ML20090A727

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Metallurgical Investigation of Cracking in Boric Acid Piping at Prairie Island 1, Summary of Results
ML20090A727
Person / Time
Site: Prairie Island 
Issue date: 02/08/1983
From: Rao G
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20090A716 List:
References
FOIA-84-257 NUDOCS 8407120093
Download: ML20090A727 (28)
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Text

.,

e ETALLURGICAL INVESTIGATION OF CPACKING IN THE BORIC ACID PIPING AT PRAIRIE ISLA'O 1 l

(SLFt'MY OF RESULTS)

NRC PRESENTATION, FEB. 8,1983 G. V. RA0 fhTALLURGICAL no NDE ANALYSIS NUCLEAR TECHNOLOGY DIVISION WESTINGHOUSE ELECTRIC CORPORATION l

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h2 8407120093 840521 B

Le 257 PDR

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SlffARY OF RESULTS THIS REPORTS SlttiARIZES THE PRELIMIf4ARY FINDIf4GS OF THE TETALLURGICAL INVESTIGATION OF CRACKING IN THE STA94 NIT EORIC ACID PIPIf4G LIADING FROM BORIC ACID STORAGE TN4K TO THE SAFETY INJECTION PLfiP AT THE PRAIRIE ISLAf4D UNIT 1 STATION. THE EXAMINATION ARE BASED ON A SIX INCH LONG 8-INCH DIAMETEP PIPING TAKEN FROM THE PIPE-TO-ELBOW WELD REGION OF THE BORIC ACID LINE. THE EVALUATIONS CONSISTED OF SURFACE EXNi! NATIONS, METALL0 GRAPHIC EXNilNATIONS, FRACT0 GRAPHIC EXN41 NATIONS, NO CHEMISTRY EVALUATIONS.

VISUAL No LOW POWER LIGHT MICROSCOPIC EXAMINATION OF THE INSIDE DIAMETER (ID)

NO OUTSIDE DIAMETER (0D) SURFACES OF THE PIPE SAMPLE IN THE AS-RECEIVED CONDITION SHOWED THAT YHE ID SURFACE OF THE PIPING IS COVERED WITH THICK BLACK OXIDE. THE OD SURFACE APPEARED RELATIVELY CLEAN.

REMOVAL OF THE ID SURFACE OXIDE BY LIGHT SNOING (DRY) REVEALED THE PRESENCE OF Nlt1EROUS CRACKS IN THE PIPE MATERIAL STARTING FROM NEAR THE WELD NO EXTEf0!NG 1 TO 1.5 INCHES INTO THE PIPE MATERIAL.

MINOR CRACKING WAS ALSO SEEN OCCASSIONALLY IN THE ELBOW PARTICULARLY NEAR THE WELD. THE ID SURFACE OF THE ELBOW REGION SHOWED SHALLOW PITTING. THE CRACKING IN THE PIPE APPEAR TO BE CIRCLtiFERENTIAL IN GENERAL. HOWEVER, SOME REGIONS WHERE CRACKS APPEAR TO BE AXIALLY ORIENTED WERE ALSO SEEN. BASED ON THE RESULTS OF SURFACE EXAMINATIONS, THE PIPE WAS SECTIONED NO SPECIMENS WERE TAKEN Olff FOR VARIOUS fETALL0 GRAPHIC, FRACT0 GRAPHIC AND CHEMISTRY EVALUATIONS.

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a A rusER OF CIRCUGEREtiTIAL NO AXI AL SECTIONS TAKEt1 FROM THE WELD REGION OF THE PIPING WERE POLISHED AtO EXAMINED fETALLOGRAPHICALLY BOTH IN THE AS-POLISHED AfD POLISHED AfD ETCHED C0tOITION, TO ESTABLISH THE DEPTH NO DISTRIBUTION OF CRACKS At0 THEIR RELATIONSHIP TO THE LOCAL MICROSTRUCTURE. THE RESULTS SHOWED THAT THE CRACKING WAS INITIATED ON THE ID SURFACE AND PROP 0 GATED RADIALLY OUTWARD.

