ML19211A015

Summary of 791101 Meeting W/Vendors,Utils & Licensees Re Cladding Rupture Temp,Cladding Strain & Assembly Flow Blockage

ML19211A015
Person / Time
Issue date:	11/20/1979
From:	Denise R Office of Nuclear Reactor Regulation
To:	Mattson R Office of Nuclear Reactor Regulation
References
NUDOCS 7912140253
Download: ML19211A015 (64)

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UNITED STATES

[}-;, V"d[/ j NUCLEAR REGULATORY COMMISSION E

WASHINGTON, D. C. 20555 q L%j#l NOV 2 01979 MEMORANDUM FOR: Roger J. Mattson, Director Division of Systems Safety, NRR FROM:

Richard P. Denise, Acting Assistant Director for Reactor Safety, DSS

SUMMARY

MINUTES OF MEETING ON CLADDING RUPTURE TEMPERATURE, CLADDING STRAIN, AND ASSEMBLY FLOW SLOCKAGE On November 1,1979, the NRC staff met with reactor fuel vendors, some plant licensees, and other interested parties (Enclosure i lists the attendees who signed the meeting roster) to discuss recently developed staff views on the safety analysis for emergency-core-cooling systems (ECCS). The staff presentations addressed (1) Zircaloy cladding behavior during a loss-of-coolant accident (LOCA) and (2) the sensitivity of the ECCS evaluation models to cladding swelling and rupture. The meeting agenda is attached as Enclosure 2.

During the neeting the Core Performance Branch presented preliminary results (a portion of a draft report is attached as Enclosure 3) of an ongoing generic review of three Zircaloy correlations used in the ECCS evaluation models. The three correlations are cladding ruoture temperature, cladding circumferential strain at failure, and assembly flow blockage (i.e., reduction-in-flow area). Based on the interpretation of experi-mental data (most of which have been obtained subsequent to the 1973 rulemaking hearing and are listed in Enclosure 4), the staff had developed preliminary audit correlations (see Enclosure 5) for these three models.

Anc a comparison of the staff correlations to the approved vendor models showed that over certain temperature and stress regimes the vendor odels underpredicted the degree and incidence of cladding swelling and rupture, thus appearing to violate the requirements of Appendix K of 10 CFR 50.

The Analysis Branch presented the staff's assessment (see Enclosure 6) of what the significance of the discrepancies between the draft audit correlations and the approved vendor models could mean in tems of continued plant compliance to the ECCS acceptance criteria (10 CFR 50.46). Based on limited computer runs, it was thought that the signiff-cance was on the order of hundreds of degrees of peak cladding temperature decending apon the particular ECCS model depencency to cladding strain and ructure. Such was then the reason for the request to meet with the

'ndustry in order to detemine the specifics of the limiting-LOCA analysis for all comreccial plants that use Zircaloy cladding.

Contact:

D. A. Powers, x27603 7912140 2 5_3 1573 300

.-f. As a result of the information received during the meeting, the staff concluded that, while the aporoved vendor models deviated significantly from the staff correlations on an overall basis, within the ranges of interest, one of two circumstances prevailed:

(1) the vendor models were either conservative or very close to the staff correlations within these ranges, or (2) peak cladding temperature was relatively insensitive to the discrepancy. In either case, the fuel vendors (including Yankee Atomic) agreed to provide letters of confirmation to the staff showing that all operating plants would continue to be in conformance with the requirements of 10 CFR 50.46. These letters were to be received by 5:00 p.m. on November 2,1979.

The details of the model discrepancies and the sQni cance of the discrepancies are provided in Enclosure 7.

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t Richard P. Denise, Acting Assistant Director for Reactor Safety Division of Systems Safety cc:

H. Denton E. Case F. Schroeder Q. Eisenhut D. Ross R. Tedesco P. Check R. Woods Reactor Fuels Section L. Phillips G. Lauben POR P. Boehnert, ACRS staff ACRS - 16 opies D. Bessette, ACRS staff G. Marino, RES

0. Hoatson, RES M. Picklesimer, RES W. Johnston, RES 1573 301 L. Olshin S. Schwencer R. Reid W. Gammill H. Rood R. Chapman, ORNL

ENCLOSURE 1 ATTENDEES AT THE MEETING ON CLADDING SWELLING AND RUPTURE NOVEMBER 1.1979 NRC Baltimore Gas & Elec. Co.

D. Eisenhut R. Olson R. Denise G. Lauben ORNL M. Picklesimer P. Boehnert R. Chapman G. Marino F. Coffman Carolina P&L W. Johnston J. Voglewede R. Farr K. Kniel G. Alberthal Wisconsin Public Service Coro.

L. Olshin R. Woods M. Stern S. Rubin R. Hanneman S. Schwencer R. Reid Westinchouse W. Gammill H. Rood S. Kopelic R. Tedesco D. Powers V. Esposito R. Meyer P. Check D. Burman Florida Power & Light Co.

Boston Edison S. Sarkar J. Gosnell J. Keyes Duke Power Co.

EPRI-NSAC S. Rose B. Leyse Atomic Industrial Forum GE F. Stetson R. Elkins Southern Co. Services A. Rao N. Shi rl ey K. Folk Press AEPSC W. Dillehay V. Manno 1573 302

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2 Yankee Atomic Electric Combustion Engineerino A. Husain G. Meuzel S. Schultz E. Jageler J. Cicerchia EXXON Nuclear C. Brinkman D. Kreps R. Collingham G. Owsley Phila. Electric Co.

G. Cook L. Rubino JCP&L B&W - NPGD S. Chan A. Lowe GPU B. Short B. Dunn G. Bond H. Bailey PASNY S. Iyer 1573 303 e

ENCLOSURE 2 MEETING ON CLADDING SWELLING AND RUPTURE NOVEMBER 1, 1979 Introduction R. Denise 10 minutes Cladding Swelling and R. Meyer 60 minutes Rupture Information Potential Effects of New N. Lauben 45 minutes Models on ECCS Discussion Lunch Vendor and Licensec Feedback

- comments on data or models

- suggestions of work to be done

- suggestions for coerating clant actions in the meantime 1573 304

EriCLOSURE 3 ORAFT 10/31/79

0. 3cwers/ R. Meyer CLACOING SWELLING AND RUPTURE MODELS

CR LOCA ANALYSIS D. A. Dewers and R. O. Meyer 1573 305 O

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1.

