ML19257C060
| ML19257C060 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1979 |
| From: | Meyer R, Powers D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19257C053 | List: |
| References | |
| TAC-12731, TAC-30209, NUDOCS 8001240053 | |
| Download: ML19257C060 (51) | |
Text
.
ORAFT 10/31/79
- 3. Powers / R. Meyer CLACDING SWELLING AND RUPTURE MODELS FOR LCCA ANALYSIS
- 0. A. Powers and R. O. Meyer 1804 172-8 0 012 4 0 O O
1.
INTRCCUCTICN Ouring a postulated loss-of-ccolant accident (LCCA), the reactor ccolant oressure may drco belcw the internal fuel red gas pressure causing the fuel cladding to swell (balloon) and, under some ccnditions, ructure. Core behavior during a LCCA would depend on thE time at which. welling and ruoture occurred, the magnitude of swelling, and resulting coolant ficw blockage (i.e., reduction in flow area).
Such chenomena were among the many reactor safety issues discussed during the 1973 rule-making hearing on Acceptance Criteria for Emergency Core Coolina Systems (ECCS). The adcoted acceptance criteria (Ref.1) limited predicted (calculated) reactor oerformance such that if certain oxidation and tem erature 1imits were not exceeded, then core ccoling would be assured. It was recuired that each licensee use a safety evaluatien medel to analytically demcnstrate ecmoliance with the accectance cri teri a.
Accendix K (Ref. 2) gives recuirements for seme features of evaluation medels, and, in particular, states that to be acceptable the swelling and rupture calculations shall be based on acolicable data in such a way that the degree of swelling and incidence of ructure are not under-estimated. D e degree of swelling and incidence of ructure are tren used to calculated other core variaoles including gao conductance, cladding temcerature, oxidation, emerittlement, and hydrogen generation.
After the conclusion of the ECCS hearing, the AEC reviewed and accreved clacding behavice models for each U.S. fuel manufacturer for their use in ECCS analyses.
1804 173
During the ECCS hearing uncertainties were apparent in predicting fuel behavior during a LOCA. Therefore, in the Comission's concluding opinion (Ref. 3), the Ccmission directed the AEC's research office (now the NRC Office of Nuclear Regulatory Research) to undertake a maj1r confimatory research program on cladding behavior under LOCA conditions. The resulting multi-million 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-size bundle tests.
The research programs are not all finished, but with the comoletion of
- any cut-of-pile 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 confir::ed all of our previous conclusions. The trend of these recent data shews the likelihecc Of more ruptures, larger rupture strains, and greater ficw blockages, than we previously believed.
Consecuently, we see the need to reevaluate all LOCA cladding models to assure that licensing analyses are performed in accordance with Aopendix X.
In the 'olicwing sections we will display the relevant body of data, cescribe our evaluatien of these data to arrive at useacle correlations (curves), and ccmcare these correlations with those currently used in licensing analyses. Since the data snew streng heating-rate
- effects,
- 30th heating rate and strain rate are imcortant factors in determining cladding burst pressure and strain. Mcwever, most burst exceriments
(
are not designed to distinguish between heating-rate effects and strain-
~
rate effects. For the cur::oses of this recort, the actual differences are r0bably unimcortant. Therefore to avoid c:nfusion, in the remainder
.ef.
Of this recort we will refer to both effects simoly as heating-rate o
effects.
m we have derived df fferent curves for slow ramp rates and fast ramo rates. But most current ECCS models do not include a ramp rata effect, so we have also disolayed ccmcosite curves that enveloce the sicw-ramo and fast-ramp curves.
1804 175,
f
2.
DATA P.ASE The ballooning and rupture behavior of Zircaloy are fairly complex phenomena in part because (a) the stresses are biaxial and the material is anisotrooic in the temcerature range of most interest, (b) the procerties of zirconium-base alloys are susceptible to heating-rate effects, (c) oxygen embrittlement increases yield and failure strengths, and (d) the cracking of oxide coatings results in failure sites that can localize stresses. Consequently the behavior of Zircaloy depends strongly on the cladding's environment and hence on test conditions (Refs x-y).
