ML17284A546

LER 88-023-01:on 880629,Tech Spec Violation of Secondary Containment to Outside Differential Pressure Occurred.Caused by Sys Configuration Error.Instrument Setpoint Change Request initiated.W/880930 Ltr

ML17284A546
Person / Time
Site:	Columbia
Issue date:	09/30/1988
From:	Powers C, Washington S WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-88-023, LER-88-23, NUDOCS 8810110371
Download: ML17284A546 (9)

Text

ACCEIZRATED DRIBUTION DEMONSKTION SYSI'EM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8810110371 DOC.DATE: 88/09/30 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION WASHINGTON,S.L. Washington Public Power Supply System POWERS,C.M. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION R

,I

SUBJECT:

LER 88-023-01:on 880629,Tech Spec violation of secondary I containment to outside differential pressure.

W/8 ltr. D DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES h

ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D

PD5 LA 1 1 PD5 PD 1 1 SAMWORTH,R 1 1 D

INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 AEOD/DOA 1 1 S AEOD/DS P/NAS 1 1 AEOD/DSP/ROAB 2 2 AEOD/DSP/TPAB 1 1 ARM/DCTS/DAB 1 1 DEDRO 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB 8H 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB 8D 1 1 NRR/DEST/RSB 8E 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/QAB 10 1 1 NRR/DOEA/EAB 11 1 1 NRR/DREP/RAB 10 1 1 NRR/DREP/RPB 10 NUDOCS-ABSTRACT RES TELFORD,J 2

1 1

2 1

1 QZQ- ~

NRR DRIS/SIB 9A RES/DSTR DEPY 02 1

1 1

1 1

1 RES/DSIR/EIB 1 1 RGN5 FILE 01 1 1 EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY,A 1 1 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC HARRIS,J 1 1 D NSIC MAYS,G 1 1 S

.I D

TOTAL NUMBER OF COPIES REQUIRED: LTTR 46 ENCL 45

NRC FornL355 ~

U.S. NUCLEAR REGULATORY COMMISSION (04)3)

APPROVED OMB No. 315001St LICENSEE EVENT REPORT (LER) EXPIRES: SI3lldd FACILITY NAME (1) DOCKET NUMBER (2) PAGE 3 Mashin ton Nuclear Plant - Unit 2 0 s 0 0 039 7 ~or06

'""'"Technical Specifscatson V)olatson o econ ary ontalnmen to u sl e -

1 eren sa Pressure Caused by Design Due to Programatic Errors EVENT DATE ISI LER NUMBER (5) REPORT DATE (7) OTHER FACILITIES INVOLVED (SI SEQUENTIAL to REVISION FACILITYNAMES DOCKET NUMBER(SI MONTH DAY YEAR YEAR NUMBER NS NUMBER MONTH DAY YEAR 0 5 0 0 0 0 6 29 88 8 0 2 3 01 0 9 8 8 0 5 0 0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR ()I (Cheep one or more of the follovnnPI (ll Moor. (5) 1 20A02(S) 20.C05(c) 73.71(SI 60,73(e l(2 I ov)

POWER 20A06( ~ ) ( I I (il 60,36(cllll 50.73( ~ l(2)(v) 73.71(cl 0 6 3 20.i0$ ( ~ )(1)lii) 50.35(cl(2) 50,73( ~ ) l2)(viil O'THER (Specify In ASNrect oefovvend In Tert, IYIIC Form

+~o 'gjivy 20,C05 (e I(1 ) liiI) 60.73( ~ l(2) li) 60,73( ~ l(2 l(viiil(A) 356'Al 20.C05 ( ~ ) ( I I Bv) 60.73( ~ l(2)(iil 50,73( ~ l(2)(viiil(B) 20.c05( ~ IV I(v) 50.73( ~ l(2) (iiil 50.73 4) (2 I I xl LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE Steven L. Mashin ton, Com liance En ineer 509 377 -2 080 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANVFAC. REPORTABLE MANVFAC. EPORTABLE " '5 rr4Fgg.

