ML20154N217
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l 6
SUPPORTING INFORMATION FOR VECTRA CALCULATION PACKAGE REVISION 1 NUH004.0409 Questions 7.1,7.2,7.3 Prepared For VECTRA TECHNOLOGIES, INC.
o 6203 San Ignacio Ave. Suite 100 San Jose, CA 95119 Prepared By Dr. E.R. Gilbert Mohr and Associates 1440 Agnes Richland,WA 99352 On May 19,1997 9810210212 981009 3-d-
C E -
Table of Contents Summary _
... 1 7.1 Justification for the Use of the PNIe6189 Methodology and Data.._
2 7.1.1 Storage Temperature and Internal Fuel Rod Pressure._
....... 3 7.1.2 Total Elongation -
4 7.1.3 Deformation Mechanisms
-................ 5 7.1.4 Fracture Mechanisms.:
-...._...........6 7.2 Bounding Fuel Types (r/t & Pressure)
............. 7 7.3 Justify Conservatism of Selected Cladding Oxide Thickness......-....
.... 10 Re fe ren c es..
............................ 1 1
^
Figu res
............ 15 N
Mohr and Associates Rev.0: May 19,1997 SUPPORTING INFORMATION FOR VECTRA CALCULATION PACKAGE REVISION 1 NUH004.0409 Questions 7.1,7.2,7.3
SUMMARY
The objective of this review is to provide additional
- formation to enhance the technical m
justification provided in VECTRA's Calculation NUH004.0409, Revision I for inened dry storage (IDS) of spent nuclear fuel (SNF) after achieving burnup levels to 55,000 mwd /MTU. A technical basis for IDS of SNF, sometimes referred to as the PNL4189 methodology, was developed for SNF with a nominal burnup level f 33,000 mwd /MTU and included the results of limited testing with SNF with a burnup level of 40,000 mwd /MTU. The results of that study were presented by Levy et al. (1987).
The concern with applying the PNL4189 methodology to high burnup SNF is that high burnup SNF may not meet the criteria applied by the PNL4189 methodology for resistance to gross degradation during IDS. Relevant considerations include observations by Yang (1997) that the experimental data based on testing SNF cladding after both low and high burnup show large data scatter. This data scatter has been attributed to hydrogen concentration and distribution, sample O
needed to obtain representative data for in-reactor application.
size and preparation, and measurement techniques. Better characterized measurements are A review of the high burnup SNF cladding data indicates that the PNL4189 methodology is applicable to SNF with burnup levels to 55,000 mwd /MTU, because the mechanisms of defonnation and fracture do not change and the cladding retains adequate strength, ductility, and resistance to creep. Changes in strength and ductility that result from irradiation ofZircaloy cladding at high bumup are attributed to effects of temperature, time, and irradiation on the precipitation and dissolution of alloying elements from precipitates into the matrix (Mahmood et al.1997). These changes in strength and ductility occur without changing the basic mechanisms controlling deformation and fracture Experimental data for high burnup SNF 'mdicate that hydrides do not affect the deformation or fracture mechanisms. Corrosion-induced hydrides up to 370 ppm do not affect the strength and ductility of either nonirradiated (Huang and Huang 1994, Bai et al.1991) or irradiated Zircaloy (Wisner and Adamson 1996) for neutron fluences up to 8.5 x 10 n/cm (E>l MeV) (Mahmood et al.1997). Although there is considerable variability in the 2
2 correspondence between neutron fluence and rod average and local burnup level, this neutron fluence corresponds to a burnup level of approximately 55,000 mwd /MTU. Most SNF with burnup levels below 55,000 mwd /MTU contain less than 370 ppm excess hydrogen, that which forms hydides. The J, type fracture toughness testing reported in the literature always results in i
completely ductile fracture of the specimen (Bai et al.1991, Mahmood et al.1997).
Page 1 of 27
mnr.o--
Rev. 0: May 19,1997 i
j Determination ofthe applicabdity ofthe PNL-6189 methodology to burnup levels above 1
55,000 mwd /MIU requires additional data and analyses, Wally forZirat / < enntaining i
high tin. The primary concern is that the effect ofhydrides on the ductdity ofSNF al Miag j
above 55,000 mwd /MTU may reduce the total stram for al-Ming breach by more than the j
factor of ten assumed by the PNL-6189 methodology. The reduced failure stram appears most i
severe in high tin Zircaloy 4. R==nea low tin zirconium and improved zuconium alloys are l
affected less by corrosion at high burnups, additional analyses may be able to justify the use of j
the PNIA189 methodology beyond 55,000 mwd /MTU for these i-M alloys.
i The section numbers in this manuscript are consistent with relevant sections in the calculation reportby VECTRA (1996).
7-1 Justification for the Use of the PNIA 189 Methodology and Data i
j Concerns regarding the use ofPNIA189 methodology and data anu. aom using this approach to j
evaluate the strength charactenstics of spent nuclear fuel (SNF) to a burnup levd of l
55,000 mwd /MTU. We recogmze that the origmal data base used to develop the methodology was developed based on data obtamed from SNF with nommal bunup levels of approxunately
)
33,000 mwd /MTU and some data with burnup levels to 40,000 mwd /MTU. Currently an j
equivalent data base for SNF to burnup levels of 55,000 mwd /MTU are not avadable to us; l
however, some data are avadable on non-fueled cinMing and several types ofmore limited data l
on fueled cladding out to >55,000 mwd /MTU that strongly suggest that the methodology in the i
PNIA189 can be used to support the evaluation of the storage of SNF with burnups to 1
1 55,000 mwd /MTU.
i l
The PNIA189 methodology is based on the ability of the cladding to resist gross degradation in l
an inerted dry storage atmosphere. N analysis indicated that the most likely meham== for i
cladding breach was stress rupture by creep inriimi by the storage temperature and the internal l
fuel rod pressure. N conditions forcalcaleng rupture =<==nari an order ofmagnitude j
decrease in total elongation from inadiation embrittlement of the claMing h mehanient for j
creep and rupture were based on deformation maps and fracture maps. A pmbole breach was I
identified as the fkacture type. The conservatism of the PNI4189 methodology can be i
consideredjustified for high burnup SNF, if the mehaniens do not change at high bunrup and if i
adequate resistance to gross degradation dunng inerted dry storage is maintamed.