Therefore, for final calibration of the data correlations, we have selected only those data frem exceriments in aqueous atmospheres that utilized either internal fuel-pellet simulators (i.e., indirect cladding heaters) or actual fuel ;ellets in reactor. This selection emchasizes the more recent and more ex;ensive prototyohal test data and deemphasizes c:uch of the earlier data. Appendix A provices a tabulation of all of the data we have used, their references, and a legend of symbols that are used for these selected data sets in the later figures.
inere are holes in this data base, however, particularly with regard to the absence of large bundle tests, and we nave utilized the results from simpler less typical tests to bridge the sacs. These more aristine tests are atypical in a sense, but they do reveal fundamental features of Zircaloy behavior that allow one to interpret.the sparser prototypical data.
1.
1804 176
3.
NEW CORRELATIONS 3.1 Rupture Temperature The incidence of rupture depends on the -differential pressure across the cladding wall, the cladding taperature, and on the length of time those conditions are main ained. Time duration under burst conditions manifests itself as a heating-ramp-rate effect, and this effect will be treated exolicitly. We have converted differential pressures to hoco stresses to eliminate design-specific dimensional effects. The conversion was made using the thin-shell fornula,
= = (d/2t}aP, where 2 is cladding heap stress, d is the undefe med cladding mid-wall diameter, t is the undeformed cladding thickness, and AP is differential pressure across the cladding wall at rupture. Table 1 shows scme ccmeuted values of hoco stress in terns of differential cressure for cannon consercial fuel designs.
Figure I shows rupture temperature data as a function of hoco stress for a wide range of test ccnditions. While this figure shcws the general trend -- tubes burst at lower temperature wnen the pressure differential is higher -- the data are scattered primarily because of ramo-rate effects and excerimental uncertainities in detamining burst temcerature. 1804 177 0
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Figure 2 shows ORNL data at 28'C/sec (a ccanon ramp rate used in the ORNL experiments) and the basic carrelation we will adcot as developed by Chapman (Ref. Q) using ramerical regression techniques.
It is clear that most of the data scatter has been eliminated by restricting the data to a single rama rate. Chacman has also develooed a ramp-rate correlation (Ref. N) that can be used with the basic ructure-temoerature correlaticn in Fig. 2 to creduce a family of rupture-temperature curves. Ramp-rate has little effect on rupture temcerature for rates faster than 2S*C/sec.
Three curves that scan the imocrtant rame-rate range are shewn in Fig. 3 along with the data of Fig. 1.
Chapman has shcwn that mest of the original scatter is exclained by ramp-rate effects, and the curves in Fig. 3 are seen to span most of the data.
The up-facing triangles still deviate frem the correlations and the major body of data. Difficulties in temoerature measurement for these TREAT in-reactor data (Ref. X) are believed to be rescensible for this deviation, and such discrecancies will be seen in later disclays as well.
3.2 Burst Strain Ceformation (burst strain) at the location of a ructure depends on temcerature, differential pressure (whien is related to temcerature by the ccrrelation in Fig. 3), ramo rate, and several other variables such as local temcerature variations. These effects have been discussed previcusly (Refs. x-y).
Figure 4 shews burst strain as a
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function of one of these variables, burst temcerature, 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 frem less prototypical but more controlled tests to help derive a correlation. Figure 5 shows burst strain versus burst temperature frem Chung and Kassner's work {Ref. T) with short Zircaloy tubes heated by passing an electrical current directly through the Zircaloy. Several fundamental features are apparent.
There are three sucerplastic peaks -- one in the icw temcerature aloha chase around 800 C and two in the hign-temoerature beta enase around 1050'C and 1225 C.
The very important valley at about 925'C is a consecuence of mixed alpha-olus-beta-pnase material, which exhibits icw ductility.
