ML19289E375
| ML19289E375 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/09/1979 |
| From: | Wessman G GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | Gammill W Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904170259 | |
| Download: ML19289E375 (28) | |
Text
'N k ,
GENE At. MIC COMPANY DIEGO CAUFCTINIA 92138 April 9, 1979 Mr. William Gammill Assistant Director for Advanced Reactors Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Gammill:
Enclosed are fifty (50) copies of General Atomic Company's response to Lead Item VIII contained in the February 16, 1979 letter, " Review of Thorium-Oxide and Uranium Carbide Fuels". Lead Item VIII pertains to the review of uranium carbide. Responses to the remaining Lead Items, which pertain to the review of thorium oxide, are being transmitted under separate cover to enable GAC and NRC to maintain separate documen-tation of these reviews.
If you have any questions regarding the enclosed infomation, please do not hesitate to contact us, as we are anxious to expedite the review of uranium carbide toward completion by the end of June.
Sincerely, 4-(<:5 ni-G. L. Wessman, Director Plant Licensing Division GLW:mk Encl.
790417DJR57
e' s STATE OF CALIFORNIA ss.
COUNTY OF SAN DIEGO )
After being dulyCompany, sworn, the person Wessman of General Atomic signed thekrown wit'11ntodocument me to be this G. L.
9 A day of April 1979.
WITNESS my hand and official seal.
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L Notary Public
o o RESPONSE TO NRC LEAD ITEM VIII ON UC., REVIEW VIII. Earlier review efforts (circa 1976) on HTGR fuel particle failure models and data culminated in an evaluation report NUREG-Oll1. In that report the TRISO UC particle irradiation test data base, analyses, and coating fa,ilure models were critiqued, weaknesses were identified, and model modifications were suggested (see, in particular, section III of NUREG-Olll). It would be helpful, there-fore, to have a summary of the advances in the state-of-the-art regarding TRISO UC7 particle testing and fuel failure model development that hive been made since issuance of NUREG-0111. For example, what new evidence has been generated in support of (1) the sic tensile stress and pressure vessel failure correlation and (2) the sic fission-product-interaction failure rate at temperatures above 1500 C? What are the predicted failure rates for each identified potential failure mechanism as a function of particle age and tempera-ture, and what specific data support the predictions?
RESPONSE
The irradiation data base on reference TRISO coated UC 2 fuel particles has been extended since the issuance of NUREG-Olli and is summarized in Figure 1.
This figure is a cumulative burnup distribution which defines the fraction of UC 2 irradiated to different burnups and includes capsules P13R and P13S as well as number of additional capsules irradiated since issuance of NUREG-0111. 5 Figure 1 applies to a total of ~3.16 x 10 TRISO coated UC2 particles 25 irradiated between 885 -1240 C and 3.5-12.2 x 10 n/m The 5 2(E>29fd)HTGR.
figure shows that ~2.5 x 10 UC2 particles have been irradiated to burnups 3
271% FIMA and ~9 x 10 to burnups 277% FIMA. These burnup limits represent the mean and peak values for 6-year-old fuel in FSV.
The ranges of TRISO coated UC 2 kernel and coating properties tested in the irradiation capsule test program are listed in Tables 1 and 2. These tables consist of three rows: the first row presents the nominal properties for the reference design; the second row defines the range of irradiation test experience for each of the properties; and the last row presents the nominal property values for the core heatup simulation test (CHST) samples used in the safety analysis test program. (The CHST program for TRISO UC is 2
described below.) Two important points are evident from Tables 1 and 2:
- 1. The range of irradiation test experience brackets the nominal design values for all critical properties.
- 2. CHST sample properties are representative of the nominal reference design properties.
Consequently, the irradiation capsule test and CHST results can be used as a representative basis for judging TRISO coated UC2 fuel performance _under normal and accident related conditions.
