ML20010G700

Paper Entitled, Corrosion of Structural & Poison Matl in Spent Fuel Storage Pools, Presented at 810406-10 Intl Corrosion Forum in Toronto,Ontario,Canada

ML20010G700
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Issue date:	04/06/1981
From:	Czajkowski C, Protter S, Weeks J BROOKHAVEN NATIONAL LABORATORY
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[ / 7 PAPER NUMSER m RR@SPW81, e 1 The international Corrosion Forum Sponsored By the National Association of Corrosien Engineers / April 6-10, 1981 /

Sherston Centre. Toronto, Ontario, Canada.

! CORROSION OF STRUCTURAL AND POISON MATERIAL IN SPENT FUEL STORACE POOLS

• C. Czajkowski, J. R. Weeks and S. R. Protter

'Jepart=ent of Nuclear Energy and Reactor Division Brookhaven National Laborntory Upton, New York 11973 I. INTRODUCTION At the ti e cos: nuclear power plants now operating were conceived and constructed, part of the over all fuel cycle sche:e was that the spent nuclear fuel from these plants would be shipped to i fuel reprocessor where the plu-tonium and other fissionable caterial would be separated from the radioactive residue. Consequently, a typical nuclear power station is designed to have within its' .e t t e boundaries, facilities for storing only enough spent nuclear fuel to allow the short lived radioactivity to decay for a period of approxi-

=ately one year before this fuel would be shipped off-site. For .various reasons, it cay be necessary at some time during the operation of the reactor, to unload the entire core f rom a typical nuclear power plant. Consequently, the spent fuel storage pools were designed to hold the amount of fuel discharged in one year (or approxicately one/chird of a reactor core) plus the ent're contents of the reactor core. Recent government decisions, hcwever, have stopped the reprocessing of spent nuclear fuels, and until new govern =ent policies are developed re.rding long-term storage of the fuel, it has becoce necessary for the typica_ utility to store nuclear fuel on-site for =uch longer than a year.

Clearly this situation has required extensiu codification and expansion of the on-site fuel storage pools, including the installation in many cases of high density fuel storage racks containing nuclear poisons to prevent accidental criticality (l).

N Figure 1 shows a typical reactor s peitt fuel storage pool arrange =e nt showing the fuel transfer area between the storage pool and the reactor pit.

Because the water in the spent fuel storage pool mixes with the pri=ary coolant of the nuclear reactor during fuel unloading and loading processes, it is necessary to ::uintain the fuel pool coolant to reactor p ri=a ry coolant specifications. Tables 1 and 2 give the typical spent fuel storage pool chemistry and te=perature conditions at a boiling water reactor site where the coolant is high purity neutral water, and at a pressurized water reactor site where the coolant contains dissolved boric acid as a soluble neutron poison.

This boric acid in the spent fuel storage pool serves additionally to prevent accidental criticality of the fuel in storage, although no credit for this boric acid is given in the design of the pools.

• Work performed under the auscices of the United States Nuclear Regulatory Commission 8109220425 810914 PDR ADOCK 0500C461 _

A /ublication Right Cu;yrignt my the authen st where ec,pyrignt is e;;lics=le. Reproduced by tne Nattenal Asscciation of Corrosion Engineers with per Sten of tne autncqst NACE has been given first ngnts of puolication of this manusen;t. Requests for permissiort to puolish this 218310. Houston. Texas manuscrict in any form, in part or iri whole. mast be made in writing to NACE Publications Dept.. P.O. Box 77218. The mar'uscript has not yet Deen review ed by N ACE. and accordingly, the matettal presented and the views expressed are schly those of the au.hcrtsi and 3re not necessarily encorsed by the Assoc:Jt on.

Pnntad in USA

lkterials of cons t ruc t io n of sper" fuel storage pools are typically stainless steels for the pool liner and most equipment for. the cooling system, and stainless steels or alu=inus alloys for the fuel storage racks. Nuclear po: sons such as boral or a boron carbide organic resin are used between the fuel co=part=ents in some high density racks. The purposes of this paper are three-rold: first, to review the corrosion experience in general of the structural and nuclear poison caterials in spent fuel storage pools around the country; second, to describe our in-house experience at the spent fuel storage facilities at Brookhaven on the behavior of the nuclear poison boral and alu-inum and stainless steel structures; and third, to present the results obtained at Brookhaven on evaluating the causes cd the stress corrosion cracking that occurred in the Three Mile Is12nd - Unit I s p e nt. fuel storage pool heat exchangers.