Heating-rate effects are also visable; slow-ramo rates creduce large strains in the temperature regime belcw about 950'C as a result of feedback effects discussed in Refs. x-y.
But slow-rama rates produce very small strains at terceratures greater than about 950t because the Zircaloy has time to oxidi:e and embrittle before significant balleoning can cccur.
Fast-ramo rates creduce the occasite effects in both tamcerature regimes.
To derive the slow-ramo correlation, which is shown in Fig. 6, we have thus taken Chung and Xassner's 5t/sec curve and scaled the peaks and valleys to pass througn the mere crototypical data in our data base. The aloha-chase ceak at 775'C was assigned the value 1B04 184
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of 80% in order to bound Chapman's 10*C/sec bundle test. The five highest points in Fig. 6 (0-10*C/sec heated-shroud single-red tests) are preliminary and have not been fully evaluated, but they were disregarded because the heater power was so law (about 3W) that the tubes were in effect burst in a muffle furnace (the heated shrouds). Direct or external heating methods are known to exaggerate rupture strains by maintaining artificially small local temoerature variations (see Ref. X), and such experiments were excluded frcm our data base. Since the majority of the data is bounded by the curve, we believe that the correlation satisfies the intention of Accendix K not to underestimate the degree of swelling.
It snould be cautioned that some very recent, unevaluated data frem Gemany (Ref. X) also shcw large strains (up to 120%), so the catential exists that Fig. 6 may have to be revised u: ward.
The fast-ramp correlatien is shown in Fig. 7.
In this case, there are no data fem prototypical bundle tests and limited single-red tests with heated shrouds and unifom heaters in the area of the icw-temcerature peak. The correlation was obtained by scaling Chung and Kassner's 55'C/sec curve and adjusting the alcha-phase
- eak height in relation to the peak height in Fig. 6 according to the relation that a 23*C/sec peak would have (based on intereolation) in Chung and Kassner's curve (Fig. 5) to the 5'C/sec ceak in Fig.
5.*
When cratatyoical bundle tests and heated-shroud tests are cer#cmed in the future, we excect the data to fall near the curve in Fig. 7.
Tonsideration is being given to adjusting these cur /e ceak Iccations
- o higher temceratures -- to around 325'C..
1804 187
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Figure 8 shcws the composite (i.e., envelope) of the curves in Figs. 6 and 7 along with all of the data from Fig. 4 The composite curve gives a good representation of the data, providing that the causes of small strains (Ref. X) are kept in mind.
3.3 Assembly Flow Blockage Very few measurements of bundle blockage have been made under prototypical conditions and the best attempts are shown in Fig. 9.
It is therefore necessary to derive bundle blockage from single-rod A
burst strains, but this is not straight fomard, test results have shown that ruptures in a bundle are not coplanar.
Figure 10 is a cross section from Chapman's first bundle test (Ref.
X). Notice that only a few of the rods have burst in this plane.
'de have chosen the most realistic (minimum ficw restriction) of Chaoman's definitions of blockage for the following analysis.
Figure 11 snows the axial distribution of blockage for Bundle No.1, from which the maximum blockage is seen to be 49%.
Figure 12 shcws the geometric relation between average red strain and bundle blockage for a scuare array of comercial-si:e tubes.