Advances in TRISO coated UC2 particle design since NUREG-Olli can best be des-cribed by a joint discussion' of the TRISO coated particle stress models and irradiation test data. The TRISO coated UC2 particle design is developed on the basis of calculated sic stress distributions which assure that the expected particle failure from internal fission gas pressure is less than or equal to that calculated for FSV TRISO coated (Th/U)C 2 fuel. In order to obtain a conservative design, the stress calculations are done for peak fissile fuel exposure conditions, i.e., 1250'C, 8x1025n/m2 (E>29fd)HTGR, 20% FIMA for (Th/U)C '
2 and 75% FIMA for UC2. Figure 2 is a schematic diagram which illustrates the methodology used to calculate the sic stress distribution in a TRISO coated particle population. The explicit functions which define the maximum tensile stress in a TRISO coated particle are based on the Kaae performance model (Ref. 1). These functions are used to calculate the maximum sic stress for individual particles based on a random selection of critical particle pro-perties, i.e., kernel and coating dimensions and densities. This part of the calculation is performed by a Monte-Carlo type routine which accounts for random permutations of critical properties in TRISO coated particles. The output of the calculational routine defines a histogram which describes the frequency of occurence of a stress value, f(o). This distribution is then used to define a failure criterion based on the assumption that TRISO coated fuel particles with the highest ~ stresses fail. The failure criterion is defined as:
f= g f(o)de FC where f (a) = calculated sic stress density distribution, and f = experimentally observed failure fraction.
FC = Failure Criterion
The empirically detemined criterion for UC is 231MPa (33,500 psi) which 2
is consistent with the threshold value of 207MPa (30,000 psi) originally evaluated in NUREG-Oll1.
The expected failure fractions for TRISO coated UC2 and(Th/U)C 2are shown in Table 3. They are 0.002 and 0.005, respectively. These failure projec-tions are based on nominal fuel properties and expected property distributions, peak irradiation exposure conditions, and empirically determined failure criteria for each fuel type. The expected failure fraction for TRISO coated UC 2 is less than one half of the (Th/U)C 2value; consequently, ample performance margin exists in the TRISO coated UC2 particle design relative to the (Th/U)C2 design.
The exceptional performance of reference type TRISO coated UC2 particles has been demonstrated in the large scale SSL-2 irradiation test (Ref. 2).
Table 4 lists the fuel types, nominal particle conditions, and irradiation conditions for fml tested in SSL-2. The test contained 85,200 UC2 particles tested in 27 reference type HTGR fuel rods. In addition, the fuel rods were assembled in a graphite body which simulated the actual design geometry of an HTGR fuel element; consequently, SSL-2 constitutes a large scale proof test of UC2 performance. Figure 3 is a plot of in-pile irradiation test results for SSL-2. The uppe: curve in Figure 3 shows predicted and observed Kr85m R/B (rate of release / rate of birth) versus effective full power days.
The predicted coating failure leading to increased release rates is based on TRISO coated UC2 particle stress calculations using nominal particle dimensions and standard deviations along with SSL-2 exposure conditions.
The important point is that the observed Kr-85m R/B values are substantially less than predicted, which implies that the in-pile failure of UC 2 is less than determined by the TRISO coated particle stress model and that performance is equal to or better than the current (Th/U)C2FSV fuel.
Fuel rod tests conducted in cell 2 of capsule GF-4 provide added confirmation of the expected performance of reference type TRISO coated UC fuel. Table 5 2
lists the fuel types, nominal particle dimensions, and irradiation conditions for fuel tested in cell 2 of capsule GF-4. Figure 4 is a plot of Kr-85M '/B
for GF-4 (cell 2) measured in-pile versus fast fluence. As a basis of comparison, Figure 4 contains both measured (solid line) and predicted (dashed line) Kr-85m R/B results. . The predicted values are based on TRISO coated particle stress calculations for Th02 and UC2 fuel using nominal particle dimensions and standard deviations along with GF-4 exposure conditions. Figure 4 shows that the predicted Kr-85M release values are in agreement with observations.