CRACKING EXTENDED UP TO ONE TO TWO INCHES FROM THE WELD. SPECIENS POLISHED ON THE ID SURFACE SHOWED THAT THE CRACKING IS CIRCU4FERENTI AL IN GENERAL, ALTHOUGH AXIAL CRACKS WERE ALSO SEEN OCCASIONALLY. AT LOCATIONS NEAR THE 00 (TOP REGION)

CRACKING EXTEtOED FROM THE ID SURFACE TO ALL THE WAY UP TO THE OD SURFACE OF THE PIPING. THE CRACKS WERE PRIMARILY TRANSGRANULAR AND BRANCHED OUT IN NATURE ALTHOUGH OCCASIONALLY SOME DEGREE OF INTERGRANULARITY WAS ALSO SEEN. THIS CRACKING BEHAVIOR RESEMBLES TYPICAL CHLORIDE TYPE CRACKING IN 30l4 STAINLESS STEEL. THE MICROSTRUCTURE DID NOT REVEAL EVIDENCE OF APPRECIABLE SENSITIZATION. WET CHEMISTRY ANALYSIS OF THE PIPING AND ELBOW MATERIALS SHOWED THAT THEY CONFIRM TO THE TYPE 30l4 STAIN ESS STEEL SPECIFICATION.

X-RAY POWDER #4ALYSIS OF THE ID SURFACE OXIDE SHOWED EVIDENCE OF CHLORIDES (UP TO 70 PPM) IN THE OXIDE LAYER.

FRACT0 GRAPHIC EVALUATIONS AND CHEMISTRY EVALUATION OF CRACK DEPOSITS ARE CURRENTLY PROGRESSING.

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MAJOR TAS*/S t

o SURFACE EXAMINATIONS o

ETAl10 GRAPHIC EXAMINATIONS i

t-o FRACT0 GRAPHIC EXR11 NATIONS

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o CHEMISTRY EVALUATI0t!S 4

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f LEGEND - FIGURE-1 A1, A2... A10:

AXIAL SECTIONS FOR METALL0 GRAPHY CA, C2 AND C3:

CIRCUMFERENTIAL SECTIONS FOR METALL0 GRAPHY F1, F2 AND F3:

ID SURFACE METALL0 GRAPHY SPECIMENS FOR SENSITIZATION TEST S1, S2 FR1 AND FR2 SPECIMENS FOR FRACTOGRAPHY 5

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CHEMISTRY EVAL.UATIONS 1.

PIPE AND ELBOW MATERI AL CONFIRMED TO THE TYPE 304 SS SPECIFICATION 2.

X-RAY POWDER ANALYSIS RESULTS OF THE ID SURFACE OXIDE SHOWED A)

BASED ON X-RAY % PEAK INTENSITY CA TI CR FE NI MO ZR 0.35%

0.56 6.9 85.2 2.56 0.7 0.7 B)

CONTAMINANTS CHLORIDES 79 PPM SULPHATE 114 FLORIDES 18

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MARCH 27, l.975 A C I I. I T Y NOIIFICATION IlEM OR EVENT REGIONAI. AC T Ior, Y

M (<' ')

Au rop RKiHSAS POWER ANO INSPECTOR AT IJPD A TE O A I 8. Y R. IIRT OF 11/I8 AND EVAltJATE REPnRTS A LIGHT CilMP ANY SITE 3/21/75 11/19 /75 At'D 1/30/

, CONCERNING LEAKS FOLLOWLJP AT SITE.

NI8-1

( S T R E S S-CORRI15 IliN CRECKS1 IN THE REACTOR l

'N 50-313 Bill t.D I NG SPRAY SYSTEM SCHEDULF i n P I P I N(,.

