ML19262C665
| ML19262C665 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/08/1980 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Gammill W Office of Nuclear Reactor Regulation |
| References | |
| P-80020, NUDOCS 8002150247 | |
| Download: ML19262C665 (8) | |
Text
.
public service company oe cdleraale 16805 Weld County Road 19 1/2, Platteville, Colorado 80651 February 8,1980 Fort S t.
Vrain Unit No. 1 P-80020 Mr. William P. Gammill Assistant Director for Standardization and Advanced Reactors Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C.
20555 At ten tion :
Ifr. Mike Tokar Docket No. 50-267
Subject:
Supporting Documentation for Change to DF 6.1 Re ference :
P-80003
Dear fr. Cammill:
As requested in discussions between your Mr. Mike Tokar and Mr. Fred Swart of PSC, enclosed are the supporting documents for our requested change to De-sign Feature DF 6.1.
If you have any further questions, please contact Mr. J. W.
Gahm, (303) 571-7436.
Ve ry truly yours,
<*h~ )Y
- w.i m '
Don Ware $ bourg f
Manager, Nuclear Production DW/cls Enclosure Doo3 s
///
241 p
-se e
i
Attachment to GLP-5868 Perfor=ance of FSV Fissile Fuel with a Th:U Ratio of 3.6 to 1 S m ry on Septenber 14, 1971, prior to final AEC approval o'f the proposed Fort St. Vrain Technical Specifications, the FSV fuel specification was modified to allow a decrease in the nomi.al Th:U ratio of fissile fuel kernels from 4.25/1 (10,5) to 3.6/1 (+1.2, -0.2).
However, the Th:U :atio described in Technical Specifi-cation D.F.6.1 re=ained 4.25/1.
The change was made after the =aximum lifeti=e of initial core seg=ents was reduced from six years and to enable initial core metal loading require =ents to be =et.
The allowable range of Th:U ratio in the newer specification overlaps that of the original specification.
The 3.6/1 fuel was used in the latter stages of initial core production, in seg=ents 7 and 8, and will be used for future seg=ents containing (Th/U)C7 fissile fuel. The =ajor i= pact of the decrease in Th:U ratio is a slight increaoe in the fissile kernel peak burnup for six year old fuel in FSV.
The peak burnup for 4.25/1 fuel is expected to be 20% FIMA; the peak burnup for 3.6/1 fuel is expected to be 22.4%
FIMA.
(None of the fissile fuel in initial core seg=ents is expected to exceed 20% FIMA.) Although approxi=ately 20% of the 3.6/1 fissile fuel in any one seg-ment of an equilibrium FSV core will experience burnups exceeding 20% FIMA at end of life, only 3-4% of the fissile fuel on a core average basis will experience burnups greater than 20% FIMA at end of cycle. This s=all shift in burnup dis-tribution will not have a noticeable i= pact on FSV fuel perfor=ance.
Fuel Descriction The nominal dimensions for FSV (Th/U)C2 fissile fuel particles are given in Table 1.
The kernel and coating dimensions of fuel having Th:U ratios of 4.25/1 or 3.6/1 are the same.
Changes in fuel perfor=ance only reflect the s=all change in burnup associated with the decrease in Th:U ratio. Perfor=ance during nor=al reactor operation and under hypothetical accident conditions are discussed below.
Fuel Perfor=ance - Norm l Reactor Ooeration Three phenomena must be considered when discussing TRISO (Th/U)C2 perfor ance during nor=al reactor operation.
The first is kernel migration.
Coating failure due to kernel nigration is assu=ed if a nigrating kernel contacts the structucal layers of a TRISO coating.
Since the coating di=ensions are the same for 4.25/1 and 3.6/1 kernels, this pheno =enon will be affected only if the change in Th:U ratio changes the rate of kernel migration.
The rate of =igration, which is a function of te=perature and te=perature gradient, is described by the kernel migration coefficient (KMC).
KMC values have been determined for Th:U ratios in the range 4.25/1 to 1.60/1 (Ref. 1, 2).
No relationship between Th:U ratio and KMC over this range of Th:U ratios has been noted.
A specific co=parison of KMC values determined for kernels having Th:U ratios of 4.25/1 with those derersined for kernels having Th:U ratios in the range of 3.48/1 to 4.08/1 showed no dependence of KMC on Th:U ratio (Ref. 2).
The change in Th:U ratio from 4.25/1 to 3 5/1 will not, therefore, have any i= pact on KMC values or on coating failure associated with kernel migration.
The second phenomenon is failure of particles =anufactured with missing or de-fective coatings.
Since the s=all decrecse in Th:U ratio does not result in a change in specified coating properties or coating proc-sses, the nu=ber of par-ticles =anufactured with nissing or defective coatings would not change and coating failure and fission product release due to this pheno =enon w0uld not change.