From this figure it can be seen that an average red strain of 27%
would cause a bundle blockage of 19%. Since -he average rupture strain for rods in Sundle No. I was 42% (see Apcendix A), the blockage can be obtained from the rupture strain by multiclying by 0.54 (the ratio of 27 to a2) and utilizing Fig.12. The similar ratics for Bundles No. 2 and No. 3 are 0.67 and 0.70 giving an overall averace for the three bundle tests of 0.57.,
o 1804 189
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Assuming that the distributions of ruptures in Chapman's bundle tests are typical, the local blockage correlation is thus formed by multiplying strains in Figs. 6 and 7 by 0.67 and then utilizing Fl a. 1 ?.. We have called this result " local blockage," as distinct from the desired assembly blockage, because it does not yet represent large connercial-size bundles or include the effects of non-fueled tubes, which would not balloon. The slow-and fast-ramp local blockage curves are shown in Figs. 13 and 14 where they are ecmeared with the sparse collection of data. Figure 15 shows the comoosite flow blockage curve, which envelopes the curves in Figs.13 and 14 Finally, to obtain assembly flow blockage, two adjust =ents are required. First, it must be recognized that bundle-average block-age, which is desired, is a function of bundle size. This can be seen by envisicning an 8x8 test bundle that is analyzed quadrant by cuadrant. If each 4x4 quadrant is viewed as a small bundle, the planes of maxuum blockage for the cuadrants would be expected to occur at different elevations because of some randomness of the process. One would therefore expect to find the olane of maximum blockage in each cuadrant to have greater flow restriction than the olane of maximum blockage in the bundle taken as a wnole. That is, the large bundle size introduces an averaging effect.
To account for this effect for ccmmercial fuel bundles ranging from 7x7 (SWR) to 17x17 (FWR), we have used an averace blockage frem Chapman's bundle tests rather than the :.aximum value used in develooing Figs.13 - 15 (that process was appreeriate for the 1804 195
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data ecmparisons because the bundles represented in Fies.13-15 were all small arrays). For Bundle No.1, the average (41%) of the blockages was found between the 23-cn and 47-cn locations in an attemot to eliminate the superessing effect of scacer grids at 10 cm and 65 cm. Similar averages were found for Bundles No. 2 and No. 3.
Using these values the ratio to be used to derive larce-bundle blockages from rupture strain data is 0.55 (ce= cared with 0.67 for small arrays). This factor was used to derive all of the blockage curves in the next section of this recort.
The second adjustment is a reduction of about 5" to account for instrument tubes and guicetubes that would not balloon. The exact scaling factor SF decends on the fuel design and is given by SF = N,.A,j(N A +NA),
7p gg where N is the numcer of fuel reds, A is the flow area areund an g
7 undeformed fuel red, N is the number of guidetubes or instrument g
tubes, and A is the ficw arts arcend an undefamed guidetube or g
instement tube. This scali.7q facter was also emolayed in deriving the blockage curies in the next section. 1804 199
APPENDIX A FUEL CLA0 DING BURST DATA CATA REFERENCE A (Upright Triangle)
FRF-1 R. A. Lorenz, D. O. Mcbson, and G. W. Parker, " Final Report on the First Fuel Roa Failure Transient Test of a Zircalcy-Clad Fuel Rod Cluster in TREAT," Oak Ridge National Lacoratory Report, ORNL-4635, March 1971.
Availacle in public tecnnical liLraries. Also availaole from National Tecnnical Information Ser/ ice (NT!S), Springfield, Virginia 22161.
R. A. Loren:, D. O Mcoson, and G. W. Parker, " Fuel Rod Failure Uncer Loss of-Coolant Conditions in TR. EAT," Nuclear Technology, II, p. 502 (August 1971).
Availacle in puDlic tecnnical libraries.
Incile, 7-roc bundle, steam atmosphere.
Maximum reduction in bundle ficw area = 48 %.
Mean roc curst strain = 26 %.
Mean roa burst temoerature 389*C.
Mean roa engineering curst stress.,1.71 Xcsi.
e RCO PAMP PRES 5URE SURST BURST ENGINEERING RATE AT SURST TEMPERATURE STRAIN BURST STRESS
(*C/S)
(PSIC)
(ac)
(%)
(< PSI)
H 25-36 172 966 26 1.39 4-1 25-36 250 799 35 2.02 R
25-36 205 743 36 1.56 4-2 25-36 290 316 22 2.34 L
25-36 162 315 36 1.31 I
25-36 190 327 35 1.54 C
25-25 215 310 40 1.74 1804 200
DATA REFERENCE B (Cross)
R. H. Cha; man, *Multired Burst Test Program Progress Report for April-June 1977, Oak Ridge National Laboratory Report, ORNL/NUREG/TM-135, June 1977.
