ML20002C964
| ML20002C964 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/11/1969 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101150482 | |
| Download: ML20002C964 (19) | |
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. C0mp20y o.n... on.c.., 2,2 w..e u.cnio.n Avenu.Jac=.on. u.cn>o.n 4osoi. Ar.. coa. sir tas-osso February 11, 1969
'Dr. P. A. Morris, Director Re:
Docket 50-155 Division of Reactor Licensing DPR-6 ZEK United States Atomic Energy Commission Washington, DC 205h5
Dear Dr. Morris:
Attention:
Mr. D. J. Skovholt Transmitted herewith are three (3) executed and thirty-seven (37) conformed copies of a request for a change to the Technical Speci-fications of-License DPR-6, Docket No 50-155, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant.
The proposed change (No 17) vill enable Consumers Power Company to insert into the reactor at Big Rock Point fuel bundles having one or two removable fuel rods containing plutonium oxide. The plutonium fuel will be irradiated under a joint program sponsored by the Edison Electric Institute, General Electric Company, the USAEC and Consumers Power Com--
pany. 'ihis specific irradiation is one phase of a broader program which has,as its main objective, the demonstration of the economic and technical.
feasibility of utilizing plutonium in light-water reactors. This specific experimental program includes the design, fabrication and irradiation of a significant number of fuel rods and bundles in an operating reactor.
The program is aimed at obtaining. data on the performance of a reference plutonium fuel form to provide design information for a reload core pro-gram in a commercial reactor.
The Joint Committee on Atomic Energy and the Atomic Energy Commission, aiing with the whole nuclear power industry, have expressed great interest in an experimental plutonium recycle program. This. program is intended to be directly responsive to that interest and was brought to this point with concerted effort over a contracted time schedule.
It is our intent to insert approximately sixteen (16) plutonium bearing fuel bundles into the Big Rock Point reactor during our next refueling outage, which is curyen y scheduled for April,1969 We vould, therefore, be p&.U&
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Dr. P.' A. Morris -
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February 11, 1969 most appreciative of an expeditious handling of this Request for a Technical Specification Change.so that we might; receive approval before
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- April 1,1969 Yours very truly,
4 GJW/ map R. L. Haueter-Assistant Eleetric. Production
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Superintendent'- Nuclear i
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4 CONSUMERS POWER COMPMiY Docket No 50-155 Request for Change to the Technical Specifications License No DPR-6 ZEK For the reasons hereinafter set forth, it is requested that the Technical Specifications of License DPR-6, Docket 50-155, issued to Consumers Power Company on May 1,1964, for the Big Rock Point Nuclear Plant be changed as follows:
I.
Section 5 A.
In Section 5 1.1, change to read as follows:
" Fuel (Sintered Pellets or UO2 Compressed Powder) or UO -Pu0 "
2 2
B.
In Section 5 1 5(a), change to read as follows:
" Enrichment of Fuel approximate weight percent of U-235 from 2.6 to 5 2 inclusive. Approximate weight percent of Pu (fissile - Pu-239 and Pu-241) 1.0 to 4.0 in nor-mal (0 7 w/o U-235) UO "
2 C.
In Section 5 1 5, revise Figure 5 7 under E-G Fuel heading, Note B, to read:
"R - Removable fuel rods - low enrichment or UO -Pu0g rods."
2 D.
In Section 5 1 5, replace the present table of fuel bundle parameters with the following table:
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' FUEL BUNDLES-5 1 5 (contd)
Research and Development' Original Reload Reload Reload Centermelt -
Centermelt' General (A)
B&C E
E-G "D" Fuel.
Intermediate,
. Advanced 1
.Ceometry, Fuel Rod Array-12.x.12 11 x 11 9x9 9x9 11 x 11
'8 x'8
'7x7' Rod Pitch, Inches 0.533 0.577
'O.707-0.707 0.580 0.807.
0.921-Standard Fuel Rods per Bundle. 132 109 74 70 109
'36 29..