CAUSE SYSTEM COMPONENT SYSTEM COMPONENT TVRER TO NPRDS TURER To NPADS $ ,

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SUPPLEMENTAL REPOAT EXPECTED (ICI MONTH DAY YEAR EXPECTED SUBMISSION DATE (15)

YES III yN. COmPlete EXPECTED SUSIIIISSIOIY 04 TEI X NO ABSTRACT ILlmlt to Ie00 tpeced I e., epprovlmerely IIINen rlnple tpere ryPrwrl tren Ilnetl (id)

On June 29, 1988 it was determined that the plant configuration for sensing the Reactor Building (Secondary Containment) to outside differential pressure was incorrect and that the setpoint relied upon to maintain the technical specification value did not take into account instrument loop inaccuracies and drift. In addition, a rereview of event reportability determined that the event should have been reclassified as reportable per 50.72 (b)(2)(iii) and 50.73 (2)(2)(v) once the instrument setpoint calculation was completed and the margin required was determined to be greater than - .25. This event was previously evaluated as reportable per 50.73(a)(2)(i)(B). The 50.72 verbal notification report was made August 12, 1988 at 1546 hours0.0179 days <br />0.429 hours <br />0.00256 weeks <br />5.88253e-4 months <br />. In addition, the Architect/Engineer Burns

& Roe Inc. has been asked to evaluate this event for reportability per 10CFR Part 21.

A system configuration error caused by ambiguity in the Architect/Engineer (AE)(Burns &

Roe Inc.) functional description of signal select instruments in the Reactor Building differential pressure control circuit, and the lack of clarification on the the signal select configuration required, caused the low value select option to be enabled instead )

of the correct high value select option.

8810))03')

l<<SSOUSNTIAL rS A(VISION .Im" NUM 8 8 NUM 8 II Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 8 8 0 3 OF TEXT ///moro <<>>co d o/O/rorL o>> ////Ooo/ //RC /(orIIIm('c/ (ll) Event Descri tion On that June 29, 1988 a Plant configuration problem was discovered ahd under certain conditions the Reactor Building (Secondary Containment) to it was determined outside differential pressure could have unknowingly exceeded the Plant Technical Specification (3/4. 6. 5. l) requirement to maintain at least a - .25U of vacuum water gauge differential pressure. A system configuration error created a configuration condition that did not compensate for wind direction as described in the FSAR. In addition, the Reactor Building differential pressure setpoint could not be verified to account for instrument loop component inaccuracies and drift and therefore could have contributed to a non-conservative Reactor Building differential pressure. Both of these conditions have existed since Plant Startup. During normal operation, the Reactor Building Ventilation is provided by a constant supply and regulated exhaust system. Inaccuracies and drift of the control loop could have resulted in a positive building pressure during normal operation only. Since the Standby Gas Treatment system takes a suction on the building, pressure would be reduced upon initiation and have been maintained negative provided the accumulated dr ift did not exceed the controller setting. Excessive drift, if present would have caused a fan flow reduction until the fan tripped on low flow. At this point operator action would have taken place and the necessary controller setpoint adjustment made to sustain SGT operation. Current analysis for SGT response assumed initial building pressure to be atmospheric. 'he Reactor Building secondary containment pressure control system (part of the Reactor Building Heating and Ventilation System) utilizes eight differential pressure transmitters (REA-DPT-1A1 thru 1A4 and REA-DPT-lB1 thru lB4 ) (one on each side of the building for each redundant Reactor Building exhaust fan (REA-FN-lA 8 1B) to monitor building to atmospheric differential pressure. The signal select device for each exhaust fan controller (REA-LWS-1A and 1B) should select the least negative differential pressure signal; however, since Plant Startup the select devices have been configured to select the most negative differential pressure signal. The selection of the least negative pressure ensures that the required -.25" vacuum water gauge differential pressure is maintained on all Reactor Building walls, regardless of wind direction. The result of the configuration error is that the Reactor Building to outside differential pressure would not ~ compensate for a wind condition. The current WNP-2 design base does- not specify a design bases wind for pressure control maintenance on SGT system response. However, during some wind, portions of the Reactor Building could become p'ositive with respect to the outside pressure. This omission and other conditions than can affect SGT 8 Reactor Building HVAC System performance are currently being evaluated (reference corrective action to be taken 0'2). NAC FOAM SCSA o U.S.OPO:1088 0-834 538/455 (843) NRC Form 3SSA U.S. NUCLEAR REGULATORY COMMISSION (84)3) LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMS NO. 3)SOW)DE EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMSER (2) LER NUMSER (8) PAGE (3) VEAR gj~< SEOVENTIAL

i'ii REVISION NVMEER r K NVMEEII Washington Nuclear Plant - Unit 2 0 s 0 0 o 3 9 7 8 8 023 01 0 4oF 06 TEXT /// more EPece /e rtu/hler/ use SEE/'one//YRC Form 388r('s/ (I7)