Availaole in puolic technical libraries. Also available from National Tech-nical Information Service (NTIS), Springfield, Virginia 22161.
R. H. Chacman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of Creep Time and Heating Rate on Deformation of Zircaloy-4 Tubes Test in Steam with Internal Heaters,* Cak Ridge National Lacoratory Report, NUREG/CR-0343:
ORNL/NUREG/TM-245, October 1978. Available in puolic technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
Out-of-elle, single rod, steam at.mosphere.
E30 RAMP PRESSURE SURST 3URST ENGINEERING RATE AT BURST TEMPERATJRE STRAIN BURST STRESS
('0/S)
(PSIG)
(*C)
(U (KPSI)
PS-1 23 922 393 18 7.47 PS-3 28 809 373 29 6.56 PS-4 28 850 371 21 6.38 PS-6 23 330 882 26 6.72 PS-10 23 870 901 20 7.05 PS-12 28 891 398 18 7.21 PS-14 28 844 883 25 6.84 PS-15 28 893 385 17 7.24 PS-17 23 1760 778 25 14.2 SR-1 23 116 1166 26 0.94 SR-2 28 146 1082 44 1.19 SR-3 28 249 1011 43 2.02 SR-4 23 650 921 17 5.26 SR-5 29 1380 310 26 11.2 SR-7 25 2090 736 20 17.0 SR-3 23 173 1020 43
- 1. 14 SR-13 28 155 1079 79 1.26 SR-15 28 2780 714 la 22.5 SR-17 23 154 1049 53 1.25 SR-19 28 2760 688 16 22.4 SR-20 23 154 1049 55 1.25 SR-21 23 162 1023 18 1.32 SR-22 28 129 1081 50 1.05 SR-23 28 139 1077 35 1.13 SR-24 23 144 1057 67
- 1. 16 SR-25 23 139 1092 73
- 1. 13 SR-26 23 120 1130 34 0.98 SR-27 23 133 1084 41 1.08 SR-23 23 1220 335 27 9.37 SR-29 23 1170 843 27 9.25 SR-37 28 1967 760 23 15.3 SR-38 23 1998 770 20
- 16. 2 1804 201
DATA REFERENCE C (Plus)
MRST-3-1 R. H. Chacman, "Multired Surst Test Program Progress Report for July-Oecember 1977," CaK Ridge National Lacoratory Report, NUREG/CR-0103: CRNL/NUREG/TM-200, June 1978. Availaole in puolic tecnnical libraries.
Also availacle from National Technical Information Service (NTIS), Soringfield, Virginia 22161.
R. H. Chacman, " Preliminary Multf red Burst Test Program Results and Imolications of Interest to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD., Novemcer 7, 1978. Availacle in POR for inspection and copying for a fee.
Out-of-pile, 15-red bundle, steam atmosenere.
Maximum reduction in buncle flow area = 49 %.
Mean rod burst strain = 42 %.
Mean rod strain in plane of maximurt blockage = 27 %.
Mean rod burst temcerature = &as* - Sfo%
Mean rod engineering burst stress = S.72 (psi.
RCD RAMP
?RES3URE SURST SURST ENGINEERING RATE AT SURST TEMPERATURE STRAIN SURST STRESS 1
(*C/S)
(PSIG)
(*C)
( *,' )
(KPSI) 1 29 1124 352 36 9.10 2
29 1075 367 32 3.71 3
29 4
29 1052 360 36 9.33 5
29 1005 372 45 3.14 6
29 1104 872 43 3.94 7
29 1052 369 36 3.52 3
29 1074 372 42 S.70 9
29 1030 370 3.34 10 29 1059 373 45 3.5a 11 29 1054 347 53 3.54 12 29 1114 363 37 9.02 13 29 1091 373 59 3.34 14 29 1066 375 42 3.53 15 29 1C62 365 22 3.50 16 29 1092 348 39 3.35 1804 202
DATA REFERENCE C (Plus)
MRST-B-2 R. H. Chacman, "Multired Burst Test Program Progress Report for July-Decemoer 1977," CaK Ridge National Lacoratory 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 1973," CaK Ridge National Lacoratory Report, NUREG/CR-0655: ORNL/NUREG/TM-297, June 1979. Available in public technical libraries. Also available from National Tecnnical Information Service (NTIS), Springfield, Virginia 22161.