INTRODUCTION During a postulated loss-of-coolant accident (LOCA), the reactor coolant oressure may drop below the internal fuel red gas pressure causing the fuel cladding to swell (balloon) and, under some conditions, ructure. Core behavior during a LCCA would decend on the time at which swelling and ructure occurred, the magnitude of swelling, and resulting coolant ficw blockage (i.e., reduction in flow area).

Such chencmena were among the many reactor safety issues discussed during the 1973 rule-making hearing on Acceptance Criteria for Emergency Core Cooline Systems (ECCS). The adcated acceptance criteria (Ref.1) limited predicted (calculated) reactor cerformance such that if certain exidation and temcerature limits were not exceeded, then core cooling would be assured. It was recuired that each licensee use a safety evaluatien medel to analytically demonstrate comoliance with the ac:ectance criteria.

Aapendix K (Ref. 2) gives recuirsents for seme features of evaluaticn medels, and, in particular, states that to be accectable the swelling and ructure calculations shall be based on acplicacie data in such a way that the degree of swelling and incidence of rupture are not under-estimated. The degree of swelling and incidsnce of ructure are then used to calculated other c:re variables including gao conductance, claccing tercerature, oxicatien, emerittlement, and hydrogen generaticn.

After the ::nclusien Of tne ECCS hearing, tne 2EC reviewed and accreved clacding benavior mecels for each U.S. fuel manufacturer f:r -heir use in ECCS analyses.

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Curing the ECCS hearing uncertairtties were apcarent in predicting fuel behavior during a LOCA. Therefore, in the Ccmmission's concluding coinien (Ref. 3), the Commission directed the AEC's research office (new the NRC Office of Nuclear Regulatcry Research) to undertake a major confirmatory research program on cladding behavior under LOCA conditions. D e resulting multi-millien dollar program includes simple bench-type Zircaloy tests, single-and multi-red burst tests that simulate some in-reactor conditions, and actual in-reactor tests ranging to full-si:e bundle tests.

The research programs are not all finished, but with the comoletion of many cut-of-oile and a few in-cile tests, we are at a plateau of under-standing that greatly exceeds our understanding in 1971, and the results have not confirmed all of cur previcus conclusions. The trend of these recent data shcws the likeliheed of more ruptures, larger ructure strains, and greater ficw bicckages, than we previously believed.

Consecuently, we see the need to reevaluate all LOCA ' cladding :r.cdels to assure that licensing analyses are cerformed in accordance with Accendix X.

In the follcwing sections we will display the relevant body of data, describe our evaluatien of these data to arrive at useable correlations (curves), and ccm are these correlations with those currently used in licansing analyses. Since the cata snew streng heating-rate' effects,

• Soth neating rate and strain rate are 1 :ccrtant fact:rs in determining

laccing burst ressure and strain. Mcwever, rest burs ex:eriments are no cesigned t distinguish :e-neen nea-ing-rate ef#ects anc strain-rate effects. 20r the Our oses Of :his re crt, One actual differencas are :recaoly uniccortant. Theref:re to avoid c nfusicn, in :ne remaincer of tnis reccr. we will refer to bo:n effects simsly as heating-ra:e effects.

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we have derived different curves for slew camp rates and fast rama rates. But most current ECCS models do not include a ramo rate effect, so we have also disolayed ccmcosite curves that enveloce the slow-ramo and fast-ramp curves.

1573 308

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DATA BASE The ballooning and rupture behavior of Zircaloy are fairly complex phenomena in cart because (a) the stresses are biaxial and the material is anisotrooic in the temoerature range of most interest, (b) the

rocerties of

irconium-base alloys are susceptible to heating-rate effects, (c) oxygen embrittlement increases yield and failure strenaths, and (d) the cracking of oxide coatings results in failure sites that can locali

e stresses. Consequently the behavior of Zircaloy depends strongly on the cladding's environment and hence on test conditiens (Refs x-y).

Therefore, for final calibration of the data correlaticns, we have selected only those data frem experiments in acueous atmosaheres that utili:ed either internal fuel-pellet simulators (i.e., indirect cladding heaters) or actual fuel cellets in reactor. This selection emphasi:es the more recent and more excensive prototypical test data and da-basizes much of the earlier data. Appendix A provides a tabulation cd all of the data we have used, their references, and a legend of symbc,ls that are used for these selected dats sets in the sater figures.

here are heles in this data base, newever, carticularly witr regard to the absence of large bundle tests, and we have utilized the "esults fr:m simoler less typical tests to bridge the gaos. These more cristine tests are atycical in 3 sense, but they do reveal fundamentti features of Zircaloy behavice that allow one to intercret the scarser proto:ycical data.

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1573 309

3.

NEW CORRELATIONS 3.1 Rupture Temeerature The incidence of rupture decends en the differential pressure across the cladding wall, the cladding temcerature, and on the length of time these conditions are maintained. Time duration under burst conditions manifests itself as a heating-ramo-rate effect, and this effect will be treated exDlicitly. We have converted differential pressures to hoco stresses to eliminate design-specific dimensional erfects. The ccnversion was made using the thin-shell formula, 1= {d/2t):P, where is cladding hoco stress, d is the undefomed cladding mid-wall diameter, t is the undefamed cladding thickness, and tP is differential pressure acrcss the cladding wall at rupture. Taele I shews scme ccmouted values of heco strees in tems of differential cressure for c......cn cenreccial fuel designs.

eigure i shcws ruoture temperature data as a functicn cf heco stress fer a wide range of tes cenciciens. While tnis figure shcws the general trend - tuces burst at icwer tercerature when the pressure differential is higher -- the cata are scattered crimarily because of ramo-r'tte effects and excerimental uncertainities in detemining burst temcerature.