Secondly, further evaluation determined that instrument loop component inaccuracies and drift were not taken into account when the Reactor Building pressure controller was set. The calculated instrument loop component inaccuracies is 1.756% . The span of control is 10 inches vacuum water gauge pressure (-3R to +7R) and the instrument loop component inaccuracy is +/- .175". Instrument drift was not specifically established by trend data and was estimated conservatively by doubling the instrument loop component inaccuracy ( Instrument drift +/- .35") until a design engineering setpoint calculation is complete. The Reactor Building to outside differential pressure was controlled by, manually setting the pressure controller (REA-DPIC-1A or 1B) for each Reactor Building exhaust fan so that a -0.25" vacuum water gauge differential pressure was maintained on the Reactor Building to outside differential pressure recorders. Therefore, when instrument loop component inaccuracies and instrument drift are considered, Reactor Building to outside differential pressure could have been greater than the technical specification if limit the postulated component inaccuracy and instrument drift occurred during normal operation. Immediate Corrective Action The signal select devices (REA-LWS-lA and 1B) were modified to select the least negative differential pressure signal for Reactor Building pressure control. An Instrument Setpoint Change Request ( ISCR) was initiated to change the Reactor Building to outside differential pressure control value from -0.25R to -0.6". This value will account for instrument loop component inaccuracies and drift in a conservative manner until a final setpoint calculation can be completed. With the approved ISCR the pressure controller was set to -0.6" on June 29, 1988. Further Evaluation and Corrective Action A. Further Evaluation

1. This event was Initia11y reported under the provisions of i 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the Plant's Technical Specifications. The actual dates and times when the Plant was outside technical specifications due to actual instrument drift cannot be determined. The Plant could have been outside the technical specification limit for some period of time since Plant Startup, December 1983.

A rereview of the applicable reportability criteria for this event was initiated following discussions with Region V of the Nuclear Regulatory Commission. During the investigations undertaken to prepare the original issue of this LER information became known which should have caused a re-evaluation of reportability per 10CFR50. 72. The rereview concluded the event should have been reported as a four hour report per criterion 10CFR50. 72(b)(2)(iii). The four hour report was made on August 12, 1988 at 1546 hours. ,This event is also reportable per criterion 10CFR50. 73(a)(2)(v). Burns and Roe Inc. has been requested to review this event for 10CFR Part 21 reportability. NRC IEORM 3EEA *U.S GPO:1988.0.824 838/ESS (883) NRC Form 355A V.S. NUCLEAR REOULATORY COMMISSION (94)3) LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMS NO. 3150-01(M EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (5) PACE (3) SEOUENT/AL dsp aEY/srON NI/Meea NVMeea Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 2 3 0 1 0 5 OF 0 6 TEXT /// more g>>oe /r reerr)ed, rrre sdrR/ono/ HRC form 3855 3/ I IT)

2. There were no Plant structures, systems, or components inoperable prior to this event that contributed to the event.

3. The instrument configuration error was caused by design drawings supplied by the Plant Architect/Engineer, Burns and Roe, Inc.. The signal select instruments were erroneously identified on design drawings as low value select, and in order to select the least negative signal they should have been identified as high value select. The root cause is attributed to the ambiguity of the method used by the Architect/Engineer to describe the function of the instruments and the lack of clarification as to which signal select configuration was required. A single jumper is relied on to configure the signal select instrument for high or low value select.

4. The failure to document calculations demonstrating the inclusion of instrument drift for the building pressure control setpoint is due to failure to apply the setpoint calculation program executed by the Architect/Engineer. The building pressure controller is a dial controller set by Licensed Plant Reactor Operators to maintain Reactor Building- to outside differential pressure at less than or equal to -.25R vacuum water gauge. There is no analytical value associated with this setpoint other than the Technical Specification value of -.25 inches vacuum water gauge and; therefore, any setpoint inaccuracies'r drift could cause a non-conservative pressure condition. The root cause of this deficiency is a programmatic error in the Architect/Engineer setpoint methodolgy program in that only devices that had a analytical limits were evaluated for instrument drift.