Out-of-pile, 16-red bundle, steam atmosphere.
Maximum reduction in bundle flow area = 53 *4.
Mean rod burst strain = 42 %.
Mean rod strain in plane of maximum blockage = 23 %.
Mean red burst temcarature = 358*C.
Mean red engineering burst stress = 3.38 Kosi.
R00 RAMP PRESSURE SURST BURST ENGINEERING RATE AT SUR57 TEMPERATURE STRAIN BURST STRESS 1
('C/51 (PSIG)
('C)
(*41 (KPSI) 1 29 1117 370 35 9.05 2
29 1115 S46 39 9.02 3
29 1096 353 40 S.38 4
29 1100 372 42 S.91 5
29 1127 866 35 9.13 5
29 1004 357 58 8.13 7
29 1067 361 56 8.64 3
29 1097 856 38 3.39 3
29 10 29 1065 356 43 3.53 11 29 1112 353 40 9.01 12 29 10c4 351 20 3.86 13 29 1134 283 41 9.19 14 29 1048 353 42 3.49 15 29 1152 336 35 9.33 15 29 1117 348 42 9.05 1804 203
DATA REFERENCE C (Plus)
MRST-S-3 R. H. Cha: man, " Preliminary Multirco Burst Test Program Results and Imolications of Interest to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD., November 7, 1978.
Availacle in POR for inspection and copying for a fee.
R. H. Chacman, "Multired Burst Test Program Progress Report for April-June, 1979,* Cak Ridge National Laboratory Recort, NUREG/CR-1023: ORNL/NUREG/TM-351, in cuolication.
Out-of pile, 16-rod bundle, staam atmosphere.
Maximum recuction in bundle flow area = 75 "..
Mean red ourst strain = 57 %.
Mean rod strain in plane of saximum blockage = 10 %.
Mean red burst temcerature = 764*C.
Mean rea engineering burst stress..,11.07 Kosi.
R00 RAMP PRESSURE BURST SURST ENGINEERING RATE AT SUR5T TEMPERATURE STRAIN SURST STRESS 1
(*C/5)
(PSIG)
(*C)
( *.)
(KPSI) 1 10 1393 771 18 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 5
10 1327 770 61 10.75 7
10 8
10 1320 756 73 10.69 9
10 1220 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 1486 760 29 12.04 la 10 1405 769 a2 11.38 15 10 1335 753 53 10.31 16 10 1a07 747 59 11.20 1804 204
DATA REFERENCE D (Closed Circle)
F. Ertacher, H. J. Neit:el, and K. Wiene, " Interaction Between Thermanydraulics and Fuel Clad Ballooning in a LOCA, Results of RESEKA Multirod Burst Tests with Flooding," paper presented at the 6th NRC Water Reactor Safety Researcn Information Meeting, Gaithersburg, MO, Novemcar 7,1978. Availaole in file for USNRC Report, NUREG-0536.
F. Ercacher, H. J. Neit:el, M. Reimann, and K. Wiene, " Fuel Rod Senavior in tne Refilling and Reflooding Phase of a LOCA-Burst Test with Incirectly Heated Fuel Rod Simulators," pa:er presented at the NRC Zircaloy Cladding Review Grouc Meeting, Idaho Falls, May 23, 1977. Availaole in file for USNRC Recort, NUREG-0536.
K. Wiehr and H. Scnmidt, "Out-of-Pile Exceriments on Ballooning of Zircaloy Fuel Rod Claddings Test Resul;.3 with Shortened Fuel Rod Simulators,"
<ernforschungs:entrum Karlsruhe Report, KfK 2345, Octocer 1977. Available in file for USNRC Report, NUREG-0536.