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rigure 2 shows CRNt. data at 28'C/sec (a common ramo rate used in the ORNI. experiments) and the basic correlation we will adoot as develcoed by Chacman (Ref. Q) using numerical regressien technicues.

It is clear that most of the data scatter has been eliminated by restricting the data to a single ramo rate. Chacman has also develoced a ramo-rate correlation (Ref. N) that can be used with the basic ructure-temperature correlatien in Fig. 2 to creduce a family of ructure-temcerature curves. Rame-rate has little effect on ructure te perature for rates faster than 28'C/sec.

Three curves that sean the imcortant ramo-rate range are shewn in rig. 3 alcng with the data of Fig.1. Chacman has sncwn that most of the original scatter is explained by came-rate effects, and tne curves in :ig. 3 are seen to span most of the data.

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function of one of these variables, burst tem:erature, and the data scatter is therefore due to temperature measurement dif-ficulties and the other variables mentioned above.

The scatter in Fig. 4 is bewildering, so we have relied on data from less prototypical but more centrolled tests to helo derive a correlation. Figure 5 snows burst strain versus burst temcerature frem Chung and Kassner's work (Ref. T) with short Zirealoy tubes heated by cassing an electrical current directly through tne Zircaloy. Several fundamental features are accarent.

There are three sucerplastic peaks - One in the low tem erature al:ha : nase around 200 C and two in the high-temcerature beta phase around 1050'C and 1225'C. The very imcortant valiey at about 925'C is a consecuence of mixed al ha-olus-beta-onase material, whicn exhibits icw ductility.

Heating-rate effects are also visable; slow-ramo rates creduce large strains in the tc..7erature regime beicw abou: 950'C as a result of feedback effects discussac in lefs. x-y.

5ut slow-ramo rates creduce very mall strains at terceratures greater than about 950% because the Zir:alay has time to oxidi:e and embrittle before significant ballocning can accur.

sst-rame rates = reduce -he occcsite effects in both temcer*,:ure regimes.

To derive the sicw-rama Or elation, which is shcwn in Fig. 5, we aave tnus taken Chung nr.d Kassner's 5'*/sec curve and scaled the

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of 30% in order to bound Chacman's 10*C/sec bundle test. The five hignest coints in Fig. 6 (0-10*C/sec heated-shroud single-rod tests) are cre11minar/ and have not been fully evaluated, but they were disregarded because the heater ;cwer was so icw (about 3W) that the tubes were in effect burst in a =uffle furnace (the heated shrouds). Direct or external heating methods are kncwn to exaggerate racture strains by maintaining artificially small local temcerature variations (see Ref. X), and sucn exceriments were excluded frem our data base. Since the majority of the data is bounded by the curie, we believe that the correlation satisfies the intention of Accendix < not to underestimate the degree of swelling.

It shculd be cautioned that seme very recent, unevaluated data fecm Gernany (Ref X) also shew large strains (uc to 120".), so the catential exists that Fig. 6 may have to be revised ucward.

The fast-ramo correlation is shcwn in Fig. 7.

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Ficure 3 shcws the cemccsite (i.e., enveloce) of the curves in Figs. 6 and 7 along with all of the data frem Fig. A.

7he ecmposite curve gives a gcod recresentation of the data, providing that the causes of small strains (Ref. X) are kept in nind.

3.3 Assembly Flow 31cckage

'lery few measurements of bundle blockage have been made under prototypical conditions and the best attemots are shown in Fic. 9.

It is therefore necessary to derive bundle blockage from single-red

,m burst strains, but this is not straight forward test results have A

shewn that ruptures in a bundle are not c planar.

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X). Notice that enly a few of the rods have burst in this plane.

'4e have chosen the most realistic (minimum flow restriction) of Chacman's definitions of bicekage for the felicwing analysis.

igure 11 shcws the axial distribution Of blockage for Bundle No.1, frem which the maximum bicckage is seen to be ag5.

Figure 12 shews the gecretric relation between average roc strain and bundle blockage for a scuare array of c:mmercial-si:e tuces.

Frem this figure it can be seen that an average red strain :f 275 vould cause a bundle blockage of 195. Since the average ructure strain for reds in Bundle !!c. I was 125 (see Ac:encix A), the blockage :an be cotained ' rem the aucture strain by "Jlti: lying by 0.52 (the rnio of 27 :: 12) and utili:ing Ffg. 12. 7he similar anics for 3 uncles No. I and ?:o. 3 are 0.57 and 0.70 civinc an averal1 averace 'Or t."e tnree bundle teses cf 0.57.

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Assuming that the distributions of ruptures in Chapman's bundle tests are typical, the local blockace correlation is thus for ed by multiclyinc strains in Figs. 6 and 7 by 0.67 and then utilizing

'le have called this result " local bicckace," as distinct Fic. 12.

' rom the desired assembly blockage, because it does not yet represent large comneccial-si:e bundles or include the effects of non-fueled tubes, which would not ballcon. The sicw-and fast-ramo local blockage curves are shewn in Figs. 13 and la where they are compared with the scarse collection of data. Figure 15 shcws the ccmcasite

1. 1cw blockage curve, which enveloces the curves in Figs.13 and 14 Finally, to obtain asseatly flow blockage, two adjust ents are recui red. First, it must be reccgni:ed that bundle-average bicck-age, which is desired, is a function of bundle si:e. This can be seen by envisioning an 8x8 test buncle that is analyzed quadrant by cuadrant. If each 4x2 cuadrant is vieved as a small buncle, the olanes of maximum bicckage for the quadrants would be ex:ected to Occur at different elevations because of some randceness of the process. Cne would therefore expect to 'ind the plane of maximum blockage in each quacrant to have greater ficw restricticn than ne clane of maxi: um bicckage in :ne bundle taken as a whcie. ~ hat is, the large bundle si:e introduces an averaging effect.

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$a De 15 g.

