ML20011F678

LER 89-026-00:on 891112,gaps in Piping Supports on Supply & Return Piping for Containment Coolers Identified.Caused by Inadequate Design Technique Used in Original Support Design. Shims Added Before Restart from outage.W/900301 Ltr

ML20011F678
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/01/1990
From:	Ewing E, Taylor L ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN039004, 2CAN39004, LER-89-026-01, LER-89-26-1, NUDOCS 9003070138
Download: ML20011F678 (4)

Text

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l s Arkansas Power & Ught Company l

, . 42 V cs!Captal l Lee Rack AR 72203 l Tel 501377 4000 j H

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March 1, 1990 l 2CAN639004  ;

i U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 l Washington, D. C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 2 #

Docket No. 50-368 License No. NPF-6 Licensee Event Report No. 50-368/89-026-00 Gentlemen: ,

In accordance with 10CFR50.73(a)(2)(ii)(B), attached is the subject report concerning reactor building cooler nozzles which could be subjected to  ;

unacceptable stresses during a seismic event due to excessive gaps in piping supports which resulted from an inadequate design technique. >

Very truly yours, I .)

. C. Ewihg r

i. General Manager, '

l Technical Support and Assessment l

l ECE/RHS/sgw l attachment cc: Regional Administrator l Region IV '

U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Records Center 1500 Circle 75 Parkway Atlanta, GA 30339-3064 9003070138 900301 h 413 N

l PDR ADOCK 05000368 d l

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NRC Fona 366 (9-83) U.S. Nuclear Regulatory C ssion Approved OMB No. 315 0104 Expires: 8/31/85 LICENSEE EVENT REPORT (L E R)

FACILITY NAME (1) Arkansas Nuclear One, Unit Two l DOCKET NUltER (2) lPAGE (3) 10151010101 31 61 Bill 0Fl013 TITLE (4) Reactor Building Cooler Nozzles Could be Subjected to Unacceptale 5 tresses During a Seisele Event Due to Excessive Gaps in Piping Supports Which Resulted From an Inadequate Design Technique

_ EVENT DATE (5) ,ER NUIWER (6) REPORT DATJ (7) 6 g I hiequentiell 1 Revision OTHER FACILITIE5 INVOLVED (8) l Month Day Year Year Number Month, Day l

i Number Year Facility Names Docket Nunber(s1 i i 0 5 0 0 0 11 1 11 2 81 9 81 9 --

' Of 21 6 --I 01 0 01 3 l 01 1 1 91 Of DPERA ING 0 5 0 0 0 lTH MODE J9) 5 5 RLFUMI 15 Summa ILU PUR5UANT D THE REQUIntntnis OF 10 CFR 5:

• PDWER (Check one or more of the followino) (11) -

20.402(b) l 20,405(c) 50.73(a)(2)(iv) l 73,71(b)

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LEVEll. 20.405(a)(1)(1)

-(10) 1 010101 1 20.405(a)(1)(11)

_t 50 36(c)(1)

[ 50.73(a)(2)(v) [l73.71(c)

_1 50.36(c)(2) 1 50.73(a)(2)(v11) Other (Specify in I l 20.40$(a)(1)(iii) l

~l 50.73(a)(2)(1) ~ l 50.73(a)(2)(v111)(A). ~l Abstract below and I 20.405(a)(1)(iv) 1

~ l 50.73(a)(2)(vi11)(B)l in Text, NRC Form Il[I20.405(a)(1)(v) l3l150.73(a)(2)11) 50.73 :a)(2) 111) l~l 50.73(a)(2)(x) l 366A)

Name LICENSEE CONTACT FOR THIS LER (12) l _ Telephone Number Area l Larry A. Taylor, Nuclear Safety and Licensing Specialist Code 51011 916141 13l11010 COMPLETE ONE LINE FOR EACH COMPONtKT FAILURE DE5CRIBED IN THIS REPORT (h3)

Reportable '

) Reportable Cause Systee Component Manufacturer to NPROS Cause System Component Manufacturer. to NPROS  !

I I l i I I I I I I l l 1 I l l I I 1 I I I l I i l I i SUPI LRtNT ret' ORT EXPECTED (14) EXPECTED Month Day (Year l~l Yes (if ves, complete Expected Submission Date) lifl No SUBMISSION L DATE (15) 11 I I AB5 TRACT (Limit to 1400 spaces, f.e., approximately fifteen single-space typewrf, ten lines) (L6)

On November 12, 1989, while performing an as built walkdown of the Service Water systes piping in the Reactor Building (RB) during refueling outage 2R7, Engineering personnel identified gaps in the piping supports on the supply and return piping for the containment coolers which were in excess of the original design gap allowables. An initial engineering review of the condition, which considered that the affected piping was a low temperature system (i.e., design temperature of 150 degrees), determined that the system remained operable pending further engineering evaluation. However, the identified niping support gap deficiencies were corrected by the addition of shims before restart from the outage.

the subsequent engineering evaluation, which was completed on January 30, 1990, determined that the ,

inlet nozzles on 2VCC-2A and the outlet nozzles on 2VCC-2B could have experienced higher loads than -

they were designed for if a design basis seistic event had occurred while the gaps existed. The cause of this condition was determined to be an inadequate design technique used in the original design of the piping supports. This technique provided support gaps to allow for free thermal growth during i

heatup, but failed position after to consider that the system piping could not be relied upon to return to its original cooldown.