F Erbacher, H. J. Neit:el, M. Reimann, and K. Wiehr, "Out-of-Pile Experiments on Ballooning in Zircaloy Fuel Rod Claddings in the Low Pressure Phase of a Loss-of-Coolant Accident," Proceedings of Saecialists' Meeting on the Sehavior of Water Reactor Fuel Elements Under Accident Conditions, Scatind, Norway, Septemcer 13-16, 1976. Available in puolic tecnnical libraries.
~
Er acher, H. J. Neit:el, and K. Wiehr, " Studies on Zircaloy Fuel Clad Sailooning in a LOCA, Results of Surst Tests with Indirectly Heated Fuel Rod Simulators," pacer oresented at tne ASTM 4th International Conference on Zirconium in tne Nuclear Industry, Stratford-on-Avon, England, June 27-29, 1978. Available from ASTM.
Cut-of pile, single rod, air and steam atmospnere.
R00 RAMP PRESSURE SURST SURST ENGINEERING RA7E AT SURST TE.MPERA7URE STRAIN BURST STRESS
_f_
('C/S)
(PSIG)
(SC)
( *.' )
(< PSI) 11
?
380 27
?
?
11 356 380 51 5.91
?
11
?
365 23
?
11
?
360 J
?
?
11
?
30 32
?
?
11
?
30 36
?
?
11
?
340 13
?
11
?
340 54
?
11
?
330 17
?
?
11
?
925 27
?
?
11
?
325 33 7
'.3 11 1420 323 33 9.31
?
11
?
320 23
?
?
11
?
320 32
?
14
'1
'.420 310 38 3.31 1804 205'
OATA REFERENCE D (Continued)
R00 RAMP PRESSURE SURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
_f_
('C/S)
(PSIG)
(*C)
( *. )
(KPSI)
?
11
?
810 42
?
?
11
?
810 14
?
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 7
?
11
?
755 52
?
1804 206
DATA REFERENCE E (0 pen Circle)
E. Karb, "In-Pile Experiments in the FR-2 OK-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. Karb, "Resulcs of the FR-2 Nuclear Tests on the Behavior of Zircaloy Clad Fuel Rods," paper presented at the 6th NRC Water Reactor Safety Research Infor-mation Meeting, Gaithersburg, MD, NovemDer 7,1978. Availacie in file for USNRC Report, NUREG-0536.
E. H. Karb, "Results of FR-2 In-Pile Tests on LWR Fuel Rod Behavior," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idano, June 22-29, 1979. Available in POR for inspection and copying for a fee.
Inpile, single rod, steam atmospnere.
RCO RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
(*C/S)
(PSIG)
('C)
(*)
(XPSI)
A1.1 7.1 725 810 64 5.01 A2.1 20 1276 820 36 8.82 al. 6 8.2 1160 825 38 8.02 33.1 10 1146 825 37 7.92 Bl. 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 Bl. 5 9
653 910 60 4.51 S1. 2 8.7 653 915 25 4.51 33.2
- 12. 1 725 915 50 5.01 1804 207
DATA REFEREN'CE F (Square)
R. H. Chacman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of Creep Time and Heating Rate on Oeformation 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 atmosonere.
RCD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS 1
(*C/S)
(PSIG)
('C)
(U (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-a3 4
1105 773 29 6.95 SR-44 5
1060 777 30 3.59 5R-41 9
1416 757 27 11.5 SR-42 10 1373 761 28 11.1 1804 208
DATA REFERENCE G (Aster 4 a REBEXA-1,
-2, -3 F. Erbacher, H. J. Neit:el, and X. Wiehr, " Interaction Between Thermohydraulic and Fuel Clad Ballooning in a LOCA, Results of RESEXA Multired Burst Tests with Flooding," paper presented at the 5th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD., Novemoer 7,1978. Available in file foi USNRC Report, NUREG-0536.