+

4 W g.

m O

D R-

[

tt Ci a

$GINBBRING IIDOP STRESS (EPSI) c~

ti

N I

4Q M

r.c

.g o 2

5 E

o E

-S 3*

8 A

1=

8=

5 3

--e @

s

=

=

-e a 8

ban s

6 6

6

(*i) TZEY.RJ12 El SOTDGGT2 f

}573 331

data c:m=arisens because the bundles rearesented in Fics.13-15 were all small arrays). For Sundle No. 1, the averace (al") of tne blockages was found between the 23-cm and 47-cm locations in an attemet to eliminate the suceressing effect o# s acer grids at 10 cm and 55 cm. Shilar averages were fcund for Sundles No. 2 and ?:o. 3.

Using these values the ratio to be used to derive larce-bundle blockages from ructure strain data is 0.55 (cercared with 0.57 for small arrays). This ' actor was used to derive all of the blockace curves in the next section of this report.

The second adjust ent is a reduction of acout 5". to acteunt for instrument tubes and guidetubes that would not balleen. The exact scaling factor SF decends On the fuel design and is civen by SF=NA/(,1A

.1A},

g7 7g 7g anere N is the num=er of fuel reds, A is the 'lew area ar0und an g

p undefor ec fuel red,1 is the number of guidetubes or instrument g

tubes, and A is the #1cw area around an undefor ed guidetube r instrsent tute. This scaling factor was also emolayed in deri eing the blockage curies in the next section.

1573 532 M

ko

ENCLOSURE 4 APPENDIX A FUEL CLAODING BURST DATA DATA REFERENCE A (Uprignt Triangle)

FRF-1 R. A. Lorenz, D. O. Hobson, and G. W.

Parker, " Final Report on the First Fuel Rod Failure Transient Test of a Zircaloy-Clad Fuel Rod Cluster in TREAT," Oak Ridge National Laboratory Report, ORNL-4635, March 1971. Availaole in public technical libraries. Also available from National Technical Information Service (NTIS), Soringfield, Virginia 22161.

R. A. Lorenz, D. O Hobson, and G. W. Parker, " Fuel Rod Failure Under Loss-of-Coolant Conditions in TREAT," Nuclear Technology, II, p. 502 (August 1971).

Available in public technical libraries.

Inoile, 7-rod bundle, steam atmosphere.

Maximum reduction in bundle flow area = 48 %.

Mean red burst strain = 36 %.

Mean rod burst temperature. 389*C.

Mean rod engineering burst stress. 1.71 Kpsi.

e RCD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS 1

(*C/S)

(PSIG)

( C)

( *. )

((PSI)

H 25-36 172 966 26 1.39 A-1 25-36 250 799 35 2.02 R

25-36 205 743 36 1.66 A-2 25-36 290 816 42 2.34 L

25-36 162 915 36 1.31 I

25-36 190 827 35 1.5A C

25-25 215 310 40 1.74 1573 333

DATA REFERENCE 8 (Cross)

R. H. Chapman, "Multired Burst Test Program Progress Report for April-June 1977," Gak Ridge National Laboratory Report, ORNL/NUREG/TM-135, June 1977.

Available in public tecnnical libraries. Also av.ailable from National Tech-nical Information Service (NTIS), Springfield, Virginia 22161.

R. H. Chapman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of Creep Time and Heating Rate on Deformation of Zircaloy-4 Tuces Test in Steam with Internal Heaters," Oak Ridge National Laboratory Report, NUREG/CR-0343:

ORNL/NUREG/TM-245, October 1978. AvailaDie in public technical libraries.

Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.

Out-of pile, single rod, steam atmosphere.

R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

("C/S)

(PSIG)

(*C)

(%)

(KPSI)

PS-1 28 922 893 18 7.47 PS-3 28 809 873 29 6.56 PS-4 28 850 871 21 6.88 PS-5 28 830 882 26 6.72 PS-10 28 870 901 20 7.05 PS-12 28 891 898 18 7.21 PS-14 28 844 883 23 6.84 PS-15 28 393 885 17

7. 2a PS-17 28 1760 778 '

25 14.2 SR-1 28 116 1166 26 0.94 SR-2 28 146 1082 44 1.19 SR-3 28 249 1011 43 2.02 SR-4 28 650 921 17 5.26 SR-5 28 1380 810 26 11.2 3R-7 23 2090 736 20 17.0 SR-3 28 178 1020 43

1. 44 SR-13 28 155 1079 79 1.26 SR-15 28 2780 714 14 22.5 SR-17

.8 154 1049 53 1.25 SR-19 23 2760 688 16 22.4 SR-20 28 154 1049 55 1.25 SR-21 28 162 1023 18 1.32 SR-22 28 129 1081 50 1.05 SR-23 29 139 1077 35 1.13 SR-24 29 laa 1057 67 1.16 SR-25 28 139 1092 78 1.13 3R-26 23 120 1130 34 0.38 SR-27 28 133 1084 41

1. C 8 5R-23 23 1220 835 27 3.57 SR-29 23 1170 343 27 9.45 3R-37 28 1967 760 23

15. 3 SR-38 23 1998 770 20

' f. 2 1573 534

DATA REFERENCE C (Plus)

MRBT-3-1 R. H. Chapman, "Multired Burst Test Program Progress Report for July-Decemoer 1977," Cak Ridge National Laboratory Report, NUREG/CR-0103: ORNL/NUREG/TM-200, June 1978. Available in public technical libraries.

Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.

R. H. Chacman, " Preliminary Multired Burst Test Program Results and Implications nf Intorcst to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Researen Information Meeting, Gaithersburg, MD., November 7, 1978. Available in POR for inspection and copying for a fee.

Out-of-pile,16-rod bundle, steam atmosphere.

Maximum reduction in bundle flow area = 49 %.

Mean red burst strain = 42 %.

Mean rod strain in plane of maximum blockage = 27 %.

Mean rod burst temcerature = &&s*c -- g (33 Mean rod engineering burst stress = 8.72 Kpsi.