II. STRUCTilRAL MA ERIALS IM POOLS The ovm_1 perfor=ance of structural =aterials in spent fuel storage pools to date has been excellent. These include stainless steels and aluminus alloys.

a) Stainless Steels

Stainless steels are used as the liners in spent fuel storare pools a l=o s t universally

n :his country, except for one or two of the ve ry early pools. Most typical spent fuel storage pools have unde 2 neath the liner a series of tracer channels in the =assive concrete foundation of the pool which serve as a leak detection system should any of the spent fuel pool coolant leak through the liner. These channels divert such leakage to a central header where it can be collected, monitored, and pu= ped either to a radioactfte coolant cleanup syste= or back into the pool, as the case requires. Although a few stainless steel fuel pool liners have developed leaks at the welds upon initial filling, these were attributed to defects in the original welds, rather than to any effect of the pool environ =ent. Intergranular stress corrosion cracking in the heat affected zone of a weld of a pool liner plate at San Onofre Unit 2 was discovered before the liner was inserted in the p- ol. This was probably caused by the da=p =arine atmosphere to which this line w . exposed while on storage at the San Cnof re site. The only instance of lea'e ':. of a pool liner developing af ter the pool had been in service for a period ot t'ce occurred at Salem Unit 1

' pool in the Spring of 1979. .a this instance, t a leakage was s=all, located

. with considerable dif ficult: ,

aad the area dam =ec of f by underwater divers and repaired by welding a plate on top of the leaking trea. No atte=pt was =ade to determine whether this leak was due to environ =enial interactions or mechanical failure.

i No stress corrosion to date has been detected on any of the fuel storage racks that have been in service, in some instances, for over ten years. How-ever, only visual inspections have been perforced on those racks that were removed from service for replacement with newer high density racks.

There nave, however, been several instances of stress corrosion cracki .g of stainless steel piping in the scent fuci pool cooling system, cost notably at Three Mile Island - Unit lW. At the request of the Nuclear Regulatory Concis s io n, 3rookhaven National Laboratory received' a portion of the piping f rom 4

ene Three Mile Island spent fuel storage pool cooling systeu containing a known crack to determine the cause of failure.

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P The portion of the piping we received is sketched in figure 2. The through-wall l intergranular stress corrosf.n crack, figure 3, occurred in an area adjacent to 3 a weld where extensive a=ounts of weld repair had been done and large amounts of l

sensitization occurred. ihe conclusions of cur investigation are sumcarized be-

- 1c a:

, 1. The pri=ary cause of the cracking appears to be intergranular stress corrosien cracking in the weld sensiti:ed heat affected :ene of a weld

} repair.

l 2. Although no definite corrc -ive species we re identified as the cause of l the cracking, the various traces of Cl on both the pipe inside surface and in the areas of the crack fracture faces deter:ined by EDX scans and the inside surface etched grain-boundaries is evidence of etching and possible contamination of the system by Cl ions perhaps during a cleaning or

pickling operation. The significance of Al, Si, Ca, P and K is not currently obvious.

3. The significance of the wideiy dispersed MnS striager-like inclusions l as pit nucleation sites is inconclusive. Also, i=portance of the widely dispersed sulfur traces throughc.. the fracture face is yet to be de-termined, but see=ingly requires further investigation into the effects of

, both S and MnS in the spent fuel environ =ent.

4 The results of Electrochemical Potentiokinetic Reactivation Analysis (EPR) did show at least one area of the pipe's cross section had been sensiti:ed significantly by the welding process which is a reasonable indication of the pipe's precracked condition.

5. It is evident from the Constant Extension Strain Race Tests (CERT) re-3 suits (corroboratad by the oxalic acid ' etched sierostructures) that the Type 304 SS =aterial, with its high (0.0L*.) carbon content and complex l thermal history (2 repairs) was seve rely sensitized at va rious locations I about its girth. This degree of sensitization was sufficient to 'cause

! intergranular-type fracture of this =sterial during CERT testing in air at see-l.

a strain rate of 8,5 x 10~0 This degree- of sensitization coupled witn the residual stresses of welding and the possible contamina-tion of Cl ions was see=ingly sufficient to crack the piping by an inter-granular stress corrosion cracking eachanism.