The following material addresses those portions of Lead Item VIII that deal with TRISO UC 2 behavior at temperatures above 1500 C. New data from fissile fuel safety studies are presented.
The performance of HTGR fuels under hypothetical accident conditions is being evaluated in an on-going core heatup simulation test (CHST) program. Groups of 50 to 200 particles are heated, in the CHSTs, from -1100 to -2500 C over periods of -8, 30, or 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Temperatures are increased linearly with time.
Fuel performance is monitored by measuring fission product release fractions as a function of time and temperature. The key fission product that is moni-tored is Kr85, since gaseous fission product release is related to total coating failure. At present, the release data are compared with predictions made using applicable fuel failure and fission product release models to evaluate the degree of conservatism associated with the models.
The test conditions ushd are chosen to simulate the range of conditions pre-dicted for those hypothetical core heatup events that are considered in reactor licensing and siting applications. In the case of FSV, a key event is Design Basis Accident No. 1 (DBA #1). The FSV FSAR shows, in Appendix D, that the minimum time required for a small fraction of the fuel to reach ~2500 C during DBA #1 would be 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, while the average fuel temperature would not reach 2500 C.
A schematic of the CHST system is shown in Figure 5. Testing is conducted in resistance heated graphite tube (King) furnaces. Each furnace is penetrated by four tantalum tubes. One tube extends about half way through a furnace, is sealed on the hot end, and contains a temperature control thermocouple. Three of the tubes are open ended and extend through the furnace. These tantalum tubes are used to house test samples and mullite traps for cesium collection.
Each group of particles used in a single test is separated into three samples.
Each sample is then loaded into a type H-451 graphite crucibic. One crucible (test sample) is then p15ced in each open ended tantalum tube. Sample test temperatures are monitored optically during heating.
The tests are conducted in flowing helium (50cc/sec/ tube) to assure that released fission products are quickly transported to their respective traps. As the helium and ,ission gases exit the furnace, they are combined into a single flow line that passes through (1) a bed of hot (-500*C) cuprous oxide to convert tritium to water, (ii) a desiccant to remove the water, (iii) a room temperature activated charcoal bed to remove radon, (iv) two ionization chambers and (v) a liquid nitorgen cold trap. The cold traps, which are used to collect Kr85, are changed periodically and gamma counted to obtain Kr85 release data. The response of one ionization chamber is integrated electronically and then related to cold trap Kr85 data to obtain a cortinuous measure of Kr85 release. The second ionization chamber is retained as a backup. Upon completing each test the Kr85 release data are compared with the pretest Kr85 inventory to obtain the release fraction as a function of time and temperature.
Kernel and coating dimensions and densities for the TRISO UC2 samples tested to date are summarized in Table 6. Nominal kernel and coating properties specified for the reference FSV TRISO UC2 design are also shown. Kernel and coating dimensions and densities of the test samples are consistent with the FSV reference design.
Irradiation conditions of the three test samples are compared with expected FSV conditions in Table 7. Three sets of conditions are shown for FSV. The first (core average) represents average conditions expected for an equilibrium core. The second (avg., 6 year fuel), represents the average conditions expected for fuel removed from FSV after 6 years of operation. The third set (peak) gives maximum conditions to be experienced by fuel after 6 years of operation in FSV. Three parameters are included in Table 7. The first two (fast neutron exposure and kernel burnup) are self explanatory. The third (fission density) is the ratio of the average number of fissions per particle to the volume (m3) per particle inside the sic layer. Fission density is proportional to the aver-age fission gas pressure or metallic fission product concentration within indivi-dual test samples. It provides a simple method for normalizing irradiation ex-
posure to kernel / particle dimensions when comparing the behavior of various test samples. As shown in Table 7, the irradiation conditions for the TRISC UC2 test samples range from core average values to peak values expected for 6 year old fuel.
