncompliance condition was 30 days. This fact, combined with the testing that was performed and the redundancy of the associated instrumentation, provides a high degree of confidence that the system could perform its safety function., Accordingly, this event was determined to have had no safety significance.

4. E -RPT tern In rumenta i n Pressure switches MS-PS-3A, 3B, 3C, and 3D were being calibrated approximately every 18 months by the PM Program to assure proper setpoint. The EOC-RPT Turbine Governor Valve - Fast Closure system instrumentation CFTs are performed monthly and satisfy Surveillance Requirement 4.3.4.2.1 when performed at a reactor power level greater than or equal to 30%. The pressure switches do not have an EOC-RPT safety function at a reactor power level of less than 30%, but serve only as an automatic logic bypass. Worst case, the longest period of operation in a Technical Specification noncompliance condition was 30 days. This fact, combined with the testing that was performed and the redundancy of the associated instrumentation, provides a high degree of confidence that the system could perform its safety function. Accordingly, this event.was determined to have had no safety significance.

5. RM Ne ive Volta e Power u I No T Plant Modification Request (PMR) 02-86-0204 added negative-voltage-low inoperative trips to each IRM and SRM chassis. Operability testing conducted during the design change process demonstrated that installed trips functioned as designed. The Supply System has no knowledge that these IRM trips have been inoperable, other than from a lack of LSFT testing, since the time of the modification. Since the testing performed indicates that the IRMs have been capable of performing their intended safety function, there is no safety significance associated with this event.

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1 LlCENSEE EVENT REPORT (eR)

TEXT CONTINUATION AGILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 36 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS

6. RM h nn 1 n Rate The Surveillance Requirement 4.9.2.c.1 SRM channel count rate verification noncompliance applied only to the "Prior to control rod withdrawal..." frequency. Plant Operators at WNP-2 performed the count rate verifications while in Mode 5 at eight hour shift intervals in accordance with Surveillance Procedure TSS 7.0.2. As a result, the longest period of noncompliance with the surveillance requirement was approximately eight hours. In addition, the SRM count rate verification information, and the instrument calibration and test data do not show a high incidence of failure. Thus, the. short intervals of noncompliance, the repetitive SRM channel verifications and testing that were performed, and the associated instrument channel redundancy combine to provide a high degree of confidence that the system could perform its safety function. Accordingly, this event was determined to have had no safety significance.

7. M IV I r Tri B a Setting the MSIV closure-trip-bypass pressure switch setpoint slightly higher than 1037 psig would have resulted in a very brief delay of the reactor scram on MSIV closure. However, this trip is redundant to the reactor high pressure trip of 1037 psig, which cannot be bypassed by the reactor MODE switch. Additional protection is provided by the Main Steam Safety Relief Valves (MSRVs), which provide electrical and mechanical overpressure relief of the reactor pressure vessel.

Therefore, the safety significance associated with this event is negligible.

8. Main C nr 1R mlRemote-IntakeRadiai n M ni r The main control room remote-intake radiation monitors were deemed to be technically inoperable due to less than adequate channel check procedures. However, there was no reason to believe that these monitors were unable to perform associated functions; therefore, the safety significance associated with this event is negligible.

9. M IV lo re Tri Func io The safety function of MSIV closure trip logic relays are for their contacts to open when the associated MSIV is not fully open. TSS 7.4.3.1.1.9 tests these relays every quarter to assure that they drop out. TSS 7.4.3.1.2.1 performs an LSFT at least annually to positively verify the relay contacts open to perform their trip and alarm functions. In addition, based on an equipment history review, there is no evidence of an incidence where these relays failed to drop out during testing or an identified condition where the contacts failed to open when the relay dropped out. This fact, c'ombined with the testing that was performed and the redundancy of the associated instrumentation, provjdes a high degree of confidence that the system could perform its safety function.

Accordingly, this event was determined to have had no safety significance.