Conservatisms in the calculations included no credit for the cooler ends of the SNF while j
residing in the storage containers, and no credit was taken fbr the improved and af mi=i i
3 Pay 2 of 27 I
. ~
Mohr ami Mana===
Rev. 0: May 19,1997 eladding alloys that show superiorperformance to Zircaloy-4 SNF al Ading A review ofnew high burnup data strongly suggest that the methodology and data in PNI4180 are applicable to calm +ing conservative storage temperatures for SNF with burnup levels to
$5,000 mwd /MTU. New data and madaling indicate that the creep rate conamm to be retarded by irradiation at high bw, the strength renamn high; except fora few specunens the total elongation remama within the criteria ofPNIA189; and the da:a are consistent with the contmustion of the same contmiling deformation and fracture r=ach==iema that are not changed byirradiation to high burnup.
7-1.1 Storage Temperature and Intemal Fuel Rod Pmssure The cladding stress is derived from the cladding dimensions and the fuel rod pressure. These terms are rested as calculational parameters in the methodolgy and have been addressed by VECTRA (1996). The storage temperature is a methodology output and has been modeled and ga&d by VECTRA (1996). The al=dding stress versus temperature r-z-+1in VECTRA's Table 4.3 is presents the SNF cladding stress as a function of the fuel rod temperature for the g
fuel rods at the specified burnup levels in storage, and is based on calculations usmg the prearradiation fill gas pressure plus the pressure increase fmm fission gases released from the fuel pellets during irradiation as inputs.
VECTRA's Table 4.4 gas the allowable storage temperature limits computed with the DATING code and also includes the effects of cooling from decay as c-=M by the ORIGEN Code. The results in VECTRA's Table 4.4 show that increasing the burnup of the SNF reduces the initial storage tempemture limit. This reduction is caused by the higher gas pressure calculated at high burnup and by the slower cooling rate of the high burnup SNF.
The bounding conditions in this review were found to be for the Wa=ria hause 15x15 fuel type in a B&W reactor system as indicated in Table 2 of this report. This fuel type was calenta+ad to have a value ofde of 8.28 based in intial dimensions in Painter et al. (1994). As indicated in Table 2, this fuel type is also indicatad to have the highest value formaximum peak LHOR and maximum fuel temperatme, as indicated in Table 2. The He fill pressure was not provided by Painteret at (1994).
1 i
m
.-.,y
Mohrand Assocuses Rev. 0: May 19,1997 7-1.2 TetalElongation Test data used in the PNL4189 methodology included postindistion creep measurements durmg storage ofPWR fuel which hadbeen irradi=*ad o aburnup of33 mwd /MTU. The t
maxunum corrosion oxide thickness was 40 micmns. The combmed prepressure and fission gas pressure with the sacrage temperature generated el Ming creep strams up to 0.47% durmg a dry storag: demonstrations without cladding breach (Porsch, Fleisch, and Heits 1986).
Also included in PNL-6189 were the results of testmg a PWR fuel assembly irradi=ad to 40 mwd /MTU (Pechs and Fleisch 1986). They also reported the results ofcreep tests cWW cn Zircaloy-4 cladding tubes at 400*C after bemg irradiatad to a fast neutron fluence of 3.9 x 2
!? n/cm. O=dding creep strams up to 0.3% were achieved; no failure were observed.
ON < creep test results for Zircaloy-4 fuel cladding tubes irradiated to 4.5 x 10 : n/cm 2
2 (E> 1 MeV) and tested at 380*C at cladding stresses of 360 and 418 MPa resulted in creep strains up to 0.8%; no failures were observed (SchafDer-Le Pichon 1997). These data are shown in Figure 1. Strams in excess of 2.5% were achieved in biaxial tensile test apacimane irradiated 28 2
to 8.5 x 10 n/cm. (E>l MeV)(see Figure 2). The creep behavior fmm these testn are compared with the results ofthe Pechs and Fleisch (1986) investigations in Figure 3. In mahng these compensons, undefined constants, X, and t, were assumed to be unity dimensional o
constants. Figure 3 se s PNL-6189 meth:dology and is reproduced from PNL-5998.
j r
Figure 4 is provided to show the emig-- '=#verm SNF burnup and the fast neutron fluence exposure for cladding test articles. The neutron fluence of 8.5 x.10 8 n/cm (E>l MeV) 2 2
corresponds to a local burnup values in the range of 45,000 to 55,000 mwd /MTU. Therefore, the post irradiation creep results for these tests with neutron f!=wwa cuirepueding to high burnup are in consistent agreement with comparable lower fluence data and the methodology of PNL-6189.
Total elongation forburst tests conducted in the temperature range of 599 to 652*F on Zircaloy-4 SNF aladding =pacimane with burnup levels up to 60,000 mwd /MTU are simnnarundin Figure 5. These data show that the total elongation generally mains above 10% of the range for nomrradianad tests. Data for tensile tests show comparable trends.
I Page 4 of 27 s
J e
r v
my,-
~.
J l
Mohrand Assoames Rev 0: May 19,1997 1
i The results ofring tenmla tests on Zircaloy 4 irradiated to high values ofneutma fluence. 8.9 x 8
2 10. n/cm (E>I MeV) are sunmuinzed in Table 1. These results demonstrate that high burnup cladding retams strength and duculity (Van Swam et al.1997). The reduction in total elongation caned by high burnup is less than a factor of ten.
W
]
Table 1.
Results of Ring Tensile Tests CnaAW on High Burnup Zircaloy-4 l
Cladding Specunens at 350'C (VanSwam et at 1997)
- Burnup, Neutron Fluence, Oxide Ultunate Total GWd/MTU 102 n/cm Thickness, Tensile Stress, Elongation l
2 j
(E> 1 MeV) microns MPa i
0 0
0 417 24 t
I 47 7
104 666 21 f
47 7
186 611 5
j 57 8.9 105 652 17 57 8.9 290 602 8
i 7-1.3 Deformanon Mechanisms 1
2 Changes in.,Loogth and duculity that result from irradiation of Zircaloy cladding at high burnup are attributed to effects of temperature, tune, and inadiation on the precipitation and dissolution of alloymg elements from precipitates into the matrix (Mahmood et at 1997). These el=g1:in
.Loogth and duculity occurwithout changmg the basic mecb=mmma controlling deformation and 2
fracture. The J, type fracture ton %s testing reported in the literature alway result in i
completely ductile fracture of the specimen (Bai et at 1991, Mahmaad et at 1997).