Attachment to GLP-5868
_2_
The third phenocenon is pressure vessel failure, which is a function of kernel burnup, coating' design, and operating te=perature.
Coating failure fractions observed during irradiation of TRISO (Th/U)C2 fissile fuels having Th:U ratios less than 4.25/1 and/or kernel burnups exceeding 20% FIMA are su==arized in Table 2.
The kernel and coating dimensions of the test fuel are consistent with those shown in Table 1 for the reference fuel.
Irradiation te=peratures are consistent with those expected in FSV.
Observed failure fractions are low.
Assuming for each particle batch the larger of the two values for failure fractions listed in Table 2, the average failure probability for (Th/U)C2 fissile fuel irradiated to 20% FIMA is 0.004.
This is consistent with results obtained in the FSV fuel proof test (Ref. 3).
The average failure probability for fuel irradiated to burnups greater than 20% FIMA but less than or equal to 27% FIMA is also 0.004.
These results show that the reference fuel design was conservative and that the s=all increase in burnup associated with the decrease in the nominal Th:U ratio from 4.25/1 to 3.6/1 will have a negli-gible effect on pressure vessel failure.
Based on the three considerations discussed above, it is concluded that de-creasing the Th:U ratio in FSV fissile fuel from 4.25/1 to 3.6/1 will not result in additional fuel failure and fission product release during nor=al reactor operation.
Fuel Performance - Design Basis Accident Conditions The failed fuel fraction at the onset of a design basis accident (DEA) and the increase in failure with increasing te=perature during a DBA must be considered in the accident analysis.
Failure at the onset of an event is the result of failure during nor=al reactor operation.
It has already been shown that ex-pected fuel failure fractions will not be affected by the decrease in Th:U ratio.
The core average failure fraction assumed in the FSV FSAR at the onset of a DBA is 0.05.
This value was established non-cechanistically and is well above the value expected for normal operation.
The assumption of 5% failure at the onset of a DBA is not affected by the Th:U ratio of FSV fissile fuel.
Core heatup siculation tests of TRISO-coated fuel particles have been conducted between 1100 and 25000C to evaluate fuel perfor=ance under loss of forced circu-lation (LOFC) conditions.
LOFC (Design Basis Accident #1) is the cost severe accident, with regard to fuel particle perfor=ance, analyzed in the FSAR.
Test results show that the do=inant failure cechanis=s include sic-actinide =etal re-actions, sic-fission product reactions, and sic deco = position.
Fuel perfor=ance appears to be independent of kernel burnup.
A specific example of the release fraction observed for Kr-85 while heating 08 TRISO (4.25 Th/U)C2 fuel particles from 1100 to 24000C is given in Fig. 1 (Ref. 4).
The particles had been irra-diated to 18.2% FIMA prior to heating.
Release of Kr-85 was not observed until te=peratures exceeded 21000C.
(At these temperatures, Kr-85 release fractions are equivalent to failed fuel fractions.) When analyzing FSV fuel performance during LOFC in the FSAR, it was assumed that the fuel failure fraction was 0.05 for te=peratures less than 17250C and 1.0 for temperatures greater than or equal to 1725 C.
This assu=ption is extrecely conservative relative to data shown in Fig. 1.
Since testing of this nature has shown a general lack of dependence on kernel burnup, it is concluded that margins suggested by Fig. 1 are not de-creased by changing the nominal Th:U ratio of FSV fissile fuel from 4.25/1 (peak kernel.burnup of 20% FIMA) to 3.6/1 (peak kernel burnup of 22.4% FIMA).
Attachment
. to GLP-5868 References 1.
O. M. Stansfield et al., " Kernel Migration in Coated Carbide Fuel Particles,"
Nucl. Tech., 25, March 1975, p. 517.
2.
J. R. Sims et al., " Migration of (Th/U)C2 Fuel Kernels," GA-A13825, May 1976.
3.
C. B. Scott and D. P. Harmon, "Postirradiation Examination of Capsule F-30,"
GA-A13208, April 1975.
4.
"HTCR Fuels and Core Development Program Quarterly Progress Report for the Period Ending November 30, 1977," GA-A14744, December 1977, p. 9-9.
e e
Attachment
. to GLP-5868 f
Table 1 Dimensions of TRISO-Coated (Th/U)C2 Fuel Particles for FSV Mean Mean Kernel Buffer Mean IPyC Mean sic Mean OPyC Particle Diameter Thickness Thickness Thickness Thickness Tvoe (um)
(um)
(um)
(um)
(um)
(Th/U)C2 100-175 50 20 20 30 (fissile A)
(Th/U)C2 175-275 50 20 20 40 (fissile 3) h e
9 m
Table 2.
Summary of Irradiation Exparience with TRISO (Th/U)C Fissile Fuel llaving Th:U Ratios 44.25/1(a,b)~.