4' 11 5 12
.28"-
L20 3
3
- Special Fuel Rods Per Bundle 121 122 7
Spacers Per Bundle 3
5 3
3
'7 5
.5
)-
Fuel Rod Cladding
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Material 304SS Zr-2 Zr-2
'Zr-2
.304SS, Zr-2 Zr-2' Zr-2 Inconel 600 and/or
'Incolay 800 Standard Rod Tube Wall,'In.
0.019 0.034 0.040 0.040 0.010 to 0.030-0.035-0.040, Inclusive.
Special Rod Tube Wall, In.
0.031 0.031 0.040 0.040 0.010 to 0.030 0.035 0.040-Inclusive-i J
Fuel Rods 4
Standard Rod Diameter, In.
0.388 0.449 0.5625 0.5625' O.425 0.570
-0.700 Special Rod Diameter, In.
0.3g0 0.344 0.5625 0.5625 0.320 0.570 0.700-6,7 90-95 Inclusive 94 Pellet 94 Pellet UO Stacked Density, Percent 94 - 1 94 - 1 Pellet 90-95 Pelle't '94 Pellet
-85 Powder 85 Powder.
2 85 Powdered Theoretical Active Fuel Length, Inches Standard Rod 70 70 69.75 70 68 to 70, inclusive 66-67.3 65-66.3 Special Rod 59 (Corner) 64.6 Central 64.9 Central Fill Gas Helium Helium Helium IIelium Helium Helium Helium I Four special fuel rods at bundle corners are segmented.
Reload S.C,E, and EG fuel bundles may contain (in the corner regions of the bundle) four Zr-2 tubes having encapsulated cobalt targets 2
sealed within.
~
Reload E and EG fuel bundles have a special central fuel rod to which the bundle spacers are fixed. In addition, two of the interior 3
bundle fuel rods are removable and may contain UO -Puo fuel.
y 4
Special rods have depleted uranium.
In addition to special rods for reload E, reload E-G has four gadolinia containing rods.
4
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6 With 3%' dishing on selected rods.
uel r d stack density will vary from 82 -.92 percent th'eoretical by using annular, dished, or nondished pellets in selected rods.
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II.
Discussion - Reload "E" or "E-G" Fuel With UO -Puo2 Fuel 2
A.
Program Description One or both of the two removable rod positions in approximately 16 of the Reload "E" or "E-G" bundles will contain UO -Pu02 sintered pellet 2
fuel rods instead of UO2 fuel rods. The purpose of this modification is to utilize the removable rod positions to further demonstrate recycling of plutonium in thermal pcwer reactors.
B.
Description of Fuel The cladding and mechanical design of the mixed oxide fuel rods is identical to the design of the removable UO2 rods they will replace (Table 1). The fuel rods contain cold pressed and sintered UO -Pu02 pellets.
2 The pellets are prepared from mechanically blended ceramic grade UO2 and Puo2 Powder. Four types of mixed oxide pellets are used:
1.
Lov density, approximately 92% theoretical without dish.
2.
High density, 95% theoretical, with 3% dish.
3 Low density, approximately 92% theoretical, annular pellets.
The central annulus will remove approximately 5 5% of the pellet.
4.
Low density, approximately 92% theoretical, annular pellets.
The central annulus will remove approximately 18% of the pellet.
The weight of plutonium in each rod is constant and this is done by adjusting the weight fraction of Pu02 in normal UO2 in each type of pellet to achieve a constant plutonium loading.
The UO2-Pu02 fuel rods will be loaded in specified bundles. The removable UO2-Pu02 rods are identified by crossed grooves on the top end plug and by serial numbers on the top retainer and lower end plug.
Previous experience with UO2-Pu02 fuel containing small amounts of Pu02 indicate that the thermal performance is essentially identical to UO2 fuel.(1,2,3,$e,5) Therefore, for the solid dished and nondished UO -pug 2 2
pellets, the thermal performance will be identical to the Reload "E" and "E-G" fuel.