4 i
A review of the We ofirradiation indeed hardening on creep indicates that the effects 1
ofinadiation hardemng on creep tends to saturate above a fast neutron fluence ofapprovima+=1y
}
4 x 10 n/cm and the creep rate may tend to increase above 3 x 1028 n/cm. The specific values 2
2 j
of fluence where these changes take place may depend upon the neuten flux and energy 1
i Page 5 of 27
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4 --
i 6
~,, -, -. - -
c l
t I
j Mohrand A======
j Rev. 0: May 19,199"I i
l whum. These effects are attributed to changes in microstructme(Fidleris 1988). Additional l-detads on microstructural a- =g-n durmg inadiation ofzirconium alloys have been provided by investigations by Ells and Willimma (1974) and by Gnffiths, et at (1987).
j MWiing ofcf= Ming creep after high burnup uradiation by Schemer-Le Pichon (1997) show-l agreementwith m - 2 3 buonal tensile,andinaxial creepdatac*=d onin=diated j
CWSR Zircaloy-4 claMing tubes. The results of theirmWIiag and v.- J 1 data show
}{
inadiation hardemng and a decrease in creep deformation due to inadiation. The tests were conducted in the temperature range ofdry storage, namely 350,380, and WC. The neutron fluences ranged from 0 to 8.5 x 10" n/cm (E>l MeV). The exposure in terms ofburnup was 2
l not provided by the authors, but this fluence conesponds to burnup levels in the range of 40 to l
55 GWd/MTU (see Figure 4). Agreement ofcreep data from these tests with data used to
}
develop the PNI4189 methodology is shown in Figure 3.
i 1
l Experunental data for high burnup SNF indicate that hydrides do not affect the deformation or fracture mechammma. Corrosion iadac~i hydrides up to 370 ppm do not affect the strength and ductdity of either nouirradiated (Huang and Huang 1994, Bai et al 1991) or inadiated Zucaloy
}
(Wisner and Adam <an 1996) fbr neutron fluences up to 8.5 x 10" n/cm (E>l MeV)(Mah=aad 2
l et al 1997).
i i
Because the creep and deformetion properties of the oprimi=d and improved alloys for avtandad l
BU are superior to those for Zircaloy-4 (Sabol et al.1997), the methodology of FNL4189 should j
also be applicable to those when evaluated as SNF claMing i
4 7-1.4 Fracture Mechanisms I
The fracture mehamam in the PNIA189 methodology results in the development of a pin hole
]
breach which vents the internal gases. When the gas is vented, additional mechanical damage is j
rme and fuel particulates are confined by the claMbg Tensile burst data are generated j
with high stresses and strain rates in regions of deformation and fracture meheni=== that are differeng then forstress or-o. The fracture mechammma fbrtensile andburstdataresultin longitudinal cicMia cracks, rather than pin holes. Even though these dsta are not generated in the same region ofdeformation and fracture mehmaien, they demanrtrate that the melanical insegnty of the alaMing remains high after bemg irr=Alad to high values of t=====
Page 6 of 27
__._______________.__..g.,
i MohrandAssocuses Rev.0: May t9,1997 l
l The experimental data based on testing SNF d= Ming afh:r both low and high bumup show large I
data scatter. This data scatterhas been attributed to hydrogen concentration and distribution, sample size and preparation, and measurement whaiques. Better drum :mi measurements arr needed to obtain wr---
- ve data forin-reactorapplication (Yang 1997). While the data generally show retention ofstrength and ductility, the total elongation appears to decrease with increasingvalues ofburnup.
I Pin hole breaches at locations ofincipient defects have led to releases of krypton durmg perfonnance ofdry storage with SNF (Johnson et al.1987, McKinnon et at 1988, Gilbert et al.
1990). The qu.ui.iy of kryrou released was typical of the krypton that wouldbe WM to be coctained in a single fuel rod. The releases were usually observed immediately followmg fuel l
handling. These types ofreleases are also PM to occur dunng dry storage of high burnup l
SNF.
Improwd and optimi=i alloys for a* ading burnup performance are showing improved j
resistance to corrosion and improved mechanical properties. Application of the PNI4189 l
methodology to these improved alloys is therefore, more conservative than application to SNF clad with Zircaloy 4.
1 The effects of burnup on the total elongation and burst stress fortests with SNF cImMing coaded in the temperature range of 599 to 652*F are shown in Figures 5 and Figure 6 (Garde 1986, Smith 1994, and Dideon 1983). 'Ibese data are for Babcock and Wilcox 15x15, and Westinghouse 15x15 types of SNF. The trend is for both the burst stress to increase and the total elongation to decrease with incressmg burnup. However, the burst stress is >600 MPs and provides a large margm to failure when cuiur-mi to cladding stresses of<l50 MPa durmg storage as indicated in VECIRA's Table 4-3.
1 7-2 BoundingFuelTypes (r/t& Pressure) 1 l
ClaMing design parameters for PWR fbel types are smnmarund in Table 2 based on information l
in Pamter et al. (1994). Of these design, the bounding fuel type is used for the calculation of limiting storage temperatures in VECTRA's Table 4-4 forPWR fbei designs. The bounding i
pressures and stresses are also r--- 2--iin Table 4-1 for PWR and Table 6-1 forBWR fuel designs. The value ofdt, by itselfdoes not dmEna the bounding fuel type. The value of r is Page 7 of 27 h
w-w+
,w
-p
Mohr and >==
Rev. 0: May 19,1997 taken as the radius to the el Miar midwall The value ofthe el-Ming stress defina= the bounding fbel type. Even though the values ofr/t, by itselffor these fhel types may not hauadia-the stress is based on the product of the fa'arnal fuel rod pressure and rit. LWon of Table 2 indicates that the 15x15 fbel type is bounding harsina it has the largest value ofr/t, the highest values ofLHGR and fbel tamnar=+nre which increases release of fission gas, e.g. the 15x15 PWR fuel design is bounding har=ne of the large value ofdt, and the higher pressures 1
pradneM by the higher fuel temperatures associated with the higher linear best generation rate (LEGR) at which this design is operated.