6. A reasonable explanation of the intergranular-type cracking observed in J the CERT tests (in air) can be made hat is similar to that po4tulated by

] Hippsley, et al.(3) It is possible that at some critical strain, the MnS/ tic inclusions start to decohere from~the catrix alloy. Since the 304SS

, was significantly sensitized (g rain boundary carbide formation) this-

, decohesion (ductile in nature) started to apply a shear stress across the more brittla carbides in the grain boundaries causing them - to " slide" across one another thereby resulting in the intergranular-type f ailure with the facets appearing in a dimpled rupture code. This theory 'is cocewhat j

substantiated- by the fact that the intergranularly f ailed specimens tested j in solution or air, all exhibited'a similar degree of elongation (strain;

of almost 50% and fell within a reasonable scatter band for tensile stress 66-32 ksi, which is quite similar to the specimens which f ailed in a ourely

ductile =anner.

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163/3

4 It is obvious from these investigations that, even in the relatively benign environ =ents and low te=peratures at which the spent fuel storage pool operates, 3 care should be taken in =anufacturing or fabricating stainless steel components i to' prevent undue sensitization of the stainless steel, high stresses associated with weld repair areas, and conta=aination of the pool with har f ul impurities such as sulfur co= pounds or chloride co= pounds.

{ b) Type 17-4 PM STAINLESS STEELS

, Type 17-4 PH stainless steels are used in an number of spent fuel storage pools as seismic restraints in order to provide stability to the racks.

In this application, the normal load on this =aterial is well below yield, and produced only by nut tightening or bo.l. :ing stretses. Mcwever, the- loads on these bolts =ay approach their design strength durin t a design basis earthquake

, for a short period ef'ti=e. Consequently, there is considerable interest in the l long-term perfor:ance of this =aterial in a spent fuel s:orage pool environment.

Further, there is a te=ptation on the part of engineers to =ake the =aterial as strong as possible by heat treating it to the H-900 condition.

) 6he literature contains =any references to stress corrosion cracking of Type 17-4 PH in the H-900 condi: ion. However, almos: exclusively these ref-4 erences are to exposures at te=peratures of 149*C (300*F) or higher. In a

=arine a:=osphere, this =aterial has failed ir the H-900 condition at te=pera-tures up to 52*C (125*F) but not when i==ersed in sea water at te=peratures of 27-30*C (80-85*F). (') Clearly the situation is so=ewhat clouded as to whether

or not this =aterial in this heat creatmen

=ight crack over long periods of ti=e in spent fuel storage pool environ =ents. Most utilities are now reco=-

=cading the H-1025 or the H-1100 heat trea t=en t for this =aterial, under which condition it has operated satisfoetorily in =any applications at te=peratures 3

and pH quite consistent with that present in both BWR and PWR spent fuel storage

! pools. In the High Flux Bea: Reactor at Brookhaien it b , performed well in a l pH 5.0, high oxygen environ =ent at 21*C-68*C (70-155* for over 15 years, j where this =aterial is used in the control rod drive pinions.

l The second potential proble= with the use of this =aterial in spent fuel j storage pools comes from reac: ions involving the heat treatment scale usually present on co=pon2nts of 17-4 PH following the H-1100 heat treat =ent. The

=anufacturers of the =aterial reco==end that the heat t rea t=ent scale be re=oved by chemical cleaning or by =echanical =eans bafore the =aterials are used in a water enviran=ent. Certainly any residual scale on the =aterial could poten-tially lac 4 to pitting in the long-term in an oxygenatec environ =ent. We ex-pesed several speci= ens of this =aterial with the heat treatment scale in:act to the spent fuel storage pool at the High Flux Bea= Reactor at Brockbaven to de-termine the long-term behavior of the heat treatment scale. Figure'4 shows that in six =onths the surface of the 17-4 PH exposed with the heat treatment scale intact was heavily coated with a brown rus t, whereas, the surface of the spect-l = ens cleaned af ter heat treatment re=ained bright and shiny. In neither case,