LICENSEE EVENT REPORT QR)

TEXT CONTINUATION ACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 37 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS

10. P T r in -Thr le 1v 1 r The RPS trip on Turbine-Throttle Valve Closure is designed to limit the reactor power transient on a turbine trip event. The testing that was performed verified that the subject relays deenergized on demand. Testing performed after this problem was found verified that the associated relay contacts opened as required when the relays deenergized. Since relay contact failure to open generally continues until corrected, this verification testing provided evidence that the relays are and were capable of performing their intended safety function, Based on a review of the available evidence, the RPS trip on Turbine-Throttle Valve Closure was always capable of performing its intended safety function. Failure of this trip to function would result in a RPS trip on either APRM high neutron flux or Reactor Steam Dome Pressure - High. Therefore, this event is deemed to have had minimal safety significance.

Min mI 1 in Iv k nr 1 t mPr r In i i n witch The MSLC system is only required to operate post LOCA. MSLC-PIS-60 is calibrated, including a CFT, on an annual basis. The pressure in the Main Steam lines downstream of the outboard MSIVs must be less than 41 psig before the system can be manually placed in service. There is no automatic start of the system. Finally, if the pressure switch fails to function as designed, the MSLC outboard depressurization line remains open for an indeterminate period of time. This line discharges into a normally unoccupied area in the Reactor Building and the effluent is processed by the Standby Gas Treatment (SGT) system. Ifdepressurization of the Main Steam lines did occur but MSLC-PIS-60 did not sense this, the MSLC system would continue to draw a small vacuum on the Main Steam lines and to process the effluent through SGT, even though the depressurization line remained open. Therefore, failure to perform a CFT on a monthly basis is deemed to have had no safety significance since the consequences of failure are that MSLC would continue to allow the Main Steam line effluent to discharge into the Reactor Building which would then be processed through SGT, instead of MSLC discharging directly into SGT. SGT is designed to process both Reactor Building atmosphere and direct influents. Area Radiation Monitors would provide sufficient warning to plant personnel relative to potential high radiation conditions if the depressurization line remained open.

12. I 1 i n u i nIn men i nR neTimeT tin in in 1 A detailed evaluation was made of the impact of not having performed response time testing of the identified relays in the Isolation Groups 3 and 4 logic. The results of this evaluation were documented in the request for discretionary enforcement and the Technical Specification Amendment requests submitted on October 2, 1993. As stated above, the existing response time testing procedures measure the system response time from the sensed parameter through two (out of a total of nine in two channels and out of ten in the other two channels) relays per channel at the

I I IOENSEE EVENT REPORT R)

TEXT CONTINUATION AGILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 I 0 38 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS appropriate level of the system logic per division. In each case, these two relays that are response time tested are in parallel with, and of the same manufacturer and model type as the untested relays in each channel.

Each of the 19 relays, including those that were and those that were not response time tested, is functionally tested on at least an annual basis as part, of the logic system functional test. Response time testing history for logic strings using the model of relay in question has confirmed the reliability and repeatability of these relays. There is no observed failure mode that has caused deterioration of the dropout time of these relays. Industry experience is that failure to function is the expected failure mode for these relays. Finally, the response time of the untested relays (approximately 110 milliseconds dropout time) is a small fraction of the Technical Specification required total time from initiation signal to valve closure (5 seconds or greater).

A failure to isolate the Group 3 and 4 containment isolation valves affected by this testing deficiency within the time frame specified in the Technical Specifications would result in the potential for release from the primary to the secondary containment. The valves in this group that communicate directly with the containment atmosphere are normally closed valves. The remaining valves are part of a closed system either inside or outside containment. A system failure would have to occur, concurrent with a LOCA, for a release to occur. In either case, the release would be to the secondary containment where the release would be processed through SGT.

Based on the successful functional testing that is performed for these relays on an annual basis, the history of consistent response time performance of the relays that were response time tested, the relatively small contribution the untested relays make to the total loop response times, and the insignificant effect on effluent release that a small increase in response time would induce, this event had minimal safety significance.