I Differentvalues ofnn/ tare r=-1 in Table 2. The different values are associated with j
corrosion wastage from oxidation of the cleMing external surface. Because the oxide has a lower theoretical density than Zircaloy 4, it replaces a claMing wall thickness ofonly 2/3 of the value of the oxide thickness (Porsch, Fleisch, and Heits 1986). Based on the Pilling-Bedworth factor of 1.75 (VanSwam et at 1997), which includes a correction for the 89 to 90% dense i
oxide, the 140 micron oxide thickness cogsposds to a reduction in el= Ming metal thickness of 80 microns. Therefore an oxide thickness of 140 iscrons corresponds to a reduction in Zircaloy 4 wall thickness of 80 microns. The oxide thickness at the inner claddmg surface is less than 7 microns (Mitsugi, Knahida and Kikuchi 1997) and is ignoredhacame it is wrthm the t
variabdity of the outer oxide thickness..
Cladding design parameters for BWR fuel types are smnmartzed in Table 3 based on information in Pamter et at (1994) and in Johnson, Gilbert, and Guenther (1983).
l 1
Page 8 of 27
f f'
~
f%
b t
MM8 M
. Rey.0: May 19,1997 Table 2. Design Parameters for Zircaloy-4 Clad PWR Fuel Types (Painter el al.1994; Johnson, Gilbett, and Quenther,1983) and r/t After Different Amounts ofClad Thinning by Ovidatian I
FtsetYemdar W
W W
W W
FinalRod Array 14xl4 15x15 17x17 17xt7 14xl4 15x15 14:14 15x15 17x17 15x15 17xl?
Max local Expoewe, GWdMlU 50 55 50 47.5 r/l 7.96 8.28 7.33 7.40 7.36 6.57 8.18 8.I4
'7.81 7.61 7.63 f
(r-40py(t-80p)'
8.99 9.46 8.41 8.44 8.23' 7.28 9.33 9.33 8.57 8.96 8.73 (r-50py(t-100p)2 9.29 9.81 8.74 8.75 8.48 7.48 9.67 9.67 9.36 8.36 9.06 l;
(r-70py(t-140p)*
9.97 10.60 9.48 9.45 9.04 7.93 10.44 10.44 9.48 10.12 9.8) j Max. Peak L.HGR, kW/A 18.8 18.8 15.2 16 3 15.8 16.3 12.9 12.6 17.4 14.7 Max. Fuel Temp. *C 2340 2290 2140
'2200 2260 2340 1870 2300 2020 435 300to 290 He FillPresswe,peig 275 to 400 450 l
80 saecross is a p===m value for optanuzed Zarconsumi alloys at 55 MWdMlU (Woods and Kisager 1997).
[
I a
100 sucross includes most Zircoloy-4 claddag for tuaraup values to 55 mwd /MTU (Kilp et al.1997).
l 3
140 ancross Me Zircaloy-4 claddag for buraup values to 55 mwd /MTU (Kilp et al.1997).
j i
Page 9 of 27 l
f 1
I
Mohr and Associates Rev. 0: May 19,1997 Table 3.
CW== Design Parameters for Zircaloy-2 Clad BWR Fuel Types (Painter et al.
1994; Johnson, Gilbert, and Guenther,1983)
FuelVendor GE GE GE SNP SNP Reactor System GE GE GE GE GE Fuel Rod Array 7X7 8X8 8x8 9X9 9X9
^
Max Local Exposure, GWd/MTU 40 45 40 55 r/t 8.30 7.05 6.75 6.57 8.12 (r-50p)/(t-100 )
9.39 7.97 7.57 7.48 9.55 Max. Peak LHGR, kW/ft 18.5 13.4 18.8 15.2 Max. Fuel Temp. *C 2440 1890 1830 2040 1^
He Fill Pressure, psig 0
30 0
60 60 i
1 7-3 Justify Conservatism of Selected Cladding Oxide Thickness (60 microns [0 0024 in.])
i The trend ofincreasing oxide thickness with increasing bumup for Zircaloy-4 is illustrated in Figures 9 and 10. These data indicate that the maximum oxide thickness to be expected at 55,000 mwd /MTU is less than the estimated maximum thickness of 140 microns 60,000 mwd /MTU cited by the expert concensus as noted by NRC staff. For conservatism the value of 140 microns can be assumed as the bounding value for a bumup level of 55,000 mwd /MTU.
Advanced alloys for high burnup cladding such as "ZIRLO" show even less oxidation and I
thirner oxides (Sabol et al 1977). A summary of the bounding oxide thickness for different zirconium cladding alloys at a burnup of 55,000 mwd /MTU is presented in Table 4. Plots oxide thickness for Optimum Zircaloy-4, ZIRLO, and Duplex Cladding are shown in Figure 11,12, and 13.
t 4
- 5 Page 10 of 27
)
.a a
1-
Mohranu--
Rw. 0: May 19,1997 Table 4.
Bounding Madding Outer Oxide Thicimens forZirconium Cf=dding Alloys ataBurnup of55,000 mwd /MTU Bounding Dadding Wall Reference Oxide Thickness Thickness, Reduction,microm CladdingAlloy microna Conventional Zircaloy-4 140 80 Kilp et al.1991, Sabol et at 1997 Opeimi=d Zircaloy-4 80 46 Woods and Klia==c 1997 ZIRLO 60 35 Sabol et at 1997 Duplex Cindding 50 29 Woods and Klinger 199?