The two types of annular pellet fuel rods will both operate well below melting at 122% overpower 500,000 Btu /hr-ft2 (Table II) and will have lower peak temperatures at normal operating conditions than the standard dished or nondished pellet rods.
C.
Nuclear Design The v1ntonium concentration was set to achieve a local pv er factor of 1 3 the removable rod position.
The bundle array is shown
h on Figure I.
The local power distribution is shown in Table III along with local. power factors of E-G" fuel without plutonium rods. Both designs have corrections for water gap effects. The calculations were made with standard GE nuclear methods. However, it should be noted that-the plutonium rods will burn down almost twice as fast as the neighboring uranium rods or the 2 5 A/o U-?35 rod that it replaced.
'The addition of two plutonium rods per bundle is calculated to make the void, Doppler, and temperature coefficients slightly more nega-tive. The bundle reactivity is essentially unchanged.
D.
Thermal Hydraulic Analysis As : stated above, the' nuclear design of the Pu02-UO2 rod increased the locsl peaking (maximum rod power) factor in the fuel bundle from 1.2 to -
a maximum of 1 3 Consequently, these bundles will be placed in core posi-tions that have lower radial power factors. The resultant thermal hydraulic performance provides additional margin from the minimum critical heat flux ratio (MCHFR) limit, 1 5 at 122% overpower, due to the reductions of water quality in the bundle.
The effect of these assemblies on reactor thermal hydraulic performance has been evaluated with a predicted core configuration and the results indicate that the desired perfomance can be achieved within all reactor limits. During the refueling outages, after fuel inspection and prior to start-up, ccre analysis will be performed c.n the selected core configurations.
E.
Related Experience Approximately 470 Zr-2 clad fuel rods containing cold pressed and sintered UO2-Pu02 pellets have been irradiated in the Plutonium Recycle Test Reactor (PRTR) and SAXTON (Table IV). The irradiation of approximately 560 full-length mixed oxide fuel rods was initiated in October 1968 in the Garigliano Nuclear Power Station (SENN - Table IV). The PRTR and SAXTON experience with sintered pellet mixed oxide fuel is sum =arized below:
1.
PRTR Experience (Ref 1, 2)
As part of the Batch Core Experiment in PRTR, two sintered pellet rods of (Pu, U)02 are being irradiated in a 19-rod PRTR type bundle.
(See Table IV.) One additional rod, removed from a bundle at 41,h00 Mwd /MT peak burnup, was post-irradiation examined. Although operated at a peak rating of178kv/ft,therewasonlyrecrystallizationandequiaxedgraingrowthin the central region of the cold pressed and sintered pellet.
Such a micro-structure is consistent with fuel operating temperatures of 1500 C - 1600 C Scene discrepancy does exist,/M ratio of the (Pu, U)0 however, since temperaturer of maximum.
C maximum were expected. The 0 2 fuel, de-42000 termined post-irradiation, was 2.008. Fission gas release was 1.4% which is reasonable for equiaxed grain growth in the fuel.
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-5 No evidence was uncovered which would indicate that the lov enrichment =(Pu,U)02 rod was significantly different from a UO2 rod operated at:a comparable rating.
2.
SAXTON Plutonium' Project (Ref 3, h, 5)
As part 'of the SAXTON Plutonium Experiment, some low 4's enrich-ment'(Pu,U)02 rods,_ have been irradiated without any known failures.
(See Table.IV.)
Peak power levels of 9 sl6 kv/ft were experienced by someoftheserodstopeakburnupsofs22,2OOMwd/T.
One fuel rod which had operated at a maximum power. rating of 10 7 kv/ft was post-irradiation examined. The fuel microstructure at the maximum power location showed only equiaxed grain growth at the pellet central region. Such a microstructure is consistent with the calculated 0
heat rating which would predict a maximum central temperature of s1500 0
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at10.6kv/ft.