13. Avera P w rR n eM ni r i emF nci n IT in The APRM E upscale trip deenergizes relays RPS-RLY-K12E and G. APRM F upscale deenergizes relays RPS-RLY-K12F and H. As shown by testing performed both before (as verified through the computer history of the most recent CFT results) and after this problem was discovered, the relays functioned as designed during testing but were not verified as part of previous testing. These same four relays are verified to function as part of the testing of the Intermediate Range Monitor (IRM) logic. The function of the APRMs was verified every three months during the CFT. Only deenergization of the two redundant relays went unverified.,

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'ICENSEE EVENT REPORT QR)

TEXT CONTINUATION FACILITY NANE (1) DOCKET NUMBER (2) LER NUNBER (8) AGE (3 ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 39 51 TITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS As shown in Attachment 2, the APRM RPS scram logic is "one out of two taken twice" for each RPS Trip system (A and B). Both trip systems must trip to complete a reactor scram. As shown on Attachment 2, three of the four APRM upscale inputs for each Trip System were tested in accordance with the Technical Specification requirements. These three inputs for each Trip System will cause a reactor scram.

Since: 1) the APRMs were verified to function properly at a three month test interval; 2) the subject two relays were verified to function at least yearly as part of IRM testing; 3) the relays functioned when tested; and 4) failure of the APRM inputs not functionally tested as part of the Logic System Functional Test on a 18-month frequency would not by themselves impact the ability of RPS to perform its intended safety function, failure to test each of the APRM inputs and relay deenergization every 18 months had minimal safety significance.

14. P wer R n e Monit r Fl w Bi ed im I ted Thermal P wer-Hi h

.Each APRM receives a flow input signal based on a summation of the recirculation loop flows.

There are four recirculation flow summer circuits. Each flow summer circuit continuously compares the unit output to the output of one of the other three fiow units. Ifa flow mismatch of greater than 10% occurs, a control rod withdrawal block and associated alarm are received. This continuous comparison by the flow summer units is comparable to the daily Channel Check required by the Technical Specifications.

The fiow signal inputs to the APRMs are calibrated on a weekly basis in accordance with Technical Specification 4.3.1.1-2.b. This calibration provides assurance that the flow signal input to the APRMs is accurate. The combination of the continuous flow unit comparator circuitry and the weekly calibration of the flow signal input to the APRMs results in minimal safety significance in not having performed a Channel Check of the APRM flow signal inputs on a daily basis.

15. M hani IVc mP m Tri ndI 1 ion Te in The MSLRM - High trip and isolation of the mechanical vacuum pumps was installed to limit the release of radiation from the main condenser to the environment. This function is credited for the design basis control rod drop accident. 'A redundant automatic isolation of the mechanical vacuum pump lines is provided by the radiation monitors on the vacuum pump exhaust lines. In addition, the MSLRMs provide an annunciator function. Operator action, as directed by the annunciator response procedure, would be to verify the mechanical vacuum pump trip and isolation.

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'ICENSEE EVENT REPORT l4}

TEXT CONTINUATION FACILITY NANE (I) DOCKET NUMBER (2) LER NUNBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 0 7 40 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS Since the trip and isolation functions performed properly when tested on October 27, 1993, a redundant trip/isolation signal is available, and since the equipment would be isolated by an operator shortly after the automatic isolation signal, should the automatic isolation fail to occur, the offsite dose consequences of this event would be expected to be insignificant and this event is deemed to have had minimal safety significance.

16. ,R rB ildin - r i n h m rV mBr ker Valves CSP-V-S, 6, and 9 are 24 inch diameter butterfly valves. These valves are in series with 24 inch check valves CSP-V-7, 8, and 10. The check valves were visually inspected in accordance with the Technical Specification requirements. These valves and the in series check valves were periodically leak tested in accordance with the 10CFR50, Appendix J test program. Valves CSP-V-5, 6, and 9 were also cycled through one complete cycle on a monthly basis in accordance with Technical Specification requirement 4.6.4.2.b.l.a. Based on the testing performed, it is concluded that valves CSP-V-S, 6, and 9 were capable of performing their intended safety function.