Four CHSTs containing a total of 600 TRISO UC2 particles have been conducted (Refs. 3,4). A sunmary of samples used, number of particles tested, and CHST conditions is given in Table 8.*
Krypton 85 release fractions measured as a function of temperature from TRISO UC2 samples having a fission density of 6.0x1026 are shown on Figure 6. The data suggest (i) that the Kr85 release fraction from TRISO UC2 increases at any temperature T as the time to reach T increases and (ii) that gaseous fission product release fractions from fuel irradiated to FSV core average conditions would not increas;, during DBA #1, until temperatures approached 2100 to 2300*C.
Krypton 85 release data collected from all four TRISO UC2 CHSTs are summarized on Figure 7. The results suggest that Kr85 release fractions from the highest exposure fuel in a 6 year old FSV segment would, during DBA #1, be less than 0.10 at 2000 C.
The results in Figure 7 were combined to determine the expected vnlue and the range at a 90% confidence level for the expected value of Kr85 release fraction vs. temperature during DBA #1. The range at a 90% confidence level is shown as the shaded region on Figure 8. Data collected from an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> test of 88 TRISO (Th/U)C2 particles (Ref. 5) are also shown. The (Th/U)C2 fuel was prepared during FSV initial core production and irradiated to 18.2% FIMA in the FSV fuel proof test (capsule F-30, Ref. 6). The results show that the performance of TRISO UC2 and TRISO (Th/U)C2 would be similar during a core heatup--leading to the conclusion that replacing the present (Th/U)C2 fuel with UC2 fuel would have no effect on FSV reactor safety margins.
- Results from CHST No. 78IHR-C-4-1 have not yet been documented. Results are contained in GA notebook No. 7443.
Predictions of Kr85 release fractions were made as a function of temperature for CHST conditions using FSV FSAR fuel failure and fission product release assumptions. It was assumed for these calculations that temperatures increased linearly with time from 1100 to 2500 C in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. The predictions are compared with experimental results for TRISO (Th/U)C2 and TRISO UC2 on Figure 9. Predicted release fractions are greater than observed for both.
types of fuel, which leads to the conclusion that FSV FSAR fuel failure and fission product release assumptions are conservative.
In summary, experimental results obtained from tests conducted under simulated core heatup conditions have shown that:
- Kr85 release fractions from both TRISO UC2 and TRISO (Th/U)C2 fuels are much lower than predicted using FSV FSAR fuel failure and fission product release models.
References
- 1. J. L. Kaae, "A Mathematical Model for Calculating Stresses in a Four-Layer Carbon-Silicon Carbide Coated Fuel Part cle", Jou. Nucl. Mat., 32,
- p. 322 (1969).
- 2. W. J. Kovacs and D. P. Harmon, " Spitfire Loop Experiment SSL-2: Pre-irradiation Evaluation of Fuel", GA-A13520, September 30, 1975.
3.
"HTGR Generic Program Technology Program, Fuels and Core Development Program, Quarterly Progress Report For The Period Ending May 31, 1978",
GA-A14953, June 1978.
- 4. "HTGR Generic Program Technology Program, Fuels and Core Development Program, Quarterly Progress Report For The Period Ending August 31, 1978",
GA-A15093, September 1978.
5.
"HTGR Generic Program Technology Program, Fuels and Core Development Program, Quarterly Progress Report For The Period Ending November 30, 1977",
GA-A14744, December 1977.
6.
C. B. Scott and D. P. Harmon, " Post Ir:adiation Examination of Capusle F-30", GA-A13208, April 1, 1975.
FIGURE 1 CUMULATIVE DISTRIBUTION DEFINING IRRADIATION TEST EXPERIENCE FOR REFERENC,E TRISO C0ATED UC2 FUEL 1
m MEAN FISSILE BURNUP
$ IN 6 YR. OLD FUEL ga .8- .
j
=9 g :>
-! NN 2.5 x 10 5 PARTICLES l
y@ ! (2.4 x 105 at time of aE .'6 j NUREG-OH 1 )
NO !