I l

THE s.IrENSEE 46PORIE0 THAT THE PIPE SENSI-T I Z AT IllN MAY RE A R E StJL T OF M ANilF AC TilR ING TECHNIOllES HY SuiPCU, AND MAY I N V OI.V E A WHOLE CLASS fiF PIPE.

A L Sil, TRACFS OF l

CHl.11R I DE S H AVi BdEN OE i d C T Ell IN THE RH l

SPRAY ADDITIVE ( SOh lllM l H I flSill.F A T F FOR IllD I N 6 R E M!lV AI. ) SYSIEM LINES AND STilRAG6 TANK.

APEL IS INVFSTIGATING THF PRFSFNCE I1F THE CHLORIOES IN THE Aufli T I V6, ANO 1HF PRFSFNCF OF ADDITIVE IN THh RH SPRAY SYSTEM PIPING.

FOLLONIJP REPORTS WII.I. HE I S SilE D.

I NC R F A S EI) PLANT SilRVE ll.L ANCE ANO INSPEF, TION PRelGR AMS REMAIN IN EAFECT.

W OI. I N A POWER AND T E I. E C f1N,

6. 0 DAY AOR

-A LIC6NSEE OA Allo I T OF PIPING REVIEW R d Pf14 T AND F LIGH1 COMPANY 3/26/75 RFCOROS REVEALED THAT THREE SFCTillNS OF PFR MC 2515.

48 H SV I CK 2 I N S T AI.L ED 2-INCH REACTOR ORAIN PIPE,

8. 9 bl 50-324 FT. LONG, HAli NitT HFEN tit INSPFCTFO, AS RFOf f ! RED RY TH2 PSAR.

IHE PIPd W E l.O S HAD BFEN PT INSPECTED AND THE PIPE WAS HYOROSTATI' Alt.Y TES TED litJRING PRIMARY p

SYSTEM HYORO LICENSEE IS PERFORMING l

A OYNAMIC 4',tLYSIS AS PART OF THE I.0 CFR 50.59 f;VAltlATION.

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION y,

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WASHINGTON, D. C. 205G5 I

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MAY 161983 ilaul Wu ChMidarTe'chn~olisiSe~ tion MEMORANDUM FOR:

c Chemicai Engineering Branc, DE~"-

I THRU:

Ralph Meyer, Section Leader, Reactor Fuels Section 1

s Core Performance Branch, DSI FROM:

Michael Tokar, Reactor Fuels Section Core Perfomance Branch, DSI 3UBJECT:

PRAIRIE ISLAND SPENT FUEL POOL FUEL ASSEMBLY ~ DEGRADATION t

(TAC 49353) t At an April 20, 1983, work session 'of the group that was established to l

develop a staff position on the issue of the fuel assembly nozzle failure in the Prairie Island spent fuel pool (TAC 49353), you asked me to look into the safety aspects of a potential nozzle separation in an operating reactor core. The safety concern centered on the fact that at the present i

time there remains a possibility that the fuel assembly nozzle degradation in Prairie Island could be the result of in-reactor processes rather. than the spent fuel pool environment. Specifically, the question was "could a j

severence occur during reactor operation, and if so, what would be the k

safety issues? For example, would control rod insertion be impaired?" I have had some infomal discussions with Westinghouse lic(::ising personnel on this matter, and it is my understanding that W did a safety analysis j

for the Prairie Island licensee, Northern States Power Company, that 1

addressed the potential consequence of a fuel assembly nozzle separation in an operating core. That analysis, which to my knowldege has not been submitted to NRC, addressed control rod insertion and fuel coolability concerns.