This event had no safety significance.

17. I I i nAc i nIn mn i nR n TimeT in indin 2 The affected relays, which do not have a time-delay feature, are electro-mechanical plunger-type (Agastat) or plate-type (HFA and HMA) with no dash pot or other dampening of the armature.

Degradation of these types of relays is typically evidenced by failure to function, rather than by degraded response times, The subject relays are verified to function on a periodic basis. The logic channels to the relays are tested on a quarterly basis as part of the CFT. However, the CFT does not test each relay individually. A LSFT is performed on a refueling outage basis and verifies proper operation of the subject relays. Response time testing of the subject model relays in other response time tests demonstrates reliable and repeatable performance of these model relays. This testing also demonstrated that the response time was small enough that the total logic response time would be within the Technical Specifications.

The need to perform response time testing is discussed in the plant Technical Specification Bases, which states:

"Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.... It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions. However, to enhance overall

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I LICENSEE EVENT REPORT Q4)

TEXT CONTINUATION AGILITY NAME (1) OOCKET NUMBER (2) LER NUMBER (B) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2

0 5 0 0 0 3 9 7 1 0 0 7 41 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME."

The portions of the isolation logic not previously response time tested were successfully tested. The combination of periodic functional testing of the subject relays, response time testing of similar relays, and industry and plant experience with these models of relays provides reasonable assurance that the relays were capable of performing their intended safety function. This event is deemed to have had no safety significance.

18. n rol R m Emer enc Fil ration ni HEPA Fil er In ec i n Although the April 9, 1985 performance of PPM 7.4.7.2.2, Revision 0, did not include all aspects of the WMA-FU-54A and WMA-FU-54B HEPA Filter visual inspection as specified in TSS Requirement 4.7.2.c.1, the procedure did include an in-place functional test that demonstrated filter unit integrity and proper system operation. As previously stated, the October 2, 1986 performance of Revision'2 to the procedure included additional visual inspection steps that satisfied the requirement, During the performance of Revision 2, no visual inspection discrepancies

'urveillance were identified for either WMA-FU-54A or WMA-FU-54B. Except for the addition of carbon adsorber, no filtration unit maintenance was performed during the eighteen month period between the Revision 0 and Revision 2 inspections. Therefore, if Revision 0 had included the visual inspection steps required to satisfy TSS Requirement 4.7.2.c.1, no visual inspection discrepancies would have been identified. Accordingly, this event was determined to have had no safety significance.

19. Iol i nAc i nIn men ionR neTimeT tin indin The affected relays, which do not have a time-delay feature, are electro-mechanical plunger-type (Agastat and Potter-Brumfield) or plate-type (HFA) with no dash pot or other dampening of the armature. Degradation of these types of relays is typically, evidenced by failure to function, rather than by degraded response times.

Each of the subject relays is functionally tested on at least an annual basis as part of the LSFT and response time testing of the relays and contacts was successfully completed on December 18, 1993.

Response time testing history for logic strings using the models of relays in question has confirmed the reliability and repeatability of these relays. There is no observed failure mode that has caused deterioration of the dropout time of these relays. Industry experience is that failure to function is the expected failure mode for these relays. Finally, the response time of the untested relays (approximately 110 milliseconds dropout time) is a small fraction of the Technical Specification required total time from initiation signal to valve closure (5 seconds or greater).