2
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^I KERNEL COATING TillCKNESS (pm)
DATA Average Kernel RETRIEVAL Diameter n/U Temperature Itur nu p Nero N tallographic 25 2
14 0.
Capsule (pm) htlo Buffer IPyc SIC OPyC (10 n/m )
(OC)
(I F1HA)
Exam Eman 3325-117E P131t 150-250 1.6 40 49 6
8 4.2 1250 22 0.1 0.2 3325-125E P13H 150-250 1.6 30 38 9
11 4.3 1250 22 0
0 3325-117E 'I P1311 150-250 1.6 40 49 6
8 3.8 925 22 0.1 0
I 4 2 63-7.5 E
- P13L 200 2.25 48 22 19 50 7.1 1575 25 0
0 4263-75E P13L 200 2.25 48 22 19 50 5.8 925 22 0
0 4000-232 P1Di 198 ffR 51 19 21 38 6.7 1340 25 IID 0
3923-87E P20 150-250 2.25 49 23 20 30 6.5 1075 21 0
0 3923-101E P20 150-250 2.25 49 24 15 29 6.5 1065 21 0
0 I
3923-149E '
P20 150-250 2.25 49 25 21 32 8.7 1125 27 0
0 e
3923-141E P20 150-250 2.25 49 24 15 31 8.7 1150 27
<0.1 0.2 3923-149 P20 150-250 2.25 49 25 21 32 7.0 865 26 1.7 0
3923-87P.
- P20 150-250 2.25 49 23 20 30 7.0 850 26 1.2 0.4 3923-113E P21 150-250 2.75 49 13 18 29 3.0 1240 12 0
0 4000-889E P21 150-250 2.75 44 11 16 48 4.3 1240 16 90 90 3923-103E P21 150-250 2.75 49 18 18 30 3.0 1240 11 0
0 1
3923-113E P21 150-250 2.75 49 13 18 29 3.6 950 17 0
0 Y
4000-88?2 P21 150-250 2.75 44 11 16 48 3.6 890 17 0.2 12 4155-149E P22 150-250 2.75 55 20 20 38 7.7 820 20 0.1 0
4155-151E r22 150-250 2.75 49 20 19 29 7.7 1095 20 0
0 4155-139E P22 150-250 2.75 49 20 18 36 7.7 1045 20 0.2 0
4155 '35E P22 150-250 2.75 57 19 15 35 7.3 996 20 1.7 1.1 a )- 131 E P22 150-250 2.79 46 18 24 33 7.3 715 20 0.5 09 4155-149E P22 150-250 2.75 55 20 20 38 5.2 695 20 0
0 4155-151E P22 150-250 2.75 49 20 19 29 5.2 950 20 h
0.4 4155-139E P22 150-250-2.75 49 20 18 36 5.2 950 20 0
0.5 4155-149E P22 150-250 2.75 55 20 20 38 6.0 920 20 0.1 0
4155-151E P22 150-250 2.75 49 20 19 29 6.0 920 20 0
0 4155-139E P22 150-250 2.75 49 20 18 36 6.0 920 20 0
0.3 4263-77E P23 150-250 2.24 48 22 19 31 2.3 1240 12 0
0 4263-61E P23 150-250 2.24 45 16 21 32 2.3 1240 12 0.1 0
4263-59n P23 150-250 2.24 45 16 21 30 2.3 1240 12 0
0 re p.
4263-75E P23 150-250 2.24 48 22 19 30 2.1 1230 11 0
0 9"
4263-77E P23 150-250 2.24 48 22 19 31 2.0 905 12 0
0 c) m 4263-61E P23 150-250 2.24 45 16 11 32 2.0 910 12 0
0 y Q.
4263-59E P23 150-250 2.24 45 16 21 30 2.0 920 12 0
1 1 0 4263-75E P23 150-250 2.24 48 22 19 30 1.8 950 9
0 0
m rt 4000-183 Y-25 175-275 2.75 50 20 20 43 4.3 1150 16 0
0 4000-101 P-25 175-275 2.75 50 20 2'l 48
.4.3 1190 16 0
0 4000-179 F-25 175-275 2.75 52 16 20 42 4.3 1150 16 0
0 4000-180 F-25 175-275 2.75 45 23 21 37 4.3 1190 16 0
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1.0 TRISO (Th/U)C Th:Uratio-d.25/1 Kernel Burnup = 18.2% FIMA 0.8 lleating Rate - 1500C/hr E:o 0
0.6 k.
E 0.4 0
4i i
0 0.2
\\
O '
O I
O I O^
0 1100 1300 1500 1700 1900 2100 2300 2500 Temperature (OC)
Fig. 1:
Kr-85 Release fraction observed as a function of temperature n--
while heating TRISO (Th/U)C fr m 1000 to 2400 C.
2
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? ii 30 8"
.