ROD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

(*C/S)

(DSIG)

(*C)

(%)

(KPSI) 1 29 1124 852 36 9.10 2

29 1075 867 32 8.71 3

29 4

29 1052 860 36 9.33 5

29 1005 372 45 3.14 6

29 1104 372 43 3.34 7

29 1052 369 36 3.52 3

29 1074 872 42 3.70 9

29 1030 870 27 3.34 10 29 1059 873 45 a.58 11 29 1054 947 53 S.54 12 29 1114 363 37 9.02 13 29 1091 378 59 3.24 14 29 1066 375 22 3.63 15 29 1C62 B65 42 3.60 16 29 1092 848 39 3.85 e

1573 3,53

DATA REFERENCE C (Plus)

MRBT-3-2 R. H. Chacman, "Multired Burst Test Program Progress Report for July-Oecember 1977," Oak Ridge National Laboratory Report, NUREG/CR-0103: ORNL/NUREG/TM-200, June 1978. Available in public technical libraries.

Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.

R. H. Chapman, "Multired Burst Test Program Progress Report for July-December 1978," Oak Ridge National Laboratory Report, NUREG/CR-J655: ORNL/NUREG/TM-297, June 1979. Available in public technical libraries. Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.

Out-of pile, 16-red bundle, steam atmosphere.

Maximum reduction in bundle flow area = 53 %.

Mean rod burst strain = 42 %.

Mean rod strain in plane of maximum blockage = 28 %.

Mean red burst temperature = 858'C.

Mean rod engineering burst stress = 8.88 Kpsi.

R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

( C/S)

(PSIG)

( C)

(4)

(KPSI) 1 29 1117 870 35 9.05 2

29 1115 846 39 9.02 3

29 1096 853 40 8.88 4

29 1100 872 42 8.91 5

29 1127 866 35 9.13 6

29 1004 857 58 8.13 7

29 1067 861 56 8.64 8

29 1097 856 38 8.89 9

29 10 29 1065 856 43 8.63 11 29 1112 853 40 9.01 12 29 1094 851 20 8.86 13 29 1134 883 41 9.19 14 29 1048 858 42 8.49 15 29 1152 836 35 9.33 16 29 1117 848 42 9.05 1573 336

DATA REFERENCE C (Plus)

MRST-8-3 R. H. Chapman, " Preliminary Multired Burst Test Program Results and Implications of Interest to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MO., November 7,1978.

Available in POR for inspection and copying for a fee.

R. H. Chapman, "Multirod Burst Test Program Progress Report for April-June, 1979," Oak Ridge National Laboratory Report, NUREG/CR-1023: ORNL/NUREG/TM-351, in publication.

Out-of-pile,16-rod bundle, steam atmosphere.

Maximum reduction in bundle flow area = 75 %.

Mean rod burst strain = 57 %.

Mean rod strain in plane of maximum blockage = 40 %.

Mean rod burst temperature = 764*C.

Mean rod engineering burst stress s 11.07 Kosi.

f <.v RAMP PRESSURE SURST SURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

_1_

(*C/S)

(PSIG)

( C)

(%)

(KPSI) 1 10 1393 771 48 11.23 2

10 1280 779 76 10.39 3

10 4

10 1318 767 55 10.68 5

10 1375 764 63 11.14 6

10 1327 770 61 10.75 7

10 8

10 1320 756 78 10.69 9

10 1320 754 59 10.69 10 10 1362 774 50 11.03 11 10 1396 775 57 11.31 12 10 1414 761 47 11.45 13 10 1186 760 29 12.04 i4 10 1405 769 42 11.38 15 10 1335 753 53 10.81 16 10 1407 747 59 11.40 1573 337

DATA REFERENCE O (Closed Circle)

F. Ercacher, H. J. Neitzel, and K. Wiehr, " Interaction Bet'aeen Thermanydraulics and Fuel Clad Ballooning in a LOCA, Results of RESEKA Multired Burst Tests with Flooding," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD, November 7, 1978. Available in file for USNRC Report, NUREG-0536.

F. Erbacher, H. J. Neitzel, M. Reimann, and K. Wiehr, " Fuel Rod Behavior in the Refilling and Reflooding Phase of a LOCA-Burst Test with Indirectly Heated Fuet Rod Simulators," paper presented at the NRC Zircaloy Cladding Review Grouo Meeting, Idaho Falls, May 23, 1977. Available in file for USNRC Report, NUREG-0536.

K. Wiehr and H. Schmidt, "Out-of-Pile Experiments on Ballooning of Zircaloy Fuel Rod Claddings Test Results with Shortened Fuel Rod Simulators,"

Kernforschungszentrum Karlsruhe Report, KfK 2345, October 1977. Available in file for USNRC Report, NUREG-0536.

F. Erbacher, H. J. Neitzel, M. Reimann, and K. Wiehr, "Out-of-Pile Experiments on Ballooning in Zircaloy Fuel Rod Claadings in the Low Pressure Phase of a Loss-of-Coolant Accident," Proceedings of Specialists' Meeting on the Behavior of Water Reactor Fuel Elements Under Accident Conditions, Spatind, Norway, Septemcer 13-16, 1976. Available in public technical libraries.

F. Ertacner, H. J. Neitzel, and K. Wiehr, " Studies on Zircaloy Fuel Clad Ballooning in a LOCA, Results of Surst Tests with Indirectly Heated Fuel Rad Simulators," caper presented at the ASTM 4th International Conference on 2irconium in the Nuclear Industry, Stratford-on-Avon. England, June 27-29, 1978. Available from ASTM.

Out-of pile, single rod, air and steam atmospnere.

R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT SURST TEMPERATURE STRAIN SURST STRESS

( C/S)

(PSIG)

( C)

(U

(< PSI)

?

11

?

880 27

?

11 356 380 51 5.91

?

?

365 33

?

?

11

?

260 JJ

?

?

11

?

340 32

?

?

11

?

30 36

?

?

1'

?

220 13

?

?

1'

?

40 5J

?

11

?

330 27

?

?

11

?

325 27

?

11

?

325 33 13 11

'220 323 33 9.31

?

320 28

?

?

320 38

?

la l'

1220 310 38 9.31 1573 338

DATA REFERENCE D (Continued)

ROD RAMP PRESSURE BURST BURST ENGINEERINC RATE AT SURST TEMPERATURE STRAIN BURST STRESS 1

(*C/S)

(PSIG)

( C)

( *.' )

(KPSI)

?

11

?