$e aH i
j d$ !
$s .4 - l 86 i zE i Sd H o i y .2- ! PEAK FISSILE BURNUP N
i IN 6 YR. OLD FUEL i
- 9 x 103 PARTICLES
! , (None at time of 0 , ,
i s
NUREG-Olli) 40 50 60 70 80 BURNUP (7. FIMA) 316,000 TRISO COATED UC2 PARTICLES TESTED BETWEEN 885 - 1240 C AND 3.5 - 12.2 x 10 25 n/m2 (E > 29 fJ)HTGR
FIGURE 2 SCHEMATIC DIAGRAM DEFINING METHODOLOGY USED IN PRESSURE VESSEL PERFORMANCE CALCULATIONS EXPLICIT FUNCTIONS
, TRISO = git,. p;. . . TEMPER ATURE. BURNUP. FLUENCE) 9 = ARBITRARY FUNCTION i =
INDEX CEFINING I* COMPONENT PROFERTIES t; = DIMENSION DISTRIBUTIONS FOR CalTICA'. CC.'.tPONFNTS: KERNEL, BU F F E R. IFyC. sic Pi = DENSITY DISTRIEUTIONS FOR CRITICAL COMPONENTS: KERNEL, BUF F E R. IPyC. sic a = MAX 1 MUM sic TENSILE STRESS MONTE-CARLO TYPE CALCULATIONS FOR VARIOUS PERMUTATIO.*lS OF RANDOM m
VARI ABLES t, AND p;. e.g.
$ 3 , 3 4
3 43 D 4 4 j 1 h 1 1 2
8 2 2 2
C
.s_
5- KERNEL BUFFER IPyC sic O DIAMETER THICKNESS THICKNESS THICKNESS HTGR FUEL PARTICLE POPULATION STRESS DENSITY DISTP.lSUTION s
3- FAILURE CRITER10N om &
E5 r- i NU "'
AREA DEFINES EXPECTED PRES $URE VESSEL FA; LURE
{-} -} l f# c(2) l I c-fil !
MAX 1MU'. TENSILE STRESS. e e
-4 g _
- PREDICTED
- -O- O
-6 R/8 (Kr-85m) 10 I EoL W
~
YO " .
~7 10 1300-PEAK FUEL
.I20O- '~~~'""""""*"""""'"'"
TEMP. ( C) 110 0 -
8-FAST FLUENCE' G-
> 0. 8 MeV) 6 do ido do abo 250 EFFECTIVE FULL POWER DAYS FIGURE 3 IN PILE IRRADIATION TEST RESULTS FOR SSL-2 CONTAINING TRISO COATED UC2 AND BISO COATED Th02
-4
'10 37 _,,
p,EASURED
.G-PREDICTED 4
, 2-g a
-5~'~ .
_ c- <
.. - V /
E G-
,c-to co 4-
- EOL.
n - .-
$ 2xf co -
N -G -
cc 10 p ,
G, -
A - .
2: _
-7 ~, >
(O < < < < < >
Q ,1 2 3 4 5 o 7 8 9- 10 I AST FLUENCE ,10 'n/cm* (E > 0.18 MeV)p7ag FIGURE 4 MEASURED AND PREDICTED Kr-85M R/B VS. FAST FLUENCE FOR CAPSULE GF-4, CELL 2 CONTAINING VSM UC2 TRISO/Th02 TRISO FUEL.