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L)CENSEE EVENT REPORT R)

TEXT CONTINUATION "AGILITY NAHE (1) DOCKET NUHBER (2) LER NUHBER (8 AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 i 0 42 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS A failure to isolate the Group 4 containment isolation valves affected by this testing deficiency within the time frame specified in the Technical Specifications would result in the potential for release from the primary to the secondary containment. The valves in this group that interface directly with the containment atmosphere are normally closed valves. The remaining valves are part of a closed system either inside or outside containment. A system failure would have to occur, concurrent with a LOCA, for a release to occur. In either case, the release would be to the secondary containment where the release would be processed through SGT.

Therefore, for the reasons stated above this event is deemed to have had no safety significance.

20. Ti T tin fE In m n i n indin 1 The relays that were not included in the response time testing are functionally tested during annual refueling outages. This functional testing provides periodic assurance that each circuit, including relays and contacts, will operate as required to perform its safety function. These relays are expected to have a response time of less than 0.5 seconds. The observed margins to the Technical Specification limits for the tested portions of the circuits ranged from 11 to 20.6 seconds. The additional time delay expected from the relays would result in a margin of at least 10 seconds. Due to the periodic functional testing performed on the relays and their small expected impact on the margin to Technical Specification limits, this event had no safety significance.

21. Re n e Tim T in fE In rumen ion indin 2 The relays that were not included in the response time testing are functionally tested during annual refueling outages. This functional testing provides periodic assurance that each circuit, including relays and contacts, will operate as required to perform its safety function. When these relays were tested on January 14, 1994, their combined response time was 100 milliseconds. Since the total HPCS response time of 22.4 seconds was within the Technical Specification limit of less than or equal to 27 seconds, this event had no safety significance.

22. ervice W er Valve P iti n V rifi i n All of the safety-related equipment room coolers that are serviced by SSW and HPCSSW contain flow switches in the cooler flow path. These flow switches provide Control Room alarms in the event of a low flow condition. Thus, if the cooler isolation valves identified as requiring position verification had been improperly positioned, the resulting low flow alarm condition would have been addressed in accordance with the appropriate alarm response procedure.

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LICENSEE EVENT REPORT IIR)

TEXT CONTINUATION AGILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 07 43 F 51 TITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS Seal water cooling from the SW system is only required for the RHR pumps utilized for shutdown cooling because it is the only mode of operation where the pump seals are exposed to water hot enough to damage them. Since RHR-P-2C cannot be used in the shutdown cooling mode by design, cooling from the SW system is not required. Thus, the RHR-P-2C seal water cooler and the associated isolation valves identified as requiring position verification do not perform an active safety function. Hence, the improper positioning of the pump seal water cooler isolation valves would not impact plant safety.

The CAC scrubber and aftercooler isolation valves are verified to be in their correct positions following each annual refueling outage in accordance with PPMs 2,3.3A, "CONTAINMENT ATMOSPHERIC CONTROL (DIV 1)," and 2.3.3B, "CONTAINMENTATMOSPHERIC CONTROL (DIV 2)." In addition, the scrubber valves are verified for correct position at least quarterly in accordance with PPM 7.4.0.5.14, "CAC VALVE OPERABILITY," and SW flow is verified through the aftercooler valves at least every six months in accordance with PPMs 7.4.6.6.1.1, uCAC-HR-1A PREHEATER OPERABILITY TEST," and 7.4.6.6.1.2, "CAC-HR-1B PREHEATER OPERABILITY TEST." Furthermore, these valves are checked periodically for proper position by the System Engineer during system walkdowns. Based on the valve position and flow verifications that were performed and the system redundancy, there is a high degree of confidence that the CAC system would have performed its safety function.

Therefore, for the reasons stated above this event is deemed to have had minimal safety significance.

23. HP In'n lv In mn i HPCS is a part of the Emergency Core Cooling System (ECCS). Its purpose is to supply water to the reactor vessel over a wide range of accident conditions. For small-break LOCAs that do not result in rapid reactor depressurization, the system is designed to maintain reactor water level. For large breaks, the system provides core spray cooling. The ECCS has built in redundancy that is comprised of HPCS, LPCS, Automatic Depressurization System (ADS), and the RHR - Low Pressure Coolant Injection (LPCI) Mode. Failure of HPCS is bounded by the ECCS single failure analysis.