REFERENCES Bai, J., C. Prioul, J. Pelchat, andF. Barcelo. 1991. " Effects ofHydrides on the Ductile-Brittle Transition in Stress Relieved, Recry=*=llin d, and Beta Treated Zircaloy 4 " Pracaadia== of l
International ANS/ ENS Conference, p. 223, Avignon, France, AHfi, LaGrange Park, Illinois, e
Chm, B.A., M.A. Khan, and J.C.L. Tarn. 1986. Deformation and Fmeture Man Mathadaloav for Pr~iietian Cinddina Behavior Durine Dry Stormaa PNL-5998, Pactfic Northwest Laboi.Eny, Richland, Wa=hinatan.
Dideon, C.G.,1983. Fuel Performanca imdae htandad-Burmm Oner=tiaa B&W 15x15 3
Roman. DOE /ET/3412, U.S. Department of Energy, Washmston, D.C.
Ells, C.E., C.E. CaI-==a and C.D. Willimma 1974. "The Temperature Danaadaaec of Irradiation Damage in Zirconium Alloys," the Saenad Sv - - h-on Maahanie=1 hhaviour of 4
Matarial= hald in Kvoto. Innan Vol 2 Society ofMaterials Science of Japan, Japan.
Fidleris, V. 1988. "The Irradiation Creep and Growth Phaanmaa= " J. Nuct Mater. Vol 159, pp.1-21, Elsevier Science Publishers, Amsterdam, North-Holland.
Page 11 of 27
_. - - -.. ~..
Mohr and M=aa=e==
Rev. 0: May 19,1997 Garde, A.M. 1986. HotCel[Framinntion ofFrtendert Burnup Fuel From Fort Calhoun.
DOE /ET/34030-11, U.S. Department ofEnergy, Wanh'r.g:., D.C.
Gilbert, E.R., W.J. Bailey, A.B. Johnson, Jr., and M.A. McKinnon. 1990. " Advances in Technology Ibr Storing LWR SpentFuel," Nuclear Tach =alonv Vol. 89. American Nuclear Society,LaGrange Park,Blinois.
Griffiths, M., R.W. Gilbert. V. Fidleris, R.P. Tucker, and R.B. Adamtan. 1987. " Neutron Damage in Zirconium Alloys Irradiatad at 644 to 710K,"I Nucl Ma+=r Vol 150, pp.159-168, Elsevier Science Publishers, Amsterdam, North-Holland.
Huang, J.H. and S.P. Huang. 1994. "Effeet of Hydrogen C-on the Mechanical Fr@iles ofZircaloy-4,"1 Nucl Ma+=e Vol. 208, Elsevier Science Publishers, Amsterdam, North-Holland.
i 1
Johnson, A.B., J.C. Dobbins, F.R. Zaloudek, E.R. Gilbert, and LS. Levy. 1987. Assessment of the ineaarity of Snent Fuel Ae=amhlies Uemd in Dry Stornan C_.. -..w.. ;.'ane at the Nev=An Test Sim. PNL-6207, PaciHc Neia Labocisay, Richland, Wanhia=*aa Johnson, A.B., E.R. Gilbert, and R.J. Guenther. 1983. Behaviorof Spent Nuclear Fuel and Stor=== System comns.w..a in Dry interrm Stornoe. PNL-4189 Rev.1, Pacific Northwest Laboratory, Richland, Wanhi.t;.--
Kilp, G.R., M.G. Balfour, R.N. Stanutz, K.R McAtee, R.S. Miller, L.H. Ramaa. N.P. Wolfhope, O. Ozer, and R.L. Yang. 1991. " Corrosion Experience with Zircaloy and ZIRLO in Operstmg PWRs," in Volume 2 Pir- = E== In*=enarianal Tonical Mena on LWR Fuel Performarw a Avignon, France, ANS, I4 grange Park, Blinois.
Levy, LS., C.E. Beyer, B.A. Chin, E.R. Gilbert, E.P. Simonen, and A.B. Johnson, Jr. 1987.
Ra.-- - = M Tamnar=*i=e T imits for Drv Stormon of Snent T.ieh? Wa*=e Rameerw Zirr alav.CinA Fuel Rods in Inert Gas. PNL-6189, Pacific Neems laboratory, Richl==d Washington.
i 4
i Page 12of 27 l
= _..
~.. - _ _ _ _. _
1 l
Mohr and A-m=
Rev. 0: May 19,199'T l
l Mahmand S.T.,K.W hay,D.M. Farkas,andR.B. Adamen.1997. "EfIe'ctsofSPP l
Dissolution on Machanical Properties of Zircaloy-2," in Pmcendings of the 1997 Internanonal Topical Meeting on LWR Fuel Performance. Portland, Oregon, M I.aGrange Park, Illinois.
McKinnon, M.A., T.E. Michaner, M.F. Jensen, and G.R. Rodman 1988. PerformanceTestmg and Anelvses of the TN-24P Snane Fuel Drv Sinreaa c'
- rn= dad with Canaalid**ad Fuel.
PNL.4631, Pacific Northwest f.aboratory, Richland, Wa<hina'an; see also EPRINP-6191, Electric Power Research Ln_ inaa, Palo Alto, Califorma.
e Mitsugt, T., N. Kushida, and K. Kilnehi 1997. " Behavior ofMOXFuel Irradiated in a. Thermal Reactor," pp. 54 61, N-mE= of the 1997 LL. rian=1 Tonical Maatina on LWR Fuel Performance. Portland, Oregon, E LaGrange Park, Illinois.
l Newman, LW. 1986. The Hot Cell Framination ofOconee 1 Fuel Rods after Five Cveles of Inadianen. DOE /ET/34212-50, U.S. Department of Energy, Wa<hing+aa D.C.
Pamter, C.L, J.M. Alvis, A.L Marion, G.A. Payne, and E.D. Kendrick. 1994. Fuel Performance Anm=1 Renart for 1991. NUREG/CR-3950, PNL-5210, Vol 9, Nuclear Regulatory Comminaion, Washington, D.C.
Pechs, M. and J. Fleisch. 1986. " LWR Spent Fuel Storage Behaviour,"1 Nucl Ma*w Vol 137, l
pp.190-202, Elsevier Science Publishers, Amd m, North-Holland.
l Porsch, G., J. Fleisch, and B. Heits. 1986. " Accelerated High-Temperature Tests with Spent PWR and BWR Fuel Rods under Dry Storage Conditions," Nuclear Technology, Vol. 74, pp. 287-298, American Nuclear Society, I.aGrange Park, Blinois.