GRAPHITE HEATING ELEMENT TA TUBE MULLITE TEST T'IBE SAMPLE HE IN (
HE OUT CUPROUS OXIDE (500*C)
> DESICCANT (20*C)
> ACTIVATED CHARCOAL (20*C) s
' LIQUID N2 -
IONIZATION -
IONIZATION -
COLD TRAP '
CHAMBER CHAMBER FIGURE S SCHEMATIC DRAWING OF THE CHST SYSTEM
1.0 . , ,
i . . n y TRISO UC2 0 e 6
z FISSION DENSITY: 6.0 X 10 0.8 - LENGTH OF CHST -
0 y 8.0 HRS D '
$ 70.75 HRS 0 o
E 0.6 - -
5 d
a:
$ 0.4 - - -
a 0
0 0.2 -
0 0
0 ' ' ' ' ^^^ ^ U 1100 1300 1500 1700 1900 2100 2300 2500 TEMPERATURE (*C)
FIGURE 6: Kr-85 RELEASE DATA COLLECTED DURING 8 AND 70.75 HR CHSTs 0F TRISO UCp FUEL IRRADIATED TO A FISSION DENSITY OF 6.0x1026
1.0 , . . . . . g 7
0.8 -
g, jg
! Q U
0.6 - 'I U
a E 0.4 - -
9 0.2 - -
0 ' ' ' ^#2 h 1100 1300 1500 1700 1900 2100 2300 2500 TEMPERATURE (*C)
FIGURE 7
SUMMARY
OF Kr-85 RELEASE DATA COLLECTED DURING CHSTs 0F TRISO UC2 FUEL
O .
D E
. T C
E P
X E
S L N E 0 O U I F
- - 0 T 5 C 9 4\ 2 A C R
F U N
/
E h 0 S T
, 0 A
E
' 3 L 0 2 E S R I R
5 T 8
- F 0 r O 0 K 1 T R S 2 O H F C
) E M C C O 0
- N R 2 0 ( E F C 9 D I D U 1 E F E R N N O U O I O ,
S I
R T C 2 k\ 0 0
T A
R E
C A
% B 0 O 9
T
) )P \a 7 P A T A
0 , U
,U-1 M A 9E E T D C HT T A TN TA H AE (E E T D H 0 G I N W EI O . i 0 A GF SR i 5 R L Nl t I l 1 E AO R E U RC T (8 H F T
2 N
\ @ 0 F 0 U C
\
\
0 3 t f
0 S O
\
\
1 S I I R
\s R T A
P M 0 M O
. - - - 0 O R 1 C F 0 8 6 4 2 0 1 1 0 0 0 0 8 E
z2 g3 i* $ R U
G I
F
10 i i : -
i f -_ . . . _ _ .
Y PREDICTED KR-85 0.8 -
~ ~ ~ ~ -
RELEASE FSV FSAR Q METHODS ,
. 2 N 0.6 -
- _ 2 _
[ _-_ _- DATA '-
M - -
0.4 -
m -
7 ~
$ 0.2 --
{ -
0' 1100 e
1300 1500
- gNgg\-
1700 1900 2100 2300 2500 TEMPERATURE (*C)
FIGURE 9: COMPARISON OF Kr-85 RELEASE FRACTIONS PREDICTED USING FSV FSAR FUEL FAILUPE AND FISSION PRODUCT RELEASE ASSUMPTIONS WITH EXPERIMENTAL DATA OBTAINED FROM CHSTs CONDUCTED ON TRISO UC2 AND TRIS 0 Th/U C2 FISSILE FUEL
TABLE 1 PROPERTY VALUES FOR TRISO COATED UC2
[ KERNhLPROPERTIES IPyC PROPERTIES lBUFFERPROPERTIESl _
- THICK- THICK- ANISO- n l DIAMETER DENSITY NESS DENSITY NESS DENSITY TROPY l CATEGORY (pm) (Mg/m3 ) (um) (Mg/m3 ) (um) (Mg/m3 ) (BAFo ) ,
_ _ , _ . 1
._ .: ,1= --_ -
NOMINAL PARTICLE 180 - 210 10.5 - 110 0.80 - 35 1.90 1.06 I l PROPERTIES FOR 11.0 1.10 REFERENCE DESIGN (
___ce
_ cr _ . _ . = _ _ , , . _ - _
.__m .
a _ __ _ __ .