Zircaloy-2 clad fuel rods containing hot pressed pellets are under test in the PRTR, SENN, and Dresden reactors. The PRIR test rods have operated satisfactorily at aklinearpowerratingsupto21kw/ft to a peak exposure of 10,000 Mwd T.
The only known experience with annular mixed oxide pellets has
'been in the liquid metal fast reactor progrsss in the US and UK.
As far as is known, the results-of these irradiations have been satisfactory since, as a result of these tests, the UKAEA has selected annular pellets for the reference design for their liquid metal. fast breeder demonstration.
reactor (Ref 8, 9).
Tests of fuel rods containing annular UO2 pellets are summariced in Table V.
The annular pellet integrity was good and there was no gross redistribution of fuel in the central hole due to pellet cracking and falling into the void in any of these tests. Although failures were en-countered in_ thepe jests, the failures were caused either by fission product swelling ( U 1 or by fuel swelling due to an excessively rapid power ine eage to allow fuel redistribution in previously molten core fuel rods 121 None cf the failures were intrinsic to annular pellet fuel.
In fact, the annular pellets were slightly superior to solid pellets in accommodating fission products, and the use of annular pellets made it possible to operate molten core pellet fuel rods under proper operating conditions.
Annular pellets have been previously approved for use in the Big Rock Point reactor in the centermelt bundles.
F.
Accident Analyses 1.
Reactivity Excursion Analysis a.
Postulated Reactivity Accidents The Big Rock Point reactor ope.stes with one specified control rod withdrawal pattern. The control rods are grouped in banks of two or
t 6
- more; all the control rods in a. bank are withdrawn together, with a pro-cedural limit aof one notch between any two control rods in a bank. This sequencing prevents large control. rod worths; ho ever, an operator error or series of errors can result in larger worths. The possible control-rod drop situations and control rod strengths when the core is critical and at hot. standby are:
' Case 1:
In-sequence potential of 0.008 ok for drop from full-in. position to drive position.
Case 2:
In-sequence potential of 0.021 Ak for drop.from full-in to full-out.
Case 3: out-of-sequence potential of. less than 0.021 Ak for drop from full-in to fu'J.-out.
Case 4: Maximum theoretical worst case of about 0.0L. 'k.
Case 1 requires the following equipment malfunctions and operator error:
- 1) ' Control rod becomes uncoupled from drive.
2)
Control rod drive is withdrawn (in-sequence), but control rod hangs up temporarily.
- 3) operator does not notice that control rod is not following.
- 4) Control rod then unexpectedly releases and drops from full-in to position of the drive due to gravity.
Case 2 requires an additional operator error of withdrawing the control rod completely rather than concurrent with the bank.
Case 3 consequences are less than those for Case 2.
Case 4 is considered hypothetical as it requires still further compounding errors beyond those enumerated above, Case 2 at the hot standby condition was used for this na ysis.
These are the same conditions used by DRL in a previous analysis 13}7 t
At the present time, the core is licensed to contain six center-melt fuel bundles. Analysis is performed for a core of "E/E-G" fuel with the centermelt bundles and plutonium rods included. To prevent a large amount of centermelt fuel from being in the peak neutron flux during a 1
reactivity accident, the six centermelt bundles are to be loaded in the core in a dispersed array with a minimum center-to-center distance of h2 cm.
This restriction means that the closest centermelt bundle spacing vill be no closer than two bundles in the x-direction and one in the y-direction.
b.
Kinetics Calculations The most important parameters in a nuclear excursion kinetics calculation are:
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7-1): Quantity of reactivity insertion.
'2) Rate.of reactivity insertion.
- 3) Specific power dist *bution.
~4) ~ Doppler coefficient.
- 5) ' Resonance neutron flux distribution.
- 6) Initial power.
The only significant difference between the *" plutonium" core and **"E". core is in the specific power distribution. The " plutonium" fuel bundle local power factor is.about 8% