ML17285A643

LER 89-027-00:on 890630,preliminary Engineering Evaluation Determined That Two Seismic Supports Missing on Each of Two PASS Containment Isolation Valves.Caused by Less than Adequate Work practices.W/890728 Ltr

ML17285A643
Person / Time
Site:	Columbia
Issue date:	07/28/1989
From:	Fuller R, Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-027, LER-89-27, NUDOCS 8908070002
Download: ML17285A643 (7)

Text

.,gc CEMRATZD Dl BUTION DEMONS ETIO'.i SYSTEM 1

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

I ACCESSION NBR:8908070002 DOC.DATE: 89/07/28 NOTARIZED: NO DOCKET N FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION FULLER,R.E. Washington Public Power Supply System POWERS,C.M. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 89-027-00:on 890630,inadequate seismic restraint of isolation valves could result in unisolatable breach of PC.

W/8 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR J ENCL J SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES h ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL 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 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 DEDRO 1 1 IRM/DCTS/DAB 1 1 NRR/DEST/ADE 8H 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/PEB 10 1 1 NRR/DOEA/EAB 11 1 1 /J) RPB 10 2 2 NUDOCS-ABSTRACT 1 1 1 1 RES/DSIR/EIB 1 1 RES/DSR/PRAB 1 1 RGN5 FILE 01 1 1 8

EXTERNAL: EGGG WILLIAMS,S 4 4 FORD BLDG HOY,A 1 1 L ST LOBBY WARD 1 1 LPDR 1 1 I NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 D S

NOTE TO ALL RIDS" R1XXPI1WIS'LEASE HELP US 'IQ REDUCE HASTE! CGMZACT THE DOQ3MEÃP (XWIK)L DESK, RDCN Pl-37 (EXT. 20079) KO EKXKBQXS KX3R MME HKH DZPHGBVZZCN LISTS FOR DOCUMEÃZS KRJ DGN~T NEED)

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 43 ENCL 42

~ r r 'Y, ~ S r, r 2 0 T ~ w,; v ." '0 ' iv , ' "v ~, v %' vv vr yv"

~ v ~ Y r v r,." r ~ v *Y 'C 1 v Lv,

WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352 Docket No. 50-397 July 28, 1989 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.89-027

Dear Sir:

Transmitted herewith is Licensee Event Report No.89-027 for the WNP-2 Plant.

This rep'ort 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.

Very truly yours, 5"/ttCo~wm C.M. Powers (M/D 927M)

WNP-2 Plant Manager CMP:lg

Enclosure:

Licensee Event Report No.89-027 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) 85'08070002 890728 PDR 'DOCK 05000397 8 PDC

NRC Form 355 U.S. NUCLEAR REOULATORY COMMISSION 04)3) APPAOVED OMB NO. 3)500104 1

EXPIRES: 5/31/SS LICENSEE EVENT REPORT ILER)

DOCKET NUMBER (2) PAO5 3)

FACILITY NAME (II Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 7 1 OF 0 TITLE (4) IN 0 ISOLATION VALVES COULD RESULT IN UNISOLATABLE BREACH OF PRIMARY CONTAINMENT CAUSED BY INADEQUATE WORK PRACTICES EVENT DATE (5) LER NUMBER (dl REPORT DATE LT) OTHER FACILITIES INVOLVED (SI MONTH OAY YEAR YEAR

'Cp~ SEQUENTIAL .re NUMddR RECON MONTH OAY YEAR FACILITYNAMES DOCKET NUMBERISI

,,c3 NUMBER Ak 0 5 0 0 0 0 30 898 9 27 0 0 0 7 8 8 9 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR (): (Check one or more ol the followlnp) (11)

OPERATINQ MODE (SI 20.402 Pol 20.405(el 50.73( ~ l(2) (irl 73.71Br) 20.405(el(1)(il 50.35(cl(1) 50.73(el(2)(r) 73.71(cl POWER LEVEL QTHER Ispeclly in Aorlrect 0 20.405(el)1)(dl 50M(c)(2) 50.73( ~ l(2) (villi(BI (rii) i>>low end In Text, HRC Form

"; ..oAP 20.405(el(1)(iiil 50.73(e I (2)(II 50.73(e I (2) (viii)(A) 366AI 20AOS(el(1) I)vl 50.73(e) (2) (5 I 50.73(el(2) 20.405(e (I I (el 50.73(el(2) (iiil 50,73( ~ ) (2)(el