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. i Form 1062.01B i NRC Fors 366A U.S. Nuclear Regulatory Commission 1 (9-83) Approved OMB No. 3150-0104 Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATIDN i FACILITY NAME (1) lDD RET NJEER (2) l JR NUMBER (61 l FAGE (3) [

l l l 1 5eouentiali i Revisioni '

Arkansas Nuclear One, Unit Two l l Year Number Number l 101510!0101 31 61 BI 86 9 --

01 21 6 --

01 of01210Fl013 TEXT (If mo n space is required, use additional NRC form 366A's) (17) j A. Plant Status 1 At the time of discovery of this condition, Arkansas Nuclear One, Unit Two (ANO-2) was in the cold ,

shutdown condition (Mode 5). Refueling outage 2R7 was in progress, f B. Event Description On November 12, 1989, while performing an as built walkdown of the Service Water system (SW) [BI) l piping in the Reactor Building (RB) in accordance with the Isometric Update Project, Engineering e personnel identified gaps in the piping supports on the supply and return piping for the containment  ;

coolers (2VCC-2A, B, C, and D) which were in excess of the original design pap allowables. An initial engineering review of the condition, which considered that the affacted piping was a low temperature system (i.e., design temperature of 150 degrees) and excessive thermal stresses were ,

not likely, determined that the system remained operable pending further engineering evaluation. l However, prior to restart from the outage, the support gap deficiencies were corrected to bring ,

them within original seismic design values. The subsequent engineering evalu; tion, which was completed on January 30, 1990, detersined that the inlet nozzles on 2VCC 2A and the outlet nozzles on 2VCC 28 could have experienced higher loads than design if a design balls seismic event had occurred while the gaps existed.

Three of the supports on the inlet piping for 2VCC-2A which had excessive gaps were located near each other and also close to the cooler inlet nozzles. The evaluation determined that the loads from these supports would most likely be transferred to the cooler inlet nozzles during a seismic event and that these loads could exceed acceptable levels.

There were six supports on the return line from 2VCC-2B which were identified as having excessive gaps. Three of these supports were located close to the cooler. The evaluation determined thet during a seismic event the loads from these supports would probably be transferred to the cooler ,

outlet nozzles, resultIng in stresses in excess of acceptable levels.

Considering the probable transfer of loads from the piping supports to the cooler nozzles during a seismic event, ANO determined (using engineering judgement) that these lines did not meet load allowables while the gaps existed. This judgement was made based on the assumption that the cooler nozzles would not have met load allowables if the supports with excessive gaps were removed from the analysis. It is possible that a detailed computer analysis could prove the system to be operable, however, it was conservatively declared inoperable.

C. Safety $1gnificance The Reactor Building Heat Removal Systems consist of the four 25 percent capacity containment coolers and two 50 percent capacity containment spray pumps and their associated spray headers.

These cooling systems are designed with sufficient redundancy so that any of the following combinations of equipment will provide adequate heat removal to attenuate the post accident temperature and pressure conditions which would be imposed upon the RB following a Loss of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB).

• All four containment coolers

• Both Reactor Building Spray systems

• Two of the four containment coolers and one Containment Spray System operating.

The significance of this condition is minimized by the fact that sufficient cooling system redundancy is available to adequately cool the RB following a LOCA or MSLB, even if 2VCC-2A and B were unavailable. In addition, the plant's safety analysis does not consider the occurrence of a LOCA or MSLB, which would require RB cooler operation, concurrent with a seismic event.

D. Root Cause The root cause of this condition was determined to be inadequate design of the piping supports.

A review of the original qualifying calculation for the SW system indicated that many of the piping supports were designed with specific gaps which would allow free thermal growth during heatup but would restrain the pipe during a seismic event. The original design attempted this by limiting gaps to a maximum of 1/8 inch and a minimum of 1/16 inch. Caps were specified on a

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~ . . s Fore 1062.018 .

NRC Fom 366A U.S. Nuclear Regulatory Commission (9 83) Approved Ole No. 3150-010*,

LICENSEE EVENT REPORT (LER) TEXT CONT!NUAT!DN ,

i FACILITY PURE (1) l DOCKET N M R (2) l LER NMR (6) l PAGE (3) '

l l 1 l%quentiall l Revision l Arkansas Nuclear One, Unit Two l l Year Number Number l ,

10151010101 31 61 81 81 9 --

01 21 6 --

01 Ol013IDF1013 T[KT (If more space is required, use additional NRC Fom 366A's) (17) particular side of the piping based on the predicted direction of themal movement. This design didn't consider that piping systems tend to ' creep' during their heatup and cooldown cycles and cannot be relied upon to return to the exact position in which they were installed. This technique of modeling piping systees to take credit for gaps in the pipe supports is unique to the piping system discussed in this report. ANO has not identified the use of this piping system modeling technique (i.e., taking credit for gaps in piping supports) in the design of other plant systems.

E. Basis for Reportability

$1nce the excessive piping support gaps resulted in a condition in which the cooler nozzles could have experienced loads greater than design during a seismic event, this event is considered  !

reportable pursuant to 10CFR50.73(a)(2)(11)(B) as a condition outside the design basis of the i plant. .

This condition was also reported in accordance with 20CFR50.72 on January 30, 1990.

F. Corrective Action The isometric update project, which identified the deficiencies discussed in this report, was initiated in 1987 to identify and resolve ex16 ting discrepancies between design drawings and the as-built condition of the plant. This project utilizes field walkdowns of piping systems as well as piping analyses reviews and has been proven effective in identifying suppcrt deficiencies.

All of the identified piping support gap deficiencies were corrected by the addition cf shiss [

before restart from the 2R7 outage.

Although the current configuration is judged to be operable, the design of this piping system will be evaluated using conventional design techniques to eliminate this unique model'ing approtch.

G Additional Information A similar event in which an inadequate modeling technique resulted in inadequate piping supports was reported in LER 50 313/88 030-00.

Energy Industry Identification System (E!!S) codes are indicated in the text as (xx). r I  !