The conditions reported in this item did not result in an actual failure of HPCS during an event where it was required to operate. In addition, performance of PPM 8.3.320 verified that the HPCS-V-4 instrumentation logic, not previously tested, would have performed as designed if called upon. Therefore, based on the HPCS instrumentation testing that was performed and the re'dundancy of the ECCS, there is a high degree of confidence that the ECCS would have performed its safety function. Accordingly, this event is deemed to have had minimal safety significance.

LICENSEE EVENT REPORT teR)

TEXT CONTlNUATION AGILITY HAHE (1) OOCKET HUHBER (2) LER HUHBER (8) AGE (3) ev. Ho.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 0 7 44 F 51 iTLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS imilar Ev n LER 91-013 reported a total of 12 items of noncompliance with WNP-2 Technical Specifications.

Following final submittal of the LER in August 1991, four additional LERs were submitted reporting similar events of noncompliance with Technical Specifications. LER 91-031 reported that IRM Control Rod Block Upscale and Downscale Trip surveillance procedures did not meet the CC surveillance requirements as defined by Technical Specifications. LER 92-004 reported that scram discharge volume scram and control rod block level instrumentation procedures did not meet the CFT sur veillance requirements as defined by Technical Specifications. LER 92-035 reported that the scram discharge volume vent and drain valves surveillance procedure did not accurately measure stroke time as required by Technical Specifications. LER 92-040 reported that the monthly High Pressure Core Spray (HPCS) diesel generator surveillance procedure did not measure start and load times as required by Technical Specifications.

The TSSIP was initiated to ensure compliance with WNP-2 Technical Specifications through improvement of the Technical Specification Surveillance Testing Program. This LER reports items relating to previous program deficiencies, and is a direct result of the TSSIP implementation.

"f T x R feren f

$ ys~em ~om m~nen Reactor. Protection System (RPS) JC Reactor Recirculation (RRC) Pump AD p RRC Circuit Breaker RPT-3A, 3B, 4A, 4B AD BKR Turbine Governor Valve TA V Turbine Throttle Valve TA V Main Turbine TA TRB Main Steam (MS) Pressure Switch 3A, 3B, SB PS 3C, 3D Intermediate Range Monitoring System (IRM) IG Source Range Monitoring System (SRM) IG Main Steam Isolation Valve (MSIV) SB Remote-Intake Radiation Monitor IL Main Steam System (MS) SB MS-PS-32A (B,C,D) SB PS WOA-RIS-31A(B), 32A(B) VH RIS Main Steam Safety Relief Valves (MSRV) MS V RPS-RLY-K3[A-H] JC RLY RPS-RLY-K14[A-H] JC RLY

LICLNSEE EVENT REPORT feR)

TEXT CONTIN UATlON FACILITY NANE (I) OOCKET NUMBER (2) LER NUNBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2.

0 5 0 0 0 3 9 7 010 45 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS EIIS Inf rm i n R f ~Eff R f

/@gem ~Com ~nn RPS-RLY-K10 JC Main Steam Isolation Valve Leakage SB Control System (MSLC)

Pressure Indicating Switch MSLC-PIS-60 SB PIS Standby Gas Treatment (SGT) BH MSLC Outboard Depressurization Valves SB ISV Isolation Groups 1,2,3,4,5,6,7,8&9 BD Average Power Range Monitor (APRM) Flow IG DET Biased Simulated Thermal Power - Upscale Logic APRMs E and F IG DET APRM Flow Biased Simulated Thermal IG Power - Upscale Main Steam Line Radiation Monitor (MSLRM) IL MON Main Condenser SD COND Mechanical Vacuum Pump SH P Reactor Building to Suppression Chamber BF VACB Vacuum Breakers (CSP-V-5,6&9)