Sabol, G.P. RJ. Comstock, G. 9haaal-ger, H. Kuni=hi. and D. L Nuhfer. 1997. "In-Reactor Fuel Claddmg Corrosion Perfbrmance at Hidy Burnups and Higher Coolant Temperatures."
l Pmceedinas of the 1997 International Topice Meenne on LWR Fuel Perrnrmance Portland Oregon,M LaGrange Park,Blinois.
Page 13 of 27
Moir and h=ar='==
Rev. G: May 19,1997 SchafTier-I4 Pichon, T., Ph. Geyer, P. Delobelle, and P. Bouffiour. 1997. W1ine of the Mechanical Behaviour ofZircaloy4 Cladding Tubes from Unirradistad State to High Burn-up,"
Prneaadin== of the 1997 Intarn=+ianal Tonir=1 Maatine on LWR Fuel b Tm - Portland, Oregon, AHS. LaGrange Park, Illinois.
Smith, G.P. 1994. The Evaluation andDemonstration ofMethods forImproved Nuclear Fuel Utilization. DOE /ET/34013-15, U.S. Department of Enenty, Wamhina'an, D.C.
Van Swam, L.F., A.A. Strasser, J.D. Coolc, and J.M. Burger. 1997. " Behavior of Zircaloy4 and Zirconium Liner Zircaloy-4 Cladding at High Burnup," pp. 421-431. Pmcendings of the 1997 International Tonical Meetme on LWR Fuel Performance. Portland Oregon, ANS, LaGrange Park, Illinois.
VECTRA.1996. Calentation P=clemae Revision 1. NUH004.0409, VECTRA Technologies, San Jose, California.
Wisner, S.B. and R.B. Ad===an 1996. " Combined Effects ofRadiation Damage and Hydrides on the Ductility ofZircaloy-2," Falaroad Waldan Pmer4== Gmun Mene Norway.
Woods, K.N. and W. Klinger.1997. "Siemens Fuel Performance Overview,"Pmceedings of the 1997 L ;-... danal Tonical Maa+ine on LWR Fuel Perfc==#
Portland, Oregon, ANS, LaGrange Park, Illinois.
Yang, R.L. 1997. " Meeting the Challenge of Managmg Nuclear Fuel in a Coe dve Envii.:------- : " pp. 3-5, PrMae of the 1997 T -.udanal Tonical Meetine on LWR Fuel Performance Portland, Oregon, AHS, LaGrange Park, Illinois.
Pap 14 of 27 y
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.__._._._._____._.__._.___..._.__._.._....._..._..._...___._.______.___._.7 Mohrand A==aase==
ltsy. 0: May 19,1997 FIGURES B
J 1 --E*,(%)
e
~
0,8- -
a 2
0,6- -
o rr a
0,4 -
a38)Mpa a418 MPa 0,Z - -
Uk 0 ::
O 40S 800 1200 1600 2000 Figure 1.
Biaxial Creep Tests at 380*C on Cladding Tubes Irradiatad to 28 2
4.5 x 10 n/cm (E>l MeV)(SchmMlar-Le Pichon 1997)
O.
Page 15 of 27
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i a
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25 1
Figure 2.
Biaxial Tensile Tests at 380*C on Ct=Ading Tubes Irradiated to 28 2
8.5 x 10 n/cm. (E>l MeV) (SchafBer-T.4 Pichon 1997) i.
Page 16 of 17 4
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Average Peak Oxi&rnickness vs. MBusnap for m M
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ESctofBumup onOxideTWh firDupiacPWR.NW (hds muirity1997)
Paoe n oi n
PNL-5210 Vol. 9 J
Fuel Perforulance Annual
! Report for 1991
~
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w "DMe B.8 %pkal 4%el A==uy rer rs Ypedar ABSCENF ABB CENF B&W B&W B&W RAW B4W '
PMar%pe PWR PWR PWR PWR PWR PWR PWR iteactorSystem CE CE B&W B&W E
R.
g.J 1%elRod Array 14:14 16 16 15:15 17x17 15x15 (5xt5 17:17.
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Mark O Haddam Mark BW adarkBW Assemblies PerCore
'/17 217 177 205 157
}57 193 I%et Rod Poskipes per Assently 176 236 225 289 225 p5 289 '
%picalNamtero(Ibe6ed Rods per 164 224 208 261 204 70(
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^****1 Rod Pisch, 14.7 12.9 14 4 -
12.3 14 3 14 3 12.6 aun(Is.)
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(0.506)
(0.568)
(0.502)
(0.563)
(0.563)
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Syseen Precogro, 15.5 15.5 15.2 15.5 13.9 13.9 15.5 5
MPa (psia)
(2250)
(2250)
(2200)
(2250)
(2015)
(pl5)
(2250)
Core Average Powst Density, 78.5 96 4 91.4 107.3 82.25 l12.25 82.25 kWAiser Average (JtGR,(*)
20.0 18.2 20.3 18.8 18 3 l84,
17.8 kWM(kWS)
(6.09)
(5.54)
(6.20)
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(5.60)
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77.6,
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(6.88) p.66)
(7,76)
(8 42) kWAR(kWA)
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(16.3)
(13.0)
(16.16)
(15.20)
(14.5)
(14.5)
(13.0)
Mas. I%el'ihop.,
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- C (*F)
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(4155)
(3900)
(3900)
(3500) h Core Average 3.89 2.36 3.30 3.15 4.00 3.41 3.40 fi Barichmentwt%mU O
Max. I.acal Exposure, 50 55 55 55 55 55 55 OW4/MT l
Cladding MascrialM Zry.4 Zry-4 Zry-4 Zry-4 304SS Zry-(
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3.71 4.09 '
3.904 3.878 3.218 3.197 3.848 in (la.)
(145.9)
(161.0)
(153.7)
(152.7)
(126.7)
(125.9)
(151 $)
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3.81 3.602 3.632 3.061 3.012 3.658 m(In.)
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(150)
(141.8)
(143.0)
(120.5)
(118.6)
(144.0)
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(8.6)
(10.0)
(11.7)
(9.5)
(4.8)
(63)
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anm (in.)