! however, was there any =easurable general corrosion, although there was possibly c so=e slight pitting underneath the rust on one specimen. Since this cust-like i

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=a:erial would contribute to the crud deposits on the fl oo r of the spent fuel storage pool and sight in some areas lead to localiced attack on the 17-4 PH, renoval of :he hea: trea: ent scale f raa any cocponents of this caterial before their insertica into a spent fuel storage pool would see: prudent.

c) Aluminum Stainless Steel Junctions Aluminum cc=pon9nts have stood up well in spent fuel storage pools, particularly in those associated with boiling wa re r reactors in which tha environment is basically neutral, high purity water. However, in a boric acid pool, or even in a neutral water pool there is always a potential f,e pitting of the aluminua a a point of galvanic contact between aluminum and stainless steel. At so=e reactor si:es, nota 51y Ver:on: Yankee (5), insulating feet have been put on the alu=inum racks wher_ they sit on the stainless steel floor of the spent fuel storage pool to minimize this pitting attack. In the spent fuel pool associated with the High Flux Beas Reactor at 3:ookhaven, we have aluminum racks and a concrete floor in the pool tver which 5 laid a '/l." stainless steel sheet to prevent da age :o the concrete. T'pical

/ water chemistry in the pool is similar to that in Table 1. To conitor the pitting reactions that =ight be occurring between aluminum racks and the stainless steel floor vf the reactor pool, a series of corrosion coupons have been kept icte rsed in the pool for several years. Figure 5 shows the exten: of pitting of the a lu=inu= surface following removal fr n the pocl after six :onths exposure. The pitting occurred only around the edges of the point of contact between the a luminum and the stainless steel. The inner portion of the sandwich, while slightly discolored, shows no pitting attack, which is probably due to the reduced availability of oxyge n in this arer. In February, 1979, while these speci= ens were present in the pool and whila the =ixed bed in the pool water purification system was being regenerated, a faulty shut-off valve allowed so=e of :he 5 sulfuric acid regenera:ing solution :o leak in:o the pool. The pH dropped to 3.6 and the resistivi:y of the wa:er to 1,000 oh=-es. Approxicately a five fold increase in the aluminua concentration of the water was observed. After a week of operating the decineralicers and the purification, the pH had risen to 5.2 and the resis:ivity to 210,000 oh -es. The pH had returned to the normal range within 11 days. The sa:ples were recoved, and the three pits shown in figure 5 were reexamined. No new pitting was observed as a resuln of this incidant, although the 0,ree pi:s shown in the fig 2re did grew slightly. Subsequently, several of the aluminus racks were re=oved f rom the pool (as part of our in-house replace-sent of the fuel s to rage system with high density fuel storage racks) and ex-acined. Again, there was pitting of roughly the sane =agnitude shown on the photograph at several places on the feet of the racks, where they were in con-tact with the stainless steel liner af the pool. However, ir no way did this minor pitting affect the strue: ural .ntegrity of the racks over the period of 15 years since they were installed in the pool and the pool filled.

Several of the boiling water reactors have replaced their aluminum racks as part of the densification progra=, and no significant corrosion has been ob-served where the aluminum racks rested on the stainless steel floor of the pool.

163/5

4 I

l III. POISON MATERIALS'I" POOLS i

The extensive use of nuclear' poisons, especially boral la spent fuel j

s torage racks, has been brought abou: by the need for increasing the storage s

capacity of the va rious nuclear power plant sites. The literature on boral, dating back over a peried of 20 years, states that its corrosion resistance is 4

the same as that of the aluminua cladding, proviced the 34 C-Al cer et in the core of the boral is not exposed to the coolant. Figure 6 shows a sche =atic of boral. As can be seen, it consists of an aluminua boron carbide cernet clad on both sides and along the- edges with alu=inuc, typically either 1100 or 6061 i alloy. Because of chis precaution in the literature, (which experience has l shown to be unnecessary) all boral exposed in the co=cercial nuclear pools was originally designed to be contained in sealed cc:part=ents between the fueled storage cells within the racks to preven: access of watt .o the boral.