, ..P.@4m...n I LICENSEE CONTACT FOR THIS LER (12I NAME TELEPHONE NUMBER AREA CODE R.E. Fuller Com liance En ineer 5 0 3 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANUFAC REPORTABLE MANUFAC. EPOATABLE CAUSE SYSTEM COMPONENT CAUSE SYSTEM COMPONENT TUAEA TURER TO NPADS TO NPRDS 1II+s)~g@p'3))rI)4%~

SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTED SUBMISSION DATE HSI YES IIIyet, complete EXPECTED SVBAIISSIOII DA TEI X NO ABSTRACT It.imit to Iepo n>>cer, l.e., epproxlmetely lllreen rlnple.rpece rypewriNen liner) (Idl On June 30, 1989 a pr el iminary engineering eval uati on determined that two sei smic suppor ts mi ssing on each of two Post Accident Sampling System (PASS) containment isolation valves, found by a Design Engineer on June 27, 1989, would probably result in failure of the pipe at its Primary Containment penetration during a Design Basis Earthquake (DBE). This would create an unisolatable breach of Primary Containment. The "as found" condition was discovered while the Design Engineer was performing a visual inspection of plant supports and while the Plant was at 3% power and in Mode 2 (Startup).

At 1650 hours on June 30, 1989 the Primary Containment Technical Specification action statement 3.6. 1.1 was entered and preparations were made to restore restraints to the required Plant configuration. At 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> when work was not completed on the restraints, a Plant shutdown was initiated. Primary Containment Technical Specification action statement was exited at 1843 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.012615e-4 months <br /> when the restraints were restored.

The root causes of the event are 1) less than adequate work practices to ensure the Plant configuration remains within design requirements, and 2) less than adequate training of project personnel to implement Plant modifications.

No further corrective actions were identified that would significantly minimize the recurrence of this condition in future Plant modifications. Current programs and procedures are considered adequate to ensure the plant configuration remains within the design requirements.

NRC Form 355 (04)3)

NRC Form'366A U rL NUCLEAR REQULATORY COMMISSION (943 l LICENSEE NT REPORT (LER) TEXT CONTINUA ON APPROV EO OMS NO. 3160-0104 EXPIRES: 8/31/68 FA ILITY NAME (ll OOCKET NUMBER (3) LER NUMBER (6) PAOE (3)

YEAR SEQUENTIAL ~R?Q REVISION NUMBER SS? NUMSER Washin ton Nuclear TEXT II/mere eeece/8 Plant - Unit rr /Ir/rerL Iree ~ I/cree/HRC lrerrrr 358481 (17) 2 osooo39 8 9 0 7 OO 2 orO 5 Abstract (cont'd)

Based on engineering judgement, there is no safety significance associated with this event because a qualitative assessment determined that a more rigorous stress analysis would indicate the pipe not fail from a Design Basis Earthquake (DBE).

Since the condition did not actually occur, this condition did not threaten the health and safety of the public and Plant personnel.

Plant Conditions a) Power Level - 25%

b) Plant Mode - 1 (Power Operation)

Event Descri tion On June 30, 1989 a preliminary engineering evaluation determined that two seismic supports missing on each of two Post Accident Sampling System (PASS) containment isolation valves, found by a Design Engineer on June 27, 1989, would probably result in failure of the pipe. at its Primary Containment penetration during a Design Basis Earthquake (DBE). This would create an unisolatable breach of Primary Containment.

The "as found" condition was discovered while the Design Engineer was performing a visual inspection of plant supports and while the Plant was at 3% power and in Mode 2 (Startup).

The two containment isolation valves (PSR-V-X82/1 5 PSR-V-X82/2) are installed in series on a >1-inch stainless steel penetration line (PI(l)-4S-X82d) leading to the Suppression Pool. The valves are located in the Reactor Building just outside of Primary Containment.

For valve PSR-V-X82/1, two angle iron braces, which provide vertical seismic restraint, were missing from the support. Two U-bolts were missing from the valve PSR-V-X82/2 support. The U-bolts provide thr ee directional seismic restraint.