i L RANGE OF IRRADIA- 91 - 240 10.0 - 35 - 0.90 - 14 - 36 1.80 - 1.03 -
R i TION TEST 11.3 110 1.40 1.95 1,08 d EXPERIENCE l h
, - , -. __ , , , - .1.. . . rm , m- . ~ . - _ - .m _ _ _ _ - _.
t
- CHST SAMPLES
- 1 L
2 6151-18-015 209 11.27 108 1,05 27 1.90 1.04 o 1
] 6151-17-015 204 11,27 98 1.08 36 1.87 1,04 !
d 4161-01-030 203 10.90 87 1.26 28 1.91 1.04 f i
l l ,
1
- w-PRESENTED FOR COMPARATIVE PURPOSES TO REFERENCE DESIGN
TABLE 2 PROPERTY VALUES FOR TRISO COATED UC2 s
ms u m_ _ _ _ ... m- -z __m f sic PROPERTIES l I DEPOSI- ,
PR ERTIES l g
TION c.=. -
- i THICK- TEMPERA- THICK- ANISO- ;
y NEFS DENSITY TURE NESS DENSITY TROPY [
q CATEGORY (um) (Mg/m3) (oc) (um) (Mg/m3) (BAFo ) f
. = _ . - - -
2 -a _- . ..
ll NOMINAL PARTICLE 35 3.20 1450 - 40 1.90 1.04 !
8 t
i PROPERTIES FOR 1700 ,
i REFERENCE DESIGN h 5
. =. , ,. ,. ,
, . , , . ,# . - _ . - i - . . m r = ,= ,= -
RANGE OF IRRADIA- 16 - 39 3.17 - 1400 - 19 - 50 1.50 - 1.02 - i l TION TEST 3.22 1750 1.96 1.07 EXPERIENCE l g
- - - ..u.m ; u w - = _, _
._ - m .-i .c , : 1 .r.. . _ . _-. . , s .i ,i, .-----..r---
i CIIST SAMPLES
- g 6151-18-015 39 3.20 1650 41 1.81 1,04 'I 6151-17-015 37 3.21 1450 38 1,81 1.03 l l 4161-01-030 29 3.20 1525 38 1.80 1.03 l
[ --.. --
m - -
PRESENTED FOR COMPARATIVE PURPOSES TO REFERENCE DESIGN
TABLE 3 COMPARATIVE EVALUATION OF PRESSURE VESSEL PERFOMIANCE FOR UC 2 /(Th,U)C2 FUEL PARTICLE SYSTEMS w._ __ . . ._ , -_ =_ _m_w w.-y_x= , m = = c. um,= = .xn l NOMINAL PARTICLE DESIGN EXPECTED PRESSUR'E KERNEL DIAM./ BUFFER VESSEL PERFORMANCE (#)
PARTICLE TYPE THICKNESS / PARTICLE DIAM. , (FAILUREFRACTION}_.,.f
?
TRISO (Th,U)C2 257/52/541 0,005 (c)
)
f(DS184A) 1
'm' .'5R M EW F". Zr kav'.daEL12EEWE'A ~ M'Marm."'. *"5."mEPPa "D k " 23.h.7 '.."1.; L9 "i'I2nr.*im".! L 1M TRISO UC2 195/110/635 0.002 (c) ;
., (DS 204A) b) s
, - . _ . - . - - .= m e., -- - - , _
= - - - - ~ . a (a) sic STRESS DISTRIBUTIONS CALCULATED FOR FUEL EXPOSED TO 8 x 10 25 n/m2 (E > 29 fJ)HTGR, 1250 C, 20% FIMA IN (Th,U)C 2, AND 75% FIMA IN UC2 (b) DESIGN CALCULATION NUMBER (c) BASED ON TRISO C0ATED PARTICLE STRESS CALCULATIONS DESCRIBED IN FIGURE 2 AND EMPIRICALLY DETERMINED FAILURE CRITERIA.