810 42

?

?

11

?

810 44

?

35 11 1380 794 27 9.54

?

11

?

780 27

?

?

11

?

780 30

?

?

11

?

780 52

?

?

11

?

770 25

?

?

11

?

770 32

?

?

11

?

760 24

?

?

11

?

755 23

?

?

11

?

755 52

?

1573 339

DATA REFERENCE E (0 pen Circle)

E. Karb, "In-Pile Experiments in the FR-2 DK-LOOP on Fuel Rod Behavior During a LOCA," paper presented at the US/FRG Workshop on Fuel Rod Behavior, Karlsruhe, June 1978. Available in file for USNRC Report, NUREG-0536.

E. H. Karo, "Results of the FR-2 Nuclear Tests on the Behavior of Zircaloy Clad Fuel Rods," pacer presented at the 6th NRC Water Reactor Safety Research Infor-mation Meeting, Gaithersburg, MD, November 7,1978. Available in file for USNRC Report, NUREG-0536.

E. H. Karo, "Results of FR-2 In-Pile Tests on LWR Fuel Rod Benavior," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979. Available in PDR for inspection and copying for a fee.

Inpile, single rod, steam atmosphere.

R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

( C/S)

(PSIG)

(*C)

(%)

(KPSI)

A1.1 7.1 725 810 64 5.01 A2.1 20 1276 820 36 8.82 Bl. 6 8.2 1160 825 38 8.02 B3.1 10 1146 825 37 7.92 31.3 12.7 885 845 34 6.12 A2.2 12.1 841 860 56 5.81 31.1 17.5 754 900' 30 5.21 81.5 9

653 910 60 4.51 31.2 B. 7 653 915 25 4.51 33.2 12.1 725 915 50 5.01 1573 540 4

DATA REFERENCE F (Square)

R. H. Chapman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of

~

Creep Time and Heating Rate on Deformation of Zircaloy-4 Tubes Tested in Steam with Internal Heaters," Oak Ridge National Laboratory Report, NUREG/CR-0343:

ORNL/NUREG/TM-245, October 1978. Available in public technical libraries.

Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.

Out-of pile, single rod, steam atmosphere.

R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE 5 TRAIN BURST STRESS 1

('C/SJ (PSIG)

(*C)

( *.' )

(KPSI)

SR-33 0

825 762 23 6.68 SR-34 0

844 766 32 6.84 SR-35 0

648 775 29 5.25 SR-36 0

660 821 29 5.35 SR-43 4

1105 773 29 8.95 SR-44 5

1060 777 30 8.59 SR-41 9

1416 757 27 11.5 SR-42 10 1373 761 28 11.1 1573 341

DATA REFERENCE G (Asterisk)

REBEKA-1,

-2, -3 F. Erbacher, H. J. Neitzel, and K. Wiehr, " Interaction Between Thermohydraulic and Fuel Clad Ballooning in a LOCA, Results of REBEKA Multired Burst Tests with Flooding," paper presented at the 6th NRC Water Reactor Safety Researen Information Meeting, Gaithersburg, MD., November 7,1978. Available in file for USNRC Report, NUREG-0536.

K. Wiehr, "Results of REBEKA Test 3," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979.

Available in POR for inspection and copying for a fee.

Out-of pile, 9-rod bundles, steam and water atmosphere.

TEST INITIAL MEAN MEAN MEAN MEAN REDUCTION RAMP PRESSURE BURST BURST ENGINEERING IN FLOW RATE AT BURST TEMPERATURE STRAIN BURST STRESS AREA

(*C/S)

(PSIG)

(*C)

(%)

(KPSI)

(%)

1 7

870 815 29 6.01 25 2

7 800 870 53 5.53 60 3

7 725 830 44 5.05 52 157L3 342

DATA REFERENCE H (Inverted Triangle)

M. Bocek, "FABIOLA," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho F&lls, Idaho, June 22-29, 1979. Available in POR for inspection and copying for a fee.

Out-of pile, single rod, steam atmosphere.

ROD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS

_f_

( C/S)

(PSIG)

(*C)

(%)

(KPSI) 1 3

563 860 66 3.92 a

11 1375 790 8

9.58 8

7.8 1375 780 35 9.58 10 10 2013 750 33 14.03 12 9

563 890 29 3.92 13 10 1810 765 10 12.62

} [)7 )

DATA REFERENCE I (Diamond)

J. L. Crowley (ORNL), personal communication to D. A. Powers (USNRC),

August 10, 1979.

R. H. Chapman (ORNL), personal communication to 0. A. Powers (USNRC),

Septemoer 11, 1979.

Out-of pile, single rod, heated shroud, steam atmosphere.

R00 RAMP PRESSURE BURST MAXIMUM ENGINEERING RATE AT BURST TEMPERATURE ROD STRAIN BURST STRESS 1

(*C/S)

(PSIG)

(*C)

(%)

(KPSI)

SR-47 10 1436 775 W 73 12.35 SR-49 5

1139 775 98 9.80 SR-51 0

1030 790 93 8.86

'R-53 0

841 760 83 7.23 9-57 0

725 775 110 6.23 1573 344

ENCLOSURE 5

. Enclosed are 3 figures that shcw correlations in the 10/31/79 draft for rupture temoerature,,pture strain, and assemoly ficw blockage.

Temperature rama rates are accounted for in the correlations, and the ramp rates that are most appropriate should be used. If it is not practical to accomodate ramp rates in the code, envelopes of these curves snould be used. The tacular values from wnich these curves were generated are also enclosed.

1573 345

,sv e_D R

BURST TEMPERATURE CURVBS W

b'>

EE5D Og.

b;-M

\\

w

-h, 0

b 3

198-Q E' g_

a ca bg.

D C]8 4 Cf8

'h 0--

i i

e 6

1D in 30 85 u

ENGINEERING HOOP STRESS (KPSI)

I l

I m

t I

~

ss S

I 8

lg l

i

... 'l.

\\

oc

-le$

)

./*

/.

/ <.

CL.

/

m

~

p I

\\

N s

8 s

N

\\'.