The cause of the missed 10CFR50.72 report is programatic in that no process was in place to cause a reportability rereview when new information becomes available. B. Corrective Action to be Taken

1. The design drawings will be revised to show the signal select devices as high value select instead of low value select.

2. An engineering study is. currently being performed to evaluate the secondary containment design bases.

3. Plant Technical Specifications will be reviewed to determine other technical specification limits that are maintained by control if there are circuits for which instrument loop inaccuracies and instrument drift may not have been calculated.

4. A letter has been sent to Burns and Roe Inc. to evaluate this event for reportability per 10CFR Part 21.

rrac FORM 35ea a U.S GPO:1988&824 538/455 19aS) NRC Form 358A U.S. NUCLEAR REGULATORY COMMISSION (983) LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMS NO. 3160M(04 EXPIRES: 8/3(/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMSER (5) PAGE I3) (mE SEOVENTIAL ~Qrr. REVISION NVMEER NVMSER Washington Nuclear Plant - Unit TEXT /// moro Epooo /E rooo)EIE oro o/E/ORE/ NRC Form 35SAS/ I (T) 2 <<<< o 3 9 0 2 3 0 1 p 60F

5. The WNP-2 reportability review program will be revised to include a process to re-evaluate reportability when new information becomes available.

Safety Si nificance There are no adverse consequences associated with this event. There were no Plant radiological events which would have caused unmonitored effluents in excess of allowable limits during this event period. If during this event period a Loss of Coolant Accident (LOCA) ocurred, there could have been unmonitored leakage through the Reactor Building in excess of that previously evaluated. The problem with instrument drift also could have caused a significant safety hazard since the maximum drift would have caused the Reactor Building to outside differential pressure to become positive during normal operating but would not have resulted in a significant release. The health and safety of the public or Plant personnel were not affected by this event. Similar Events None EIIS Information Text Reference EI IS Reference System Component Reactor Building NG Secondary Containment NG Reactor Buil ding Di fferenti al Pressure Control Circuit VA PDC Signal Select Instrument (Device)(REA-LWS-lA 8 1B) VA PDS Reactor Building Heating and Ventilation System VA Standby Gas Treatment System (SGT) BH Di fferential Pressure Transmitters (REA-DPT-1Al thru lA4) VA PDT Differential Pressure Transmitters (REA-DPT-1 Bl 'thru 1B4) VA PDT Reactor Building Exhaust Fan REA-FN-1A and 1B VA FAN Reactor Building Differential Pressure Recorders (REA-DPR-lA and 1B) VA PDR Pressure Controller REA-DPIC-1A and 1B VA PDC Pressure Controller SGT-DPIC-lAl, lA2, 1Bl, and 1B2 BH PDC NRC FORM 3MA R U.S.GPO:(988%-824 638/466 (983) ti WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washinglon 99352 Docket No. 50-397 September 30, 1988 Document Control Desk - USNRC Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO. 88-023-01

Dear Sir:

Transmitted herewith is Licensee Event Report No. 88-023-01 for the WNP-2 Plant.

The Revision to LER 88-023 (Revision 1) is submitted to correct errors made in the previous submittal and reclassify the condition as also reportable per 10CFR50.72. The errors primarily effected the manner in which the incorrect setpoint would have affected Standby Gas Treatment System response and also removed the discussion concerning the affects wind may have on building pressure control. The Supply System is currently evaluating various conditions that can affect Reactor. Building pressure control and upon conclusion of the evaluation will evaluate reportability.

This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportability, corrective action taken, and action taken to preclude recurrence.

Ver truly yours, C. Powers (M/D 927M)

WNP-2 Plant Manager CMP:lg

Enclosure:

Licensee Event Report No. 88-023-01 cc: Mr. John B. Martin, NRC - Region V Mr. C.J. Bosted, NRC Site (M/D 901A)

INPO Records Center - Atlanta, GA Ms. Dottie Sherman, ANI Mr. D.L. Williams, BPA (M/D 399)