A preliminary engineering evaluation was. performed on the 1-inch PASS line without the seismic supports on the two PASS valves. The evaluation conservatively used elastic modeling techniques, which does not allow for plastic deformation and results in higher stress values. The evaluation determined the highest stress in the pipe would occur at the penetration to Primary Containment and would be sufficient to fail the pipe.

NRC FORM 388A rU.S. CPOI 1988 520 589?00070 1903)

NRE Fenrir'366A'S~W U.S. NUCLEAR REOULATORY COMMISSION LICENSEE NT REPORT (LER) TEXT CONTINU ON APPROVEO OMS NO, 3160-0104 EXPIRES: 6/31/66 FA ILITY NAME 11) OOCKET NUMBER (2) LER NUMBER (6) PACE (3)

YEAR SEOUENTIAL ?rh/3 REVISION NUMBER NUMBER Mashin ton Nuclear Plant - Unit 2 o s o o o 97 89 0 7 0 0 3 oF0 5 TEXT ///mare <<>>ce /8 r)e/rerL u>> /6/aee/HRC Farm 3664'8/ l)T)

Immediate Corrective Action At 1650 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.27825e-4 months <br /> on June 30, 1989 the Primary Containment Technical Specification action statement 3.6.1.1 was entered and preparations made to restore restraints to the required Plant configuration. At 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> when work was not completed on the restraints, a Pl ant shutdown was initiated. The U-bol ts were instal 1 ed on PSR-V-X82/2 and the Primary Containment Technical Specification action statement exited at 1843 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.012615e-4 months <br />. An engineering evaluation had determined that the U-bolts would provide adequate seismic restraint to pr event failure of the pipe during a DBf. The angle iron braces for PSR-V-X82/1 were installed a short time later. A Plant shutdown had been planned for the evening of July 30, 1989 and the Plant was shutdown by a planned manual scram at 2323 hours0.0269 days <br />0.645 hours <br />0.00384 weeks <br />8.839015e-4 months <br />.

Further Evaluation and Corrective Action A. Further Evaluation

1. This event is reportable per 10 CFR 50.73(a)(2)(ii)(B) and 10 CFR

~

50.73(a)(2)(v)(C) 8 (D) as a condition that was outside of the 'design basis of the plant and alone could have prevented the fulfillment of the safety function of Primary Containment to control the release of

. radioactive material . and mitigate the consequences of an accident.

Preliminary engineering analyses determined the configuration could resul t in pipe failure at the containment penetration location following a Design Basis Earthquake (DBE), causing a breach of Primary Containment.

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

3. An ASNE Section XI plan (Plan No. 2-0112) was issued November 1, 1983 to reverse the flow direction of the two PASS valves. This required each valve to be removed from the line and reinstalled in reverse direction to provide a more reliable pressure seal on the Primary Containment side of the valve. The Section XI plan did not require supports or restraints to either be removed or reinstalled during performance of this task.

There is no documentation to indicate modification occurred later on these valves and/or associated lines that required removal of the restraints.

Therefore, it is assumed that the two PASS valves were left in the above described configuration following implementation of the Section XI plan 2-0112.

4. The root causes of the event are 1) less than adequate work practices to ensure the Plant configuration remains within design requirements, and 2) less than adequate training of project personnel to implement Plant modifications.

NRC SORM SBSA hU.S. CPOr (988 520 589r000)0 (633) ll .",' r...r ~ 'I' Ih', ~ hr..'. h,, 1'h M " ~ thr 0 hh' Era Vh'V( h',,h'hVI hII hI)r h, ""hhll 'rhrrIE M'CPII.', N>" '; ~, e Ih I, F1' 'I)I AN I 'TAY q',r' hh<<, h 0," ' ~ ' T., T)r

NRQ Form 3&&A U.S. NUCLEAR RECULATORY COMMISSION (943)

LICENSEE ENT REPORT (LER) TEXT CONTINU ON APPROVE'0 OMB NO. 3150-0)04 EXPIAESI B/31/BB FA ILITY NAME lll OOCKET NUMBER (1) LER NUMBER (5) PACE (3)

SEQUENTIAL REVISION YEAR gg NUMBER NUMBER n uclear Plant - Unit TEXT ///rROr& SProo /8 nq/Ir/IOIL rrro ////o&O/HRC Form 35//AB/117) 2 <<<< 3 9 7 8 9 0 2 7 0 0 0 4 oF a) A change notice was implemented to remove and reinstall supports for valve PSR-V-X82/7 on line PI (1)-4S-X82f, simil ar to work being performed on PSR-V-X82/1 at the same time. However, change notice documentation was not provided for the supports on PSR-V-X82/1 AND PSR-V-X82/2. The work practices were less than adequate to coordinate identification of Section XI plan deficiencies and to perform adequate post-modification inspections.