<,., % s v L 2 m E a w

~

p::R :1HL

,sn so

M 0

D PWR ASSEMBLY PLOW BLOCKAGE CURVES G--o M

pam1MD = SmW-RAMP DerTED = FMDAMP (c-J) p M

b 6 cb 4 B-t2 7:4 gg e~x 4

/

g b

.F

/

,\\

g.

~

/

\\

I

\\

x

" Q-l

\\

'/

\\

3

\\

f g-

)

y s._

N O

u g

g g

g 0

5 1D 15 30 B5 ENGINEERING HOOP STRESS (KPSI) co

Slow-Ramo Correlations 0*C/5

<10*C/5

<10 C/5 Surst Flcw Engineering Strain Blockage Surst Hoco Stress Temcerature

{",)

( ". )

4

('C)

(KPSI) 15.2 20 15.56 1

600 17.6 21 14.40 2

625 26.1 30 13.22 3

650 39.4 44 12.03 675 54.6 4

58 10.84 5

700 56.5 70 9.68 725 72.7 6

78 8.53 7

750 74.6 20 7.40 775 73.6 3

79 6.30 300 58.4 9

72 5.24 10 825 50.8 54 4.26 11 850 34.7 39 3.36 12 875 26.5 31 2.59.

13 900 25.1 30 1.98 la 925 25.5 1.52 31 15 950 25.2 29 1.20 16 975 18.5 0.97 22 17 1000 15.2 20 0.80 18 1025 15.2 20 0.58 19 1050 15.2 20 0.59 20 1075 15.2 20 0.51 21 1100 15.2 20 0.45 22 1125 15.2 20 0.21 23 1150 15.2 20 0.37 21 1175 15.2 0.33 2C 25 1200 1573 349

Fast-Ramo Correlations 28*C/S

>55'C/S 225*C/S Burst Engineering turst Flow Temeerature Hoop Stress Strain 81ockage

('C)

(KPSI)

(i)

(i) 1 600 31.14 20 15.2 2

625 28.74 21 17.6 3

650 26.39 23 20.0 a

675 24.01 26 22.3 5

700 21.65 33 29.0 6.

725 19.32 50 17.5 7

750 17.04 63 60.3 8

775 14.78 67 64.6 9

800 12.57 67 64.6 10 825 10.46 63 60.3 11 850 8.50 54 50.8 12 375 6.70 39 3a.7 13 900 5.17 23 20.0 la 925 3.95 22 18.5 15 950 3.04 23 20.0 16 975 2.40 34 29.9 17 1C00 1.94 57 53.7 18 1025 1.6i i8 72.7 19 1050 1.36 30 74.6 20 1075 1.17 78 72.7 21 11C0 1.03 37 32.8 22 1125 0.91 35 30.9 23 1150 0.31 37 32.8 21 1175 0.73 17 43.2 25 1200 0.67 57 53.7 1573 350

Composite Correlations Burst But-s t Engineering Flcw Temeerature-Strain Hoop Stress Blockage d

(* C)

( ". )

(KPSI)

(". )

1 600 20 31.14 22.3 2

625 21 28.74 22.3 3

650 30 26.39 22.3 a

675 44 24.01 22.3 5

700 58 21.55 29.0 6

725 70 19.32 47.5 7

750 78 17.04 60.3 8

775 80 9

800 79 14.78 64.6 10 825 72 7.10 74.6 11 850 54 6.30 73.5 12 875 39 5.24 68.2 13 900 31 4.26 50.8 la 925 30 3.36 34.7 15 950 31 2.59 26.6 16 975 34 2.40 29.9 17 1CCO 57 1.94 53.7 18 1025 78 1.51 72.7 19 1050 30 1.36 72.6 20 1075 78 1.17 72.7 21 1100 37 1.03 32.8 22 1125 35 0.91 30.9 23 1150 37 0.81 32.3 22 1175 47 0.73 13.2 25 1200 57 0.67 53.7 1573 351

ENCLOSURE 6 GENERAL APPENDIX K REQUIREMENTS MUST BE MET.

REVISED MODELS MAY BE REQUIRED FOR ALL VENDORS.

ALL BREAK SIZES NEED TO BE CONSIDERED.

IF UNCERTAINTIES ARE NOT CONSIDERED IN SWELLING AND RUPTURE CURVES, APPROPRIATE SENSITIVITY STUDIES MUST BE PERFORMED.

WIDTH OF THE VALLEY MAY BE AS INPORTANT AS HEIGHT OF THE PEAK.

1573 352

~.

o e-

~~

m

DETERMINING MAGNITUDE NEED TO SORT OUT SUBSTANTIVE CONDITIONS.

SOME TEMPERATURES AND RAMP RATES MAY NOT BE EXPECTED.

THEREFORE MODELS NEED ONLY APPLY WHERE CONDITIONS ' LLL d

OCCUR.

PWR RUPTURE TEMPERATURES 840 *C - 960

• C BWR RUPTURE TEMPERATURES 960 C - 1200 *C RAMP RATES 2.5 'c /sec. - 7.5 *c/sec - 1573 353 s

a

9 WREM & BURST TEMPERATURE CURVES g.

B

'l cJE 4

I h

0-

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c-h' i

7=

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i a0-

{

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e i

4

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ac/s 4 c/s e6944 <

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ENGINEERING HOOP STRESS (KPSI) u u

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i Apr# RPEAK REFLOOD i

PWR REFLOOD AT FLOODING RATES LESS THAN 1 IN./SEC. APP TO BE WORST CONDITION BECAUSE OF APPENDIX K REQUIREMl FOR STEAM COOLING AND BLOCKAGE.

i NRC PERF0FF.ED LIMITED SENSITIVITY STUDY ON BLOCKAGE, STRAIN, AND INCIDENCE OF RUPTURE.

1573 357 9

m

SWELLING AND RUPTURE REFLOOD STUDY 6

RUPTURED NO )E UNRUPTURED N0DE T

DLOCKAGE ELEV.

STRAIN TIME PCT ELEV.

STRAIN TIME PCT FRACTION IN.

SEC.

• F IN.

SEC.