a) Compatibility of Boral with tiate r l At 3rookhaven, boral has been exposed in the fuel storage area of the j Brookhaven Medical Research Reactor to t.ater of essentially the same quality as that in a BWR pool since January, 1959. In the Sum =er of 1978, because of the i

int.e re s t in the use of :his =aterial in spent fuel storage pools, several punch-ings were taken fros the boral plates in this reactor as shown in figure 7, and analyzed =etallographically in our labora:ory, and for boron at the University-of Michigan by neu:ron at:cnuation. Figure 8 shows that the neutron attenuation data for all six speci: ens agreed wi:hin 20%, which is probably within analy-tical error for such s=all saeples. Speci:en r/5 was analyzed chemically to contain 41. 3% 43 C in the core, which is in the upper range of boron concen-tration for caterial produced in the 1950's. Additional portions of these

punchings were exa

ined by General Electric in San Jose. It is clear from these examinations that there has been no loss of boron carbide to the coolant of the Brookhaven Medical Research Reactor over the period of more than 19 years that

his bo rs ' was exposed to the coolant. This observation gives considerable 7 assurance that, should spent fuel pool water leak into the cavities containing boral, there will be no concocitan: loss of the boron carbide from the cavity. ,

J b) Swelling of Poison Racks At severcJ nuclear power stacions, when racks were inserted into the t pool coataining boral sealed in stainless steel cavities, leakage of water into

. the stainless steel cavity .containing the boral led to a for=ation of hydrogen i

by the initial passivation corrosion of aluminum. Production of hydrogen duuag this period, of course, has been kncwn far a long tiae. Figure 9 gives some i

data ~ published by Craley and Ruther in 1953(6) showiag that the rount of cor-rosion in ter:s of ce:al lost ove r the first week or so af ter aluminum is ex-l posed to wa te r, can equal as =uch as 20 mg/de2 In terms of hydrogen produced by this reaction, this corrosion could produce as =uch as 7500 cc of hydrogen at standard temperature pressure conditions per tube of :he spent fuel storage codule. In as much as :he void space around the boral, where it is sealed into

the tube, is of the order of 100-150 c=3, this quant.'

y of hydrogen is quite sufficient to produce the 5-6 psi pressure needed to bulge the thin stainless 163/6

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steel cladding : hat seals the boral in place. The implication of this phenome-non will be discussed later in this session by Mr. VincentW. Since it is icpossible to ensure that all of the boral will be pe rma nen tly sealed, it is

desirable to vent these cavities intentionally so that the hydrogen produced can be released to the environ =ent withou t aliowing swelling of t h,

racks. This solution to the swelling problem, however,. places greater emphasis on the possible corrosion that =ight occur where aluminum and stainless steel noce in o contact in the spent fuel coolant. This is of particular concern in the poots j adjacent :o PWR's, where the-coolant contains boric acid at a pH of around 5, as

shown in table 2. Based on our experience with coupons, described above; the pitting will pri=arily occur near the points where the boral-containing cavities

are vented. Since these cavities are as =uca as 20 feet long, significant loss of aluminum from the pitting is unlikely. Further, f rom the BMRR experience,

and from studies of this pitting corrosion under e xtrece conditions of low pH i and elevated temperature, it is ve ry unlikely that the boron carbide .itself l would be lost -by the pit

ing cerrosion. Meas-tecents of pitting currents in this couple have been shewn to be strongly affected by :he amount of oxygen 4 available at the site where the active pit:ing is occurring. Fu r the r, the

alu=inum hydroxides produced by the pitting corrosion would be expected to re=ain in the cavi

1es and hold the 3 4C particles in place.

4 Most of the u:ilities utilizing the vented boral racks have committed themselves to a corrosion =onitoring system consisting of both small coupons that are vented to the environment and full length tubes containing the boralW. Since small coupons will have =uch greater access to oxygen at the l point where :he alumiau= and the stainlen steel are ;.n contact in the environ-

! ment, we think they will give a conservative =easure of the state of the nuclear poison throughout the storage racks.

IV. CONCLUSIONS a) The perior:ance of structural and poison catorial in spent fuel storage i pools to date has been excellent.

i l b) Intergranular stress corrosion of sensi:1:ed s:cialess steel can occur in these environments if the =aterial is heavily sensi:12.ed and heavily I stressed, as =igh: occur in areas wi:h extensive weld repairs.

c) Accidental contamination of :he pool with caterials such as chloride or reduced for=s of sulfur could lead to initiation cf stress corrosion cracking of 1

j such =ctarials. Clearly careful quality tinstruction techniques are recom= ended.