(0.44)
(0382)
(0.430)
(0379)
(0.422)
(0.422)
(0374)
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(0384)
(0332)
(0.377)
(0331)
(0389)
(0368)
(0.326)
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(0.028)
(0.025)
, (0.0265)
(0.024)
(0.0165)
(0.027)
(0.024)
Diametral Osp,M 190.5 178 213.4 198.1 165 178 165 micron (mil) p.5)
(7.0)
(8.4) p.8)
(6.5) p.0)
(6.5)
Ibel Pellet Diameter, 9.56 8.26 9362 8.209 9.715 9.17 8.115 mm (in.)
(03765)
(0325)
(03686)
(03232)
(03825)
(0361)
(0.3195) 1%el Pellet length, 11.43 9.91 11.05 9.53 1f.63 10.80 10.16 tam (la.)
(0.45)
(039)
(0.435)
(0375)
(0.458)
(0.425)
(0.400) ibel Pellet Density, 95 95 95 95 95 95 96
%TDM
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user -
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MilGas aqd Ppssum fic lic lic ye Ile (psig) 300-450 300-450 415 435 40 f
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(b) Dpe 301 pqipless ascel (304SS). IJrcalayi (Zsy4), and Zircaloy-2 (Zry-2).
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2 (c) P'
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(V.nSwant etat 1997, Garde 1986, Smith 1994 adDideon 1983)
.O Page18 cf 27 i
_..____.___.-.___._.________..__..___-.x mw. -
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0 10 20 30 40 50 60 70' Bumup, GWd/MTU Figure S.
Total Bangstion for BurstTests Conducted in the Temperatma Range of 599 to 652*P onZinzioy-4SNF NdiSpecimens withB===Eavels up to 60,000 MWdMIU (Garde 1986, Smith 1994, and Dideon: 1983) i Page19 af 27
8 n
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- U w. = w.,t,. m
)
d i100 1000 a m
- ~
900
.g 800 B
5@
700 e
600c3 ah 500'o g
400()
asmimim; 315c C?
O Gade1986; 315C n
300 7Dideon 1983; 321-345C g
200 100 0
0 10 20 30 40 50 60 70 Bum Up, GWd/MTU Figum 6.
Burst Stress for BustTesu Continenwi in daTemperamra Range of 599 to 652'P on LA7S SNF NW Speciesen with Bumug Insels up tr. 50,000 mwd /MTIT(Garde 1986, Shuidt 1994, amiDideca 1983) 4
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FiguurI Axial Yield and Ultimmen Tesile Stangth fir Zirtalay 4 NW Irradiasedto 55,000 MWdNIU st343*C(Newsma 1986)
O Page21 at 27 1
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Figme8.
Unifhrm and.ToadElongsnanfbrZircalay-4 N=Mi Irradissed.m 55,000 mwd /MTU ac343 *C (Newman 1986) e vie.22 < zr
i
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Figue 9.
WeninghouseCamsionDamforZiMr/ ?WatHighBanups p
(Kilp atal 1991) rs PugsUof 27
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Figure 10.
Marinnnn Measured. Oxide Thicimess vs. M heviiva khe N==(IGlp etat 1991)
~
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{O LJ h
k}
lhble B.1 (contd) o h
%mdor GE GE GE GE SNP SNP SNP SNP SNP SNP g
5 Reactor 1)pe BWR/2,3 BWR/2,3 BWR/4-6 BWR/4-6 BWR/2,3 BWR/2,3 BWR/2,3 BWR/2,3 BWR/4-6 BWR/4-6 Reactor Systest OB OE GE OB OE OB OE GE OB OE IbelRod Array 8x8 9x9 8x8 9x9 9x9 9x9 9x9 9x9 9x9 959 Wasion Barrier OB-11 GE-4a OE-11 JP-3 9-5 IX 9X JP-4,5 9-5 Assembiles Per Core 560 560 724 IbelRodPositionsperAssembly 64 81 64 81 81 81 81 81 81 81 1)pkalNumberofIbeled Rods 62 77 63 77 79 76 72 72' 79 76 Per As.sembly Rod Pitch, 16.3 16.3 14.52 14.52 14.52 14.52 min (In.)
(0.640)
(0.640)
(0.572)
(0.572)
(0.572)
(0.572)
Systent Pressure, 7.14 7.14 7.07 N
MPa (psia)
(1035)
(1035)
(1026)
, Core Average Power Density, 49.15-5031 46 kW/ liter Average LHOR,(*)
17.1 17.9 12.1 kWAn(kWM)
(5.38)
(5.45)
(3.68)
Axial Peak LHOR 21.24 21.48 17.5 la an Average Rod, (6.99)
(7.09)
(5.34) kW/m(kWM)
Max. Peak LHOR, 44.0 44.0 37.7 kWho(kWM)
(13.4)
(13.4)
(11.5)
Max.1%ct1bmp, 1890 1830 2040
'C (*F)
(3435)
(3325)
(3705)
Core Average Enrichment, 1.99 2.54 2.8 wt% n5U Max. Local Exposure, 45 40 55 OWd/M1U Cladding Material )
Zry-2 Zry-2 Zry-2 Zry-2 Zry-2 Zry Zry-2 Zry-2 Zry-2 Ziy-2 ld b #**
x gy p.
(%)
i j
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'ihbic B.1 (contd) 4 Ysador,
'"SNP ' 'Stit PuelRodlength, 4.20 4.09 4.04 4.16 m (la.)
(165.4)
(161.1)
(159.07)
(((aM)
ActiveIW.Ilength, 3.81 3.71 3.68 3.81 m (la.)
(145.24)
(146)
.(145.24)
(150.0)
Plemma 14agth, 0.25 036 0.243 0,74 m (13.)
(9.48)
(14.0)
(9.580)
(9.578)
IbelRod OD, 12.27 12.52 10.76 la76 sun (14.)
(0.483)
(0.493)
(0.424)
(0.424)
Cladding ID, 10.64 10.80 9.25 mm (in.)
(0.419)
(0.425)
(0.364)
C3 adding'Ihickness, 0.813 0.863 0.76 0,76
[
mm (14.)