b) Project personnel lacked training to ensure Plant modifications were within the required design configuration. Either the personnel responsible for modification of line PI(l)-4S-X82d did not recognize removal of the supports was not authorized by the Section XI plan, and/or they were not aware that a change notice was required for work that was not specifically identified in the Section XI plan. In either case, project personnel were less than adequate'rai.net ensure that Plant modifications are clearly documented and approved to ensure compliance with design requirements.

5. Programs and procedure revisions have been implemented since the event occurrence (and not as a result of the occurrence) to provide added assurance that the Plant configuration remains within the required design monfiguration. As a result of the Safety System Functional Inspection

- (SSFI) and subsequent- to 1987, each Plant Technical System Engineer is required to perform a visual inspection of their assigned system prior to Plant startup from a major outage. The condition reported herein was missed during two previous inspections because it is located in a hard to reach ar ea and it was known that that par t of the system had had no recent ma'jor modifications. This condition was discovered while in the course of responding to a general Plant Management directive to all engineers to perform random inspections of the Plant'n areas of their expertise for conformance to the required Plant configuration.

In addition, the Plant Modification Request (PMR) procedure (PPM 1.4.1) was revised to require a post-modification review or inspection by the Design Engineer and the Plant Technical System Engineer of selected Plant modifications based upon a selection criteria. Also, the Plant procedure PPM 1.3.19 has been revised to require Area Coordinators to be trained to identify degradation and abnormalities of equipment. Furthermore, the Project Engineer of a Plant modification is responsible for ensuring that the Plant configuration remains within the design requirements.

B. Further Corrective Action No further corrective actions were identified that would significantly minimize the recurrence of this condition in future Plant modifications. However, current Plant programs and procedures are constantly being reviewed to identify areas where improvements can be made to provide increased confidence that the Plant configuration will remain within the design requirements. Also, current programs provide for continual review of the existing Plant configuration for compliance with the design requirements.

NRC FORM 3&&A AU.ST CPOr 1988-530-589r&00)0 (943)

NRE Farm 355A U.S. NUCLEAR REGULATORY COMMISSION (943)

LICENSEE NT REPORT (LER) TEXT CONTINU ON APPROVED OMS NO. 3150-0104 EXPIRES: 5/31/ES A

FAVILITYNAME (11 DOCKET NUMBER (2) LER NUMSER (5) PACE (3) v SEOUENTIAL REVISION YEAR P?< NUMBER  ? NUMBER Washington Nuclear Plant - Unit 2 o 5 o o o 97 89 0 7 0 0 5 oF0 5 TEXT ///more tPece /t reer?/rer/ Iree er///I/oat/HRC Frmrr 36//4't/ (12)

Safety Si nificance Based on engineering judgement, there is no safety significance associated with this event. A qualitative assessment determined that a more rigorous stress analysis with sophisticated plastic modeling techniques would indicate the PASS line (PI(l)-4S-X82d) would not fail from a Design Basis Earthquake (DBE). However, since the preliminary engineering analysis determined the pipe would fail, the safety significance of the postulated event (DBA, Earthquake and LOCA) is indeterminate because the effect of radionuclide release to the Reactor Building has not been analyzed.

Since the condition did not actually occur, this condition did not threaten the health and safety of the public and Plant personnel.

Similar Events None EIIS Information Text. Reference EIIS Reference System Component Sampling and Water guality System KN Sampling and'Water guality System (PSR-V-X82/1) KN I SV Sampling and Water (}uality System (PSR-V-X82/2) KN ISV Reactor Building NG Reactor Containment NH NRC FORM 3BBA AU.S. GPOr )988-f20-SBSIOOO)O (94)3)

.'I'?SLI c?Iar v?II?4 rr 3'va/Iv?r rr ' AN> tt?AA+rr I ) rat 'rJ 'Lvcr / ?Irer ~ Q p'rvc?,.Av't,r.' Irvi? r??rvc'I, a ( rare(crrrv' r rrrra A Iry>'v pa, rg Iwr r> rr rr;