• F BLOCKAGE STRAIN t "E T['P INIT I

u Case MODEL H0 DEL 1

WREH WREM 29.8 876 1804.1

.522 80.24

.444 260.

Helt 13.51 085 260.

Helt 2

VEllDOR WREM 29.8 876 1804.1

.362 80.24

.444 44.25 l824.3 13.51 085 298.

2143.1 3

VENDOR 1.0 36.4 915 1932.0

.238 80.24 1.0 45.5

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1869.(

4 VEllDOR WREM 45.

976 1780.0

.31 0 80.24 394 44.25 1791.8 13.51 088 298.

2040.1 5

VENDOR 1.0 45.

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s 1573 359

IMPORTANT PARAMFlEE FILL GAS 2

CLADDING TEMPEPATURE FUEL TEMPERATURE BURNUP (FISSION GAS)

DIMENSIONS PLENUM TEMPERATURE POWER PLASTIC STRAIN HEAT TRANSFER v

i PIN PRESSURE (STRESS)

RAMP PATE i

RUPTURE TEMPERATURE

~

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v STRAIN

!BLO'CKaGE I

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CLADDING DIf1ENSIONS FLOW AREA RADIATI0tl CONVECTION GAP HEAT TRANSFER FLOW DIVERSION METAL-WATER REACTION 4

lHEATTRANSFER

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CLADDING TEMPERATURE OXIDATION b.

1573 360 O

ENCLOSURE 7 s

/aga N 4<<'%~

NUCLEAR REGULATORY COMMISSION UNITED STATES

~ -

g WASHINGTON. D C. 20555

- (i ***l [>%..f

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......f November 2,1 79

'S,

ME" :: -." :::-

CH3ir ar Hendrie Co -issioner Gilinsky

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v.a. n r e. d.y

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o 2-2.

COMBUSTION ENGINEERING:

CE had acently provided a new analysis model using improved ructure strain and flow blockage models that are similar to owr new data curves and that show compliance with the 22000F limit for CE operating plants. The improved CE models follow guidelines that are in approximate agreement with the nea NRC staff curves.

Inasmuch as CE used even larger blockages in their analysis than we are now reco rending, we believe that they have additioral conservatism in their PCT results for operating plants.

3.

GENERAL ELECT 3!! AND EXXDN-BWP: Both BWR models apoear to be si gni f1:artly 1'ferent fror the majority of the new data, but this is Sec use "e recert data falls in a range that is much more acplicable to DWR ocerating conditions.

In the hi gh-temoerature, slow-ramp range a::licatilit;. to coeriting EWDs, the SE and Exxon curves are in good agree ert wit 9 the tiew NPC curves.

Thus, there does not ap; ear to be a sa'et. Oroblem witn PCT's for operating BWP's.

4 EXXON :W::

The most incertan* curve used in previous PWR Exxon models i s

• e 'as t-ra : blockage curse.

This curve is conservat!ve with respect tc t*e new ND: staff :urve based on recent data over the ran ge :' at:' i t a t i ' i t., for c:erating DWR's wit 1 Exxor fuel.

The lesser impcrtant strain curve and slow-ram; blockage curve are in a:Dr0)i ate a;-ee eat wit

• the new NRC sta" curves.

In the regions whe-e t*:se Esv c curves so ew*at under:redict tne NRC curves, Exxon

inte: :: t*a cur data do not contradict thei r curves.

According!j.

t*.ere ::ss act a:

Dear to be a sa fet,

. probler witn PCT's fcr operating Ph:'s ait-Ence fuel 5.

WEI".N3-% ?-

aesting*0use strongly disagrees wir the ap;'icabili ty o f tr.e

ata and t9erefore the new NRC staff blockage model; however, its strain curve is in accroximate agreement with the new NPC sta" :arves Over the range of operating conditiors of coer3 ting plar 5.

The Westi9;" Case blockage curve agrees well with the new NPC s*a## curves On!j Over a li-ited stress range, but outside of t"at rarge t*e :ueves d' verge sharply with the '.desting"Ouse curves being less :;rse~.a-4 ve !"aa t9e NRC s ta ff curves. West'agh0use believes t*at ';e' :li::ia.g ra:* ares in its 00erati"; 0' arts aculd a'wajs Cur in !*at lici*id s* ress r3n ge atere i*s model predicts valdes very near to 090se in the new NPC sta#f Sodel.

Additional information on this issue 's er:e ted frca-Wes*inghouse by late Novem er 2, ICIO

}

6

At the conclusion of the November ' meeting, each of the five nuclear fuel suppliers was directed to provide..4RC a letter by close of business Nover.cer 2,1979 confir'r.ing the preliminary information discussed above, and specifically providing answers to the following questions:

On the basis of information presented by the NRC staff at the meeting on Neverber 1,1979 (1) Are your operating plants safe, i.e., do they meet the 22000F lirit for peak clad temperature during a LOCA?

Describe the basis.for this conclusion.

(2)

De *he evaluation models meet the recuirerents in Accendix K?

Describe the basis for this conclusion.

(3)

I' "no" to the abcve, Or: pose nodi'ied operating lirits of suc plants.

Discussions at the reeting, su-' arized above, proviced a preliminary unders*ancing of the expectec contents of the five letters due late today As incicated above, we ex ect each of tne letters will provide ar acceptable basis for continue c:eration of affected placts a+ ieast over the short terr.

However, the in'or ati:n to be received by ever.ing ':cve-'cer 2, ray indicate that a srall number of ;1 arts have rin:r safetj ceficiencies requirir; short ter 3:tice :,. tne NRC.

Ever ais;-in; in3, all c: grating :13n*,s can,;stiey ::ntinued coe-ation over the shcet *er. we believe *ha*.

it will be necessary *o take a:tice *o reqaire c"ar.ges in a numter of the evalaa* ion mocels recuirec by A::endix <

o f O Jr re ga' a

• ion.

Ha r:10 R. Dea t:. N -ec*or Office o' Nuclear Rea:*:r Regulatice 00: OG 0;E SE:V Contact-Carreli 3. Eisenn,,t X2767:

1573 363

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