In their analysis W reportedly detemined that alignment between the i

guide tube, nozzle and grids would be retained and that loose ~ parts would also not be a problem in the event of a separated nozzle. Alignment would be retained because (a) the Zircaloy thimble tubes will remain engaged I

within the sleeve portion on the top) nozzle, (b) the fuel rods will main-tain grid position and alignment, (c core pins will remain engaged within the top nozzle, (d) holddown springs will maintain axial positions, and (e) grids of adjacent fuel assemblies (containing thimble plugs) will provide additional lateral support. With regard to loose parts, W says that (a) j each portion of a failed sleeve would remain fimly attacTed to the grid and nozzle (also, each portion would be held on the Zircaloy thimble tube by the presence of the grid and nozzle), (b) any postulated grains of material freed by)intergranular corrosion will not affect control rod (RCC) l insertionand,(c effects of postulated debris are already bounded by the analysis of Salem grid damage. W thus has concluded that control rod insertion would not be affected Ty a separated fuel assembly nozzle.

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  • MAY 161983.

With regard to potential coolable geometry effects, W says that (a) there is no predicted increase in the potential physical deformation of a fuel assembly as a result of a separated nozzle, (b) lateral loads at a grid sleeve failure area are all within the strength of the Zircaloy thimble tubes acting (alone (i.e., grid impacts) alignment would be maintained with no flow blockage), (c) loads are primarily absorbed by center grids, and (d) postulated loose parts are not significant enough to affect coolable geometry.

Hence, W concludes that cool-able geometry will not be affected by a nozzle separation.

In sumary, I have at your request taken an infomal, preliminary look at the potential safety issues that might exist as a result of fuel assembly nozzle separations such as the one at Prairie Island (but occurring in-core, not in the spent fuel pool). Based on discussions with W, there appears (from a W analysis) to be no problem associated with prevention of control rod insertions or flou blockages. Thus, with regard to the order of magnitude of safety concern associated with a potential nozzle separation (i.e., thimble sleeve failure in-core), there would appear to be little concern, based on what I have heard infomally from Westinghouse.

Of course, I have not actually reviewed any analyses, but we can at least use this infomation as an indication of the degree of urgency that NRC should assign to this issue.

In my view, the infomation we have to date in no way enables us to recommend fuel assembly design or manufacturing and quality control changes, as questioned in your recent draft memorandum (copy attached), nor is there reason to think that any changes are needed at this point.

j ~b Jt.

M Michael Tokar, Reactor Fuels Section Core Perfomance Branch, DSI cc:

C. Berlinger V. Benaroya D. Dilanni C. McCracken 1

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  1. g ONITED STATES f*

NUCLEAR REGULATORY COMMISSION

,n y, g WASHWGTON, D. C. 20555

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JUN 2 ( 1983 MEM3RANDUM FOR:

Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:

C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data

SUBJECT:

POTENTIAL CONTAMINATION OF THE SPENT FUEL POOL AND PRIMARY REACTOR SYSTEM i

A recently completed Engineering Evaluation Report on the above subject is enclos ed.

Our evaluation concluded. that two separate events at the same plant site, which involved cracking of fuel assembly nozzles in the spent

, fuel pool and in, piping' from the boric acid storage tank to the safety injection system; are indicative of degradation that resulted from contami-nation of the systems.

Since the postulated sources of contamination were a contaminated batch of boric acid or resin intrusion due to the recycling system, the issue is potentially generic to PWR's.

.Some important aspects are that:

(1) this type of contamination was ndt anticipated, (2) the water chemistry monitoring programs for the spent fuel pool and the ' primary reactor system do not analyze for sulfates, and (3) an unsuspecting licensee could inadvertently place a plant in a condition that may result in significant degradation of safety related equipment.

We believe it would be prudent for NRR to consider establishment of requirements to include monitoring for sulfur contaminants in the water chemistry programs for the primary reactor system and the spent fuel pool.

l N

3k C.

He temes;Qr., Director Of for Analysis and Evaluation of Operational Data cc:

R. C. DeYoung, IE T. E. Murley, Region I J. P. O'Reilly, Region II J. G. Kepple.r, Region III J. T. Collins, Region IV J. B. Martin, Region V E. L. Jordan, IE

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