(a032)
(0.034)
(0.030,)
, (0.030)
Dirmetral Oap,(*)
229 229 190 microg(mil)
(9.0)
(9.0)
(7.5)
Ihet Pellet Diameter, 10.41 10.57 9.05 mm (19.)
(0.410)
(0.416)
(0356)
Ibel Pcilcllength, 10.41 10.67 10.41 am(la.)
(0.410)
(0.420)
(0.410) 95 95 94.5 94.5 Pp Density, F1110as sad Pressure lle He ne He (psig) 30 0
60
. 60 (e) Gadolinielefeelrods.
(b) Sossom 42 iadne of fuel sad is asainicos esect.
(c) utan - neemt hees seeeration res (4) lype301aam scelpo4ss).zircanoy-4(zsy-4),andzircaioy-2(zry.2).
p e
I Q
(c) Diensarets*P"claddi"4ID - Pe#88 d""***'-
m (0 necroucal densisy (TD) of esolc#uncaric UD is 10.96 s/c=*.
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'thble 3.1 (contd) m
%eder SNP SNP SNP SNP SNP SNP H
H H
Reactor'I)pe BWR/4-6 BWR/4-6 PWR PWR PWR PWR PWR PWR PWR Reactor System OE OB ABB CE E E
IbelRod Array 9x9 9x9 14x14 14x14 15x15 17x17 14x14 15x15 17x17 Wasion IX 9X ABB CB 'Ibp Rod E E
WE WE Watage5 Asseaddies Per Core 193 193 193 193 IbelRod Positions per Assembly 81 81 176 1%
225 289 176 225 289 W HumberofIbeled Rods per 72 72 176 179 204 264 176 208 264 f
Assembly Rod Pitch, 14.45 14.45 14.73 14.12 14.30 12.60 14.73 143 12.60 man (in.)
(0.569)
(0.569)
(0.580)
(0.556)
(0.563)
(0.4%)
(0.580) (0.563) (0.496) g System Pressure, 15.5 15.5 15.5
[
MPa (psla)
(2250)
(2250)
(2250)
Core Average Power Density, 98.1 104.7 98.1 kWAlter Average LHOR,(*)
20.76 22.0 17.8 22.0 kW/m(kW/n)
(634)
(6.60)
(5.44)
(6.70)
AxialPeak LHOR 26.40 21.4 26.40 la en Average Rod, (8.04)
(6.53)
(8.04) kW/m(kW/n)
Max. Peak LHOR, 51.9 54.5 61.7 j
- kW/mi(kW#t)
(15.8)
(16.6)
(18.8) l Max. Ibel'Itmp, 2200 1747 2340 l
'C (*F)
(3997)
(3177)
(4250)
Core AverageEnrichment, 3.34 2.55 3.65 3.00 2.80 3.42 MU wt%
Max. Local Exposure, 47.5 52 50 GWd/Mul l
l l
)...
%,/
\\
Table B.4 (contd)
Weder SNP SNP SNP SNP SNP SNP W
H H
Cladding Material (*)
Zry-2 Zry-2 Zay-4 Zay-4 Zsy-4 Zay-4 Zry-4 Zry-4 Zry-4 thel Rod 1.cagth, 3.72 3.86 3.86 3.86 3.72 3.80 3.87 m (la.)
(146.4)
(152.0)
(152.0)
(152.0)
(146 4) (149.7) (1523)
Active 1%cllength, 3.81 3.81 3.40 3.65 3.65 3.65 3.47 3.66
'3.65 m (in.)
(150.0)
(150.0)
(134.0)
(144.0) ' (144.0)
(144.0)
(136.7) (144)
(144.0)
Plenum Length, 0.18 0.l?
0.18 0.21 0.18 m (in.)
(7.28)
(6.80) p.26)
(8.2)
(74!)
I%cl Rod OD, i1.17 10.59 10.76 9.14 11.17 10.72 9.14 mm (la.)
(0.431)
(0.431)
(0.440)
(0.417)
(0.424)
(0360)
(0.440) (0L422) (0360)
Claddlag ID, 9.25 7.87 9 48 mm (la.)
(0364)
(0310)
(0373)
Claddlag'lilckness, 0.78 0.74 0.76 0.63 0.66 0.62 0.058 mm (la.)
(0.025)
(0.029)
(0.031)
(0.029)
(0.030)
(0.025)
(0.026) (0.024) (0.023)
Diametral Oap,(*)
190' 177.8 190 micron (mil) p.5) p.0) p.5) 1%cl Pelict Diameter, 9.49 9.47 939 9.05 7.69 9.68 9.29 7.85 mm (la.)
(0374)
(0373)
(0370)
(0356)
(0303)
(0381) (0366) (0309) 1%cl Pellet length, 10.80 6.93,
8.84 15.24 15.24 12.95 mal (la.)
(0.425)
(0.273)
(0348)
(0.600) (0.600) (0.510)
INel Pellet Density, 96.26 94.5 94 94 94 94 95 95 95 WID(I)
F1llGas and Pressure ife lie lie tic lie lie tic 375 305 290 290 275-400
, (psig)
(a) n.any.4 3. guel gode (b) Basom42inchesof facerodisna1=le==atcet c
(c) IJIOR = Ilment heat generation rate.
g h
(d) "lype 301 asat-le== e4cel (304SS), Zircaloy.4 (Zry-4), and Zircalcy-2 (Zry-2).
9 5,0 (c) Dinaneiral gap = cladding ID - pellet diameter.
y; 3
(f) 'Iheoreticaldeasily("ID)of atnachhmetricUO 1s1096g/cm.
2 Of o
PNL-4189 RIv. L UC-85 l
BEHAVIOR OF SPENT NUCLEAR FUEL
'. ^k AND STORAGE SYSTEM COMPONENTS IN ORY INTERIM STORAGE
>o 4
A. B. Johnson, Jr.
E. R. Gilbert R. J. Guenther 1 'O l
i i
February 1983 I
Prepared for the U.S. Department of Energy I
Under Contract DE-AC06-76RLO 1830 Pacific Northwest Laboratory Richland, Washington 99352 O
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