Isolation Instrumentation Logic Relays JM RLY Control Room Emergency Filtration Units VI AHU WMA-FU-54A & B HEPA Filters VI FLT Standby Service Water (SSW) Valves BI V High Pressure Core Spray Service Water (HPCSSW) BI V Containment Atmospheric Control (CAC) System BB Residual Heat Removal (RHR) Pump (RHR-P-2C) BO P High Pressure Core Spray (HPCS) Relay BG RLY (HPCS-RLY-K13)

HPCS-V-4 BG 20 Equipment Room Coolers VA/VI/ CLR MK

'ICENSEE EVENT REPORT (l)

TEXT CONTINUATION AGILITY HAKE 1) OOCKET KUMBER 2) LER HUMBER (8) AGE (3) ear ev. Ho.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 46 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS ATTACHMENT 1 5+rEL'lg (nO-IC~~ al cll CWiir CII CIInCIC WCN5IC CA IiCCCIeil UH WTCC ICVCL 0 CMr~

II CMMklN LII 5II 5 %NQQ ICI llI OCIIC CI CAWQR <cc IIILcrlw5lwlc NOT RESPONSE TIME TESTED CMMSCll al +I CMMRlu CM~

ICI III cflhCO4 al iiI C+r tov al )II CMMNIIC OO% CI CC I50LCIIIII5IOW ICI CI I C~r~lcu ICI C1 I C~r~l5u IJI Cil ItgMr&CIC ~Mr~c C kr jul~

III CI I 04IC CI Cia-lr~<< CMr~u lccCIa 5CCu cilcld OI IlcrM Il'>CAN Cccu ccio I IIIII EXCERPT FROM DRAWING EWD-108E-001 "MISC EQPT SYS RELAY CABINET E-CP-RC/1 ISOLATION CONTROL RELAYS"

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I L[ORNSEE EVENT REPORT IGR)

TEXT CONTINUATION AGILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 07 47 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS A%I'ACHMENT2 K-12 RELAY DEENERGIZES ON A NEUTRON MON. TRIP.

RPS TRIP SYSTEM "A" RPS TRIP SYSTEM "B" A-1 A-2 B-1 B-2 12A 12C 12B 12D I

I 12E 12G 12H I I I

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t I I

I 1 I I K-12A K-12E K-12C K-12G K-12F K-12B K-12H K-12P IRM IRM IRM IRM IBI IPI I

~ I A I~ ~ I+II ~ ~ ~

IRM IRM IRM IRM I Ql IF I ~ ~ I HI I

~ IEI ~ ~ ~ ~

APRM APRM APRM PR APRM APRM "A" IIQI ~ ~ IB ~~ ~

Ifl ~ ~ IPI ~

SIMPLIFIED APRM/IRM RPS TRIP LOGIC

LlCCNSEE EVENT REPORT III')

TEXT CONTINUATION AGILITY NAHE (I) OOCKET NUNBER (2) LER NUMBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 010 48 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS ATTACHMENT3 ANSI N510 VISUAL INSPECTION REQUIREMENTS PPM 7.4.7.2.2 Rev 0 ANSI N510-1975(1980) No N/AN'.

Mounting Frames:

Continuous seal weld between members of frame X and between frame and housing Structural rigidity X

3. Squareness of members, flatness and condition of X component seating surfaces

4. Damage to frames X Filter Clamping Devices:

Proper adjustment X

2. Sufficient number of devices to produce 50 to 80%

compression

3. Individual clamping of filter (adsorber) cells X

4. ~ Proper condition of clamping devices

5. Adequate clearances between filter (adsorber) X elements to tighten clamping devices on all sides Full penetration in welds and freedom from crack X of clamping devices HEPA Filters:

Damage to filter media, case, case corners, on both faces of filters Damage to or improper seating of gaskets

3. Burns of media or case from cutting or welding on X both faces of filters 4, Excessive dirt loading X Not applicable to WMA-FU-54A/Bor need to be inspected only during initial acceptance test and/or aAer any system modification or repair.