TABLE 4 SUFNARY OF FUEL TYPES AND NOMINAL PARTICLE DIMENSIONS IRRA3IATED IN SSL-2 (REF. 2)
Irradiation Conditions Number of Nominal Particle Dimensions Peak Fuel Fluence Particles Particle Type (um) Temperature Burnup 1025n/m2 Tested (Batch Designation) Buffer IPyC sic ( C) (% FIMA) (E>29fd)llTGR Kernel OPyC 85,200 UC2 204 108 35 30 33 1200-1300 72% 9.1 (6151-17-202)
Th02 507 84 - -
76 1200-1300 3.6% 9.1 (6542-36-020)
TABLE 5
SUMMARY
OF NOMINAL PARTICLE DIMENSIONS FOR TRISO C0ATED UCp AND Th0p FUEL TESTED IN CAPSULE GF-4(CELL 2)
Irradiation Conditions Number of Nominal Particle Dimensions Design Flygnce Particles Particle Type (um) Temperature Burnup 104an/m2 Tested (Batch Designationi) Kernel Buffer IPyC sic 10PyC (C) (% FIMA) (E>29fJ)HTGR 7,000 UC2 (6151-17-020) 204 100 35 30 33 1100 75.5 9.5
'6252-03-0101 529 63 30 33 37 16,900 Th02 ( 6252-04-010 ? 528 63 30 33 39 1100 4.8 9.5 i,6252-05-010; 519 63 30 33 49
e TABLE 6
- KERNEL AND COATING PROPERTIES (") OF TRISO CORE HEATUP SIMULATION TEST 2 UC SAMPLES Y sic Y KERNEL BUFFER DATA Tl!ICK.
DEN DIA. DEN TIIICK. DEN. TIIICK. DEN TIIICK. DEN RETRIEVAL (pm) (pm)
NUMBER (Mg/m3) (pm) (Mg/m3) (pm) (Mg/m3) (12m) (Mg/m3) (Mg/m3) 87 1.91 28 3.20 29 1.80 38 4161-01-030 10.9 203 1.26 98 1.87 36 3.21 37 1.81 38 6151-17-015 11.3 204 1.08 .
11.3 209 1.05 108 1.90 27 3.20 39 1.81 41 6151-18-015 l
110 1.90 35 3.20 35 1.90 40 FSV( ) 11.0 195 1.00 (a) AVERAGE PROPERTIES (b) REFERENCE DESIGN
TABLt 7 IRRADIATION. CONDITIONS OF TRISO UC 2 CORE IIEATUP SIMULATION TEST SAMPLES IRRADIATION CONDITIONS FAST NEUTRON KERNEL DA'I'A OSUR RETRIEVAL BURNUP FISSION (")
NUltBER (10~' n/m ) (% FIMA) DENSITY 6
4161-01-030 1.2 23 6.0 x 10 6151-17-015 5.0 60 1.3 x 10 6151-18-015 5.0 - 6u 1.3 x 10 FSV( 26 Core Avg. 2.8 39 6.5 x 10 Avg., 6 yr fuel 4.9 71 1.2 x 10 27 Peak 8.0 77 1.3 x 10 3
(a) FISSIONS /m INSIDE TIIE sic LAYER (b) REFERENCE DESIGN
I' I
TABLE 8 TRISO UC 2 CORE IIEATUP SIMULATION TEST CONDITIONS TEST CONDITIONS DATA TEMPERATURE ( C)
RETRIEVAL CIIST NUMBER LENGTil NO. OF PARTICLES (IIRS) INITIAL PEAK NUMBER 4161-01-030 78 IIIR- A-4-1 200 8.0 1105 2d00 4161-01-030 78IIIR-C-4-1 200 70.75 1120 2495 78IIIR-B-5-1 100 28.25 1145 2395 6151-17-015 6151-18-015 78IIIR-B-4-1 100 20.0 1165 2430
,