ML20215C760
| ML20215C760 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/02/1983 |
| From: | Ladd G NUCOM, INC. |
| To: | |
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| ML20215C724 | List: |
| References | |
| NUDOCS 8612150386 | |
| Download: ML20215C760 (16) | |
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hCOM inc.
P.O. SOE 60
- GREAT BARRINGTON, MASSACHUSETTS 01230 = TEL. 413-528 2560 TECHNICAL INFORMATION RELATING TO ALTERNATE FOR MATERIALS FOR BWR - CONTROL ROD - B4C - TUBE CLADDING l
hDj2150386861205 p ADOCK 05000155 PDR l George T. Ladd 3/2/83
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CONTENTS Pg. 1 INTRODUCTORY Pg. 3 INCIDENTS OF t'RESS - ASSISTED IGCC IN REACTORS Pg. 8 REFERENCES Pg. 9 . BASIS FOR ALTERNATE TUBE CLADDING MATERIALS (SIEMARY) e l
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INTRODUCTION In reeme years, evidence has momtad that nonsasitized austenitic stainless steel fuel e1 mm is s4 ject to stress-assisted intergraular corrosion cr=ddy (IGCC) in both IMts ad PWits. 100C was idenriNd as the cause of the stainless steel fuel c1*g failures in VIMt and Deasden-lO); in the lacrosse INR(2); g, the PWR BR-3 Vulcain Experinunt (3); ed very recently, in Batch 8 of the Cann-ecticut Ymkee PWR(4) The failures in the developmental Incoloy-800 clad fuel irradiated in Big Ib& Point were also attributed.to.thisy.ers.cn(5) . In all ,
these cases, tensile stresses indaced by interaction of the UOg fuel with the cladding eventually led to clad crading fran the outside in with a characteristic intergrmular pattem. Similar cladding crads have also been obserwd in two experimental military PWR cores (0) (D' enploying different fuel ccnfigurations.
These consisted of c4ular(0) ed plate-type (7) fuel elenents; both types utilized a U0 stainless steel cennet bonded to austenitic stainless steel cladding. No 2
evidence was fomd in any of the exaninaticns of these irradiated pellet md cennet fuel types that " classical" cladding sensirirarion (i.e. , carbide precipitaticn at the grain bomdaries with resulting droniun depletion effects) was involved in the intergrmular crading mechanism.
Based cn the published description of the BWR ccntrol blade absorber t4e cracking phenonenon respcnsible for B4 C loss (8) (9) (10) , and consequently premature B-10 depleticn, it is believed that the failure of the umwrcial Type 304 stainless steel absorber t4e cladding represents another case in which stress-assisted IGCC of a ncnsensitized austenitic alloy occurred.
As a ccnsequence of this t4e cracking problan and the physical loss of B 4C due to leaching, the control blade lifethne has been reduce-1 by about 207.. The resulting premature discharge of the standard ocntrol blah represents a sestantial cost penalty to the utilities operating BWRs.
While mcst of the post-irradiaticn exaninaticn pwp- associated with in-reactor cracking of stainless steels did not address the mderlying cause of the cracking mechanism, a ccuprehensiw EC-EURATU4 Progran, inwstigating the fuel failures in VBWR and Dresden(I) , concluded that the presence of siliccn md phosphorus at grain boundries increases the intergranular corrosicn susceptibility of ncnsensitized ccmnercial austenitic alloys (11) (12) It was further shown that by decreasing 1
the level of these inpurities, alloys could be produced which were resistant to intergranular attack in test media (H) (12) and could be expected to impart a ccnparable resistance to IGCC in-reactor.
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s .r the principal remedy dtich we propose to insperate in m altamate BiR control
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blade design is the rep 1mrer=rit of the present absorber tee e1=Wg with stain-less steel more resistant to in-reactor ICXX: including alloy with controlled purity ;
capositions. Particular emphasis will be placed on restricting the silicon ad phc.picrus impurity levels. It is highly probable that this simple chage alone would allow restoration of the. original design life of the blade, i.e. , correspcmding to 427. B-10 depletion averaged over the top quarter.