LlCRNSEE EVENT REPORT +I)

TEXT CONTINUATION AGILITY NAHE (1) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 10 49 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IHPROVEHENT PROJECT IDENTIFICATION OF NONCONFORHING CONDITIONS ATTACHMENT3 ANSI N510 VISUALINSPECTION REQUIREMENTS PPM 7.4.7.2.2 Rev 0 ANSI N510-1975(1980) No N/A4 Prefilters:

Damage to media, case, or gaskets X

2. Excessive dirt loading Adsorbers:

Damage to cells, including burns from welding or X cutting operations in housing

2. Individual clamping

3. Condition of clamping devices X Adsorbers; (continued)

4. Condition of gaskets X Lighting:

Adequate for visual inspection of housing and X components

2. All lights lit (replace ifout) X

3. Penetration of mounting frame by power or control conduits

4. Vapor-tight globe, guard to protect globe from physical damage Housing:

Adequate space for personnel and equipment for X maintenance, testing

2. Reasonable access to housing X Not applicable to WMA-FU-54A/Bor need to be inspected only during initial acceptance test and/or after any system modification or repair.

A'> V 4

LltfNSEE EVENT REPORT h)

TEKl CONTINUATlON FACILITY NANE (I) OOCKET NUNBER (2) LER NUMBER (8) AGE (3) ear umber ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 0 0 7 50 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS ATTACHMENT3 ANSI N510 VISUALINSPECTION REQUIREMENTS PPM 7.4.7.2.2 Rev 0 ANSI NS10-1975(1980) No N/A4

3. Space adjacent to housing amenable to isolation as X a contamination zone, adequate space for temporary storage of clean and contaminated filters during filter change Doors of rigid construction with adequate seal X between door and casing and opening outward on negative-pressure housing

5. Adequate latches on doors, with provision for X opening from inside and outside of housing and provision for locking. Adequate sills

6. Adequate structural rigidity to resist undue flexure X

-reinforcing members on outside preferably

7. Access to upper tiers with permanent service platform at approximately 6-foot level

8. Adequate clearances for access between banks of X components with door on each side of each bank

9. No back-to-back installation components X

10. Proper location of tracer injection and sample .X points Adequate guards on fans located inside housing X

12. Housekeeping in and around housing X Housing: (continued)

13. Condition of flexible connection between housing X and fan external to housing

14. Fan-shaft seal X Not applicable to WMA-FU-54A/Bor need to be inspected only during initial acceptance test and/or aAer any system modification or repair.

r c >

(,'IChNSEE EVENT REPORT TEXT CONTINUATION Q)

FACILITY NAHE (I) OOCKET NUHBER (2) LER NUHBER (8) AGE (3) ev. No.

Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 I 0 51 F 51 ITLE (4)

TECHNICAL SPECIFICATION SURVEILLANCE IMPROVEMENT PROJECT IDENTIFICATION OF NONCONFORMING CONDITIONS ATTACHMENT3 ANSI N510 VISUAL INSPECTION REQUIREMENTS PPM 7.4.7.2.2 Rev 0 ANSI N510-1975(1980) No N/A~

15. Adequate dampers to prevent intake of air from X adjacent housing or plenum during test, and to prevent bypassing of system

16. Freedom from corridors, plenums, electrical X conduits, connections, plumbing drains or other conditions that could result in bypassing of the system Dampers:

Damage to or distortion of frame or blades Bent shafts, pivot pins, or operator linkages X 3 ~ Missing seats or blade edging 4, Condition of resilient seats or edging X Recheck:

Test equipment has been removed X

2. All openings have been sealed

3. No damage to components from the test operation X Not applicable to WMA-FU-54A/Bor need to be inspected only during initial acceptance test and/or after any system modification or repair.

ML17290B029

Text

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