K. Wiehr, "Results of RE3EXA Test 3," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979.
AvailaDie in POR for inspection and copying for a fee.
Cut-of pile, 9-rod bundles, steam and water atmosphere.
TEST INITIAL MEAN MEAN MEAN MEAN REDUCTION RAMP PRESSURE SURST BURST ENGINEERING IN FLOW RATE AT SURST TEMPERATURE STRAIN BURST STRESS AREA 1
(*C/S)
(PSIG)
(*C)
(%)
(KPSI)
(%)
1 7
870 315 29 6.01 25 2
7 300 370 53 5.53 60 3
7 725 830 14 5.05 52 1804 209
DATA REFERENCE H (Inverted Triangle)
M. Bocek, "FABIOLA," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Senavior Information Exchange, Idaho Falls, Idano, June 22-29, 1979. Available in POR for inspection and copying for a fee.
Out-of pile, single rod, steam atmoschere.
RCD RAMP PRESSURE BURST SURST ENGINEERING RATE AT SURST TEMPERATURE STRAIN BURST STRESS
_f_
('C/S)
(PSIG)
(*C)
(%)
(KPSI) 1 3
563 860 66 3.92 4
11 1375 790 8
9.58 8
7.3 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 1804 210
OATA REFERENCE I (Diamond)
J. L. Crowley (CRNL), personal ccmunication to D. A. Powers (USNRC),
August 10, 1979.
R. H. Chapinan (ORNL), personal comunication to D. A. Powers (USNRC),
Septemoer 11, 1979.
Out-of pile, single red, heated shroud, staam atmosphers.
R00 RAMP PRESSURE BURST MAXIMUM ENGINEERING RATE AT BURST TEMPERATURE R00 STRAIN SURST STRESS 1
("C/S)
(PSIG)
('C)
(%)
(KPSI)
SR-47 10 1436 775 e 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 R-57 0
725 775 110 6.23 1804 2\\\\
GENERAL APPENDIX X REQUIREMENTS MUST BE MET.
REVISED MODELS MAY BE REQUIRED ?OR ALL VENDORS.
ALL BREAK SIZES NEED TO BE CONSIDERED.
IF UNCERTAINTIES ARE NOT CONSIDERED IN SWELLING AND RUPTURE CURVES, APPROPRIATE SENSITIVITY STUDIES MUST BE PERFORPED.
WIDTH OF THE VALLEY MAY BE AS IMPORTANT AS HEIGHT OF THE PEAK.
1804 212
DETERMINING MAGNITUDE NEED TO SORT OUT SUBSTANTIVE CONDITIONS.
SOFI TEMPERATUFIS AND RAMP RATES MAY NOT BE EXPECTED.
THEREFORE MODELS NEED ONLY APPLY WHERE CONDITIONS ELLL OCCUR.
PWR RUPTURE TEMPERATURES 840 *C - 960
- C BWR RUPTURE TEMPERATURES 960 C - 1200 *C
[804 213
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1804 216
IMPORTANT PARAMETERS CLADDING TEMPERATURE FUEL TEMPEPATURE SURNUP (FISSION GAS)
DIMENSIONS PLENUM TEMPERATURE POWER P uSTIC STRAIN HEAT TRANSFER v
v i
PIN PPISSURE (STRESS)
PAMP PATE 1
RUPTURE TEMPERATUPE l' s
/
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BLOCKAGE 4
t CLADDING DIISSIONS FLOW AREA PADIATI0fj!
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s, s,
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+
1804 217
+
laor.; anEAK REF100D P'dR REFLOOD AT FLOODING RATES LESS THAN 1 IN./SEC. APPEARS TO BE WORST CONDITION BECAUSE OF APPENDIX K REQUIREMENTS FOR STEAM COOLING AND 3 LOCKAGE.
NRC PERFORMED LIMITED SENSITIVITY STUDY ON BLOCKAGE, STRAIN, AND INCIDENCE OF RUPTURE.
1804 218
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