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O r INCIENIS G' STIESS-ASSISIED ICCC IN REACIORS 5
As sunnarized by Dmcm(1) , meil the occurrence of IGCC in the Type 304 stainless steel e1*ng of experimental fuel rods irr= die *d in VIMR, it was generally beliewd that IGCC could only occur if the stainless steel was sensitized (i.e., fabricated such that chrumium carbide was precipitated in the grain bomdsries). After disdiarge in May 1962, however, examination of the failed VINR fuel unm-d that the cradcing was intergreular md that the el=Mng material was not sensitized (i.e. , optical ed electron microscopy did not find any carbide precipitates at the grain bomdaries).
Similar failures also occurred in experimental assenblies in the Dresden-1 reactor.__
In additicn to the absence of evidence of sensitization, metallographic examination of fuel rod el=Mng frun both VBWR md Dresden-1 rewaled other noteworthy, character-istic features:
- 1. Cracks in the cold-worked rods were icngitudinal while those in annealed rods were predmonately circunferential usually occurring at pellet interfaces.
- 2. 'Ihe intergranular crack pattem was initiated cn the fuel rod extemal surface and propagated inward throtgh the cladding.
- 3. 'Ihere was no evidence of metal defonnaticn detected at the crack tips.
l 4. 'Ibe fracture surfaces were corroded.
- 5. Higher operating stress levels appeared to be correlated with shorter times to failure. For exmple, the rod exposures at time of failure in VBWR raged frun 3,800 Iwd/MIU for a cold-swaged, cagacted 002 Powder fuel rod to approx-imately 10,000 Nd/MrU for 002 Pellet-type fuel rods clad with annealed ttbing ad with m initial pellet-to-cladding gap.
'Ihese characteristics were ccnsidered strmg evidence that a stress-assisted IGCC failure mechmism was operating in nmsensitized Type 304 stainless steel in the high purity water envir:xnent of a BWR. While such cracking had been induced in ncn-corrosiw chemical media, it was not sensitized austenitic stainless steels generally beliewd possible in high punty water tnless the material was sensitized.
Frcm the results of these early exaninations, it was tentatively ccncluded that, l generally, an incubaticn period in the reactor environment was necessary before the corrosicn mechmi.sm becane operatiw. 'Ihis suggested that a material change was induced after sune irradiation time resulting in increased susceptibility to IGCC.
More inportantly, there appeared to be a clear relationship between cracking suscept-To better l ibility and steess lewl as well as stress and strain concentraticns.
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mderstand the IGCC me&anism, further work was the cmducted as e AEC-EURAHH sponsored program (1) , whis led to the cxnclusion that aantrolling the inpurity level of the austanitic alloy is the key to improving resistman to IGCC. The results are discussed in detail later in this secticn.
More recently, it should be noted that IGCC of austenitic stainless steel cladding has been identified as the mode of fuel failure in several other INRs using pellet type fuel.
In 1973, leaking fuelelements.were.fotnd in Core 1.of the Lacrosse BWR which utilized a stadard Type 348 stainless steel for the fuel c1Mg. A hot cell exanination(2) was the cxncheted cn intact fuel rods representing the range of average bumtps- of '
failed rods,16,000 to 22,000 Nd/MIU, with the objective of finding incipient evidence of the failure mechmism. The progran succeeded'~ifi locating a~ craw in the lowest burnty intact rod initiated at the cladding outer surface adjacent to a pellet interface. The crack was intergraular md no evidence of grain botndary precipitates (sensitizaticn) was observed. It was difficult to detennine whether the crack directicn was 1crigitudinal or circunferential. It was concluded, however, that the characteristic of the crack were similar to the stress-assisted intergranular failures reported by Duncm(1) for BWR fuel clad with Type 304 stainless steel.
In a PWR case, the BR-3/Vulcain core experienced several cracked cold-worked Type 304 stainless steel clad fuel rods among a grotp that had exmeded the target peak pellet bumtp of 40,000 Nd/MIU(3) At this exposure level, pellet swelling leading to scxne i
cladding strain was expected. The description of the crack tiorphology (longitudinal was intergranular with no wall thmning) is similar to that reported for the cracked ,
VBWR md Dresden-1 cold-worked stainless steel clad rods.
In m interim report of the currently cngoing hot cell examination of the Cannecticut Yankee (PWR) Batch 8 fuel (4) the descripticn of the observed clad cradting pattem is strikmgly similar to that reported for the above BWR and PWR cases. Batch 8 was fabricated with cold-worked Tjpe-304 stainless steel and, bYsed cn coolat activity -
data, apparently started to leak when the average rod bumtp was approximately 22,500 Nd/MIU. Exaninaticn of failed rods showed that the cracks were lcngitudmal. Several incipient cracks, partially penetrating frcm the outside of the cladding, were obserwd in intact rods. Metallography indicated all cracks were intergrmular md optical microscopy rewaled no evidence of sensitizaticn. In ewry case, the cracks were associated with a fuel chip, wedged in the pellet-to-cladding gap, which acted as a stress ccncentrator.
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As part of the AEC-sponsored High Power Density Developmmt Project (5) , developmmtal fuel rods omtaining sugge-- ==<+ad arc-fused powder Ug clad in cold-worked Incoloy- l 800 were irra, Hared in Big Bodt Point. Cladding failures sind.lar to those observed in VBWR were visible anmg peripheral rods of four amanlies with m average exposure l of abott 7,000 Nd/MIU. Hot cell avn= rim of a c1m< Ming specimen revealed tin ;
typical intergraular cracking pattem initiated cn the clarMing extemal surface.
1hus, even me of the high strmgth stper-alloys was found susceptible to IGCC.
Austenitic stainless steel clad cerment frels have also apparently experienced stress-assisted IGT in a PWR environment. In the case of Di-3A mre(6) intergrmular cracks were fotnd in the Type 347 cladding of a tubular element initiated on the outside
~ surface. The investigators were puzzled by sudt a cracking pattern in a ncnsensitized material, but noted the similarity with the earlier BR fuel rod cracks. In mother case, the Type 304L clacMing of the SM-1 reactor plate-type elments(7) also exhibited extemally initiated intergranular cracks after readling a bumtp equivalent to 387.
U-235 depletim. _ _ .
The picture that energes fran all of the above incidents is that IGCC is the ultimate life-limiting mechmism for ccupcnents clad in ncnsensitized cmmercial austenitic alloys subject to increasing stresses md strains while operating in INR environments.
Although the fuel life in a PWR appears to be 1cnger than in a BWR, this may be a
( fmeticn of design differences as well as envirtrunental differences. An essential ccntributor to failure in both cases is a mechmism for producting high cladding stresses md strains. _ It is reascnable to conclude, therefore, that the similarly-described cracking of the Type 304 stainless steel absorber ttbes in the BWR ccntrol blades (9) (10) is another manifestaticn of the sane materials limit. In this case, the driving force for the cladding stresses md strains is B 4C swelling instead of fuel swelling. Further, the remedies developed to extend the life of austenitic l stainless steel fuel cladding in these mvirtnments should also be applicable to the absorber ttbe cladding.
l Basis for Proposed High Purity mdding Remedy As an outccme of a LS-EUPATW-spcnsored progran( } ( } , the renoval of specific inpurities fran camercial austenitic stainless steel cmpositions was identified as a rmedy for the stress-assisted IGCC problan encotntered in BWR's in the late 1960's.
In place of in-reactor tests, short-teun corrosicn tests in media known to cause intergranular corrosicn in ncnsensitized stainless steel were conducted. These provided a rapid means of obtaining data on the effect of netalluq;ical and cmp-osition variables cn the susceptibility to IGCC. The selected corrosion test nedia were:
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- 1. Boiling SN 1930 3
+ 4g Cr / liter of water.
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In both envirtransnes, specimens of specially prepared high purity mustenitic stain-less steels were far more resistet to intergranular attack the "--- J stainless steels. Additional tests with ocntrolled impurity additions fomd that silicon A ed phooncrus had a significant effect on the intergraular corrosion behavior'.
sharp degradation of resistance occurred when the silian level was raised above 0.27. md when the A.c.derus level exceeded 0.0157.. This apparent depen &ncy on inpurity level also gpeared applicable to high strength Incoloy-800, whis failed in Big Rock Point.(5)
As part of the progran, the grain bomdaries of the specimens were daracterized by microhar&ess ad polarization measurements, men inpurities were present, the grain bomdaries behaved differently frun the grains stagesting that the impurities were segregating at the grain bomdaries. 1his helped explain why irradiation tended to increase the IGCC susceptibility in the corrosion media as it appeared consistent with additional segregation being promoted in the irradiation envirrmment.
Corrosion tests were also ecnducted on an tnirradiated cladding sarple representing material that mexpectedly did not fail in the VBWR. The results demcnstrated that it was much less susceptible to IGCC than other ocnventional Type 304 stainless steel tubing with similar chemical ccmpositicn. Metallographic exaninaticn revealed, however, The I that this resistmt cladding had a very fine as-fabricated grain size (vl5u).
stperior behavior of the fine-grained material is mnsistent with lower grain bomdary ccncentraticns of inpurity solute accms for the sane initial inpurity ccntent.
Althotsh no in-reactor verification test was conducted, the results of the LE-EURAIQi progran indicated that inproved resistance to stress-assisted IGCC in amtenitic stain-less steel cladding can be achieved by restricting the siliccn and phosphorus inpurity It is noteworthy that this levels as indicated md requiring a fine grain size.
approach has, indeed, been adopted in the design of the lacrosse BWR replacemen which is now tnder irradiaticn(10) .
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COMPARISON OF CLADDING YlELD STRENGTH DATA 160 g g g g g g i i i 140 _
- INCONEL 625
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4- / O 348 120 C g 304 I - _
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3 304 L (ESTIMATED) 100 __ -
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4 Se 6 IRRAD TEMP 'F i
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RANGE OF
. MAX. FLUENCE '
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I i 304L7A/ i 20 T i i i i i i i i i i O 4 8 12 16 20 24 28 32 36 40 44 l
FLUENCE x 10 21 n/cm2gyjg,h)
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REFEIEN S
- 1. R. N. Dmcan, " Stainless Steel Failure Investigatim Progran, Final Stanary Report", EtBAEC, GEAP-5530, February 1968.
- 2. G. F. Rieger, " Post-Irradiation Evaluation of Fuel Rods Fram the Lacrosse Boiling Water Reactor:, PEDC-12633, June 1976 (prepared for Dairyland Power Cooperative) .
- 3. J. Storrer ad D. H. locke, "High Bumg Irradiatim Experience in Vulcain",
Nuclear Digineering Intemational, Vol.15, No: 165, February 1970, pp.93-99.~
- 4. V. Pastpathi ed R. W. Klingensmith, " Interim Report on Investigatim of Fuel Rod Failures in the Cannecticut Yankee Reactor, BCL-585-19, November 1980. _
- 5. C. L. Howard and R. N. Dmcan, "high Power Density Development Project, 'Nenty-Fifth and Twenty-Sixth Quarterly Program Reports, Jme-November 1966, GEAP-5405, December 1966.
- 6. J. B. Brown, Jr. , V. W. Storhok and J. E. Gates, "Postirradiation Examination of the PM-3A Type 1 Serial 2 Core", ANS Trmsactions, Vol.10, No. 2, November 1967, pp 668-9.
- 7. L. D. Schaffer, " Army Peactors Progran Progress Report", OFNL-3231, January 1962.
- 8. IE Bulletin No. 79-26, Fev. l. , U.S. NRC Office of Inspecticn and Enforcement, j
SSINS No. 6820, Accession No. 8006190042, Atgust 29, 1980.
- 9. K. W. Braymm md K. W. Cook, "Cantrol Blade Lifetime Evaluations Accomting for Potential loss of B C", NEDO-24232, Jmuary 1980.
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- 10. N. Eickelpasch, J. Mullauer, R. Seepolt and W. Spalthoff, " Experience with BWR Control Elements, Peaktortagtng, Berlin, March 1980.
- 11. J. S. Armijo, " Effects of Inpurity Additicns On the Intergranular Corrosion of High Purity Fe-Cr-Ni Austenitic Alloys", EURAEC, GEAP-5047, October 1966.
- 12. J. S. Armijo, "Intergranular Corrosicn of Nonsensitized Austenitic Stainless Steel".
Corrosicn , Vol. 24. No. 1. pp. 24-30, January 1968.
- 13. " Reactor Control Blade Evaluation, Millstone-I. Special Report". Docket 50245-244 July 23,1973.
- 14. C. W. Angle. Dairfl aid Per Cooperative , private co:muiication.
- 15. A, J . Anthmv . M. D. Grow md R. II . Yotng , "l'ir,h F.xposure G ut rol thi l'ine.ci '
I! '; l'aie nt 4. I 7.' . M? Issoni Oct ober Y) . 1974 I' u'e S
BASIS FOR ALTERNATE BLADE-TUBE CLADDING MATERIAL (SIMMARY)
Stress-assisted IGCC has been identified as the mechlinism for the cracking of nonsensitized commercial austenitic alloys used as fuel cladding in BWRs and PWRs.
Crack morphology in cold worked material (longitudinal, intergranular and initiated on outside) is similar to the description of control blade absorber tube cracks.
Tests in corrosion media (e.g., HNO3 + Cr+0) that promte intergranular attack showed that susceptibility is related to silicon and phosphorus impurity levels and grain size.
Indicated remedy currently in use in Lacrosse BWR fuel-high purity Type 348.
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.30 102 103 Pit 0$ PORUS CONCEllTRATION (ppe) y Effect of Phosphorus Additions on the Corrosion Resistance of Nonsen c sitt. zed High Purity ICa Cr,1,,o Ni, Balance.Te alloys in HNO3 + Cr*g E
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. - 9-14Ct- 14N1- HIGH PURITY l t1 THl55TUOY l A 0-160- 14Ni- 0.02C. 7 CORIOU, et all19)
A-190- 14Hi- 0.01C (Type -304) _
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O I I I IIIt! I I I I I I1!! I I I I I I t I! I i t tIei 103 104 102
$1LICOM CONCCHTRATION (ppm)
Effect of Silicon Additions on the Corrosion Resistance of Non-sensitized High Purity 14 Cr,14 Ni, Balance Fe Alloys in HNO 3 + Cr*8
CORROSION RATES IN HNO 3
+ Cr OF COMRCIE PURITY AND LOW Si AND P AUSTENITIC ALLOYS ALLOY CORROSION RATE Si,w/o P, w/o mg/cm 2 /hr US-EURATOM TESTS Type 304 (Comercial) 0.43 ' O.019 8.0 18-8 Alloy (HP) 0.00010 0.0014 0.87-16-12 Alloy (Comercial) 0.47' N0.020 4.4 16-12 Alloy (HP) 0.00050 0.002 ; 0.53 l
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Xncoloy-800 (comercial) 0.33 0.020 5.75 l
Incoloy-800 (HP) 0.005 0.010 0.75 l
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AUSTENITIC ALLOY IMPURITY SPECIFICATIONS NOMINAL COMPOSITION, Wt. % IMPURITY. MAX. Wt. %
TYPE Cr Ni Mn Fe . C 51 P 304 18.0-20.0 8.0-10.5 2.0 Ba1. 0.08 1.0 0.045 304L 18.0-20.0 8.0-12.0 2.0 Ba1. 0.03 1.0 0.045 347/348 17.0-19.0 9.0-13.0 2.0 Ba1. 0.08 1.0 0.045
- Incoloy-800 21 32.5 1.5 . Ba1. 0.05 1.0 --
- Inconel-625 21.5 60 0.2 2.5 0. 1'O 0.5 0.05
, Proposed High l Purity Alloy .. -- -- -- --
0.I 0.015 1
j Typical Lacrosse High Purity -- -- -- -- -- <0.1
<0.01 348 Cladding i
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ATTACHMENT 3 Consumers Power Company Big Rock Point Plant Docket 50-155 4
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BIG ROCK POINT HYBRID CONTROL ROD EVALUATION 4
December 5, 1986 4
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36 Pages I
TSB1286-0169-NLO4
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