Lo d) Soral can be c.< posed to nunlear coolant without detectable less of the

, boron carbide from the = atrix.

e) Pit:ing corrosion of the aluainum cladding on the boral, however, is I possible where the cladding is in contac: with stainless steel, especially at

hose points of contac: where access to oxygen is highest, and in pools contain-This pitting should not effect che poison capability of the

! ing boric acid.

i boral and should not in any way dissolva or corrode the boron carbide particles l themselves.

l f

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f) Pitting of alu=inu= ' racks, where they are in contact with stainless steel can occur. However, this pitting to date in t racks . recoved from several ,

reactors in which the pools contain high purity water has not produced sig-nificant degradation of the rack materials.

g) Type 17-4 PR stainlets stee.1 should be used in the H-1025-1100 con-dition and tha heat treat =ent scale removed by either chemical or mechanical

=eans to avoid'possible stress corrosion cracking, pitting, and sludge for=ation in the pool.

V. ACKNOULEDGEMENTS This work was performed under the auspices of the U,S. Nuclear Regulatory Commission, which also provided us with the samples of tae Three Mile Island spent fuel pool piping system for examination. The assistance of Mr. R. Kar: mar

of 3 rooks and herkins, Inc., in obtaining the neutron attenuation and cherucal

analyses of the boral punchings from the BMRR is gratefully acknowledged. The assistance of C. Schnepf and W. Lindsay in performing the laboratory investiga-

tions at Brookhaven, the assistance of the Reactor Division staff in obtaining

} the punchings from the Brookhaven Medical Research Reactor for this investiga-tion and in perf o rming the corrosion monitoring studies in the High Flux Beas Reactor spent fuel storage pool, were all essential to this evaluation.

VI. REFERENCES i

i 1. NUREG-0448, Proceedings f rom the URC-IAEA Spent Fuel Storage Meeting, i Feb. 28-March 2, 1978.

2. F. Scott Giacobbe, General Public Utilities, This symposium.

3. C. A. Hippsley, J. F. Knote, and B. C. Edwards, Acta Met 28 (1980) 869.

3

4. P. Gaugh, ARMCO, private co==unication, 1978, citing experience at INCO's

. Kure Beach and Wrightsville Beach Facilities.

5. Vernon: Yankee, Docket 50-271, Spent Fuel Pool Modification Licensing Request, 1977.

6. J. E. Draley and W. E. Ru the r, Report ANL-5001, Feb., 1953.

7. C. N. Vincent, Monticello Sent Fuel Storage Module Corrosion Experience, this sy=posius.

8. i.e. Cocconuealth Edison, Co., Zion Units 1 and 2 and Dresden Units 2 and 3 Spent Fuel Fool Modifications, Public Service Electric and Gas, Salem Unit 1 Spent Fuel Pool Modification; see also ref. 7.

1 4

163/8

. TABLE 1 SilR Spent Fuel Pool Chemistry

• Specification Typical Range conductivity <5 <1

(' ho/c=)

Chloride (ppa) <0.1 <0.02 pH 5.6-8.6 end of lower range SiO2 (pp=) <0.5 .1 .5 Te=perature <52*C (<125*F) 37*C (93*F) 8/15/S0

• Courtesy Coc=onwealth Edison, Co. , Dresden Station TABLE II Pi!R Spent Fuel Fool Chemist ry*

Specificacian Typical Range Boric Acid, as >200C 2300 ppm 3oron Chloride (ppm) <0.15 <0.10 F, pp:a (0.15 <0.02 pH 4.0-7.0 4.7-5.2 Tecpetature <65*C (<150*F) 18-29'c (65-S5*F)

" Composite of data from Portland General Electric Co. ,

Trojan Specifications, and '!e t ropolitan Edison Co. ,

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1. 7 H-1100* followed by grit blasting 1

l #8 & 9 H-1100* Type 304 with stainless steel not on threaded end l

• H-1100 = 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 580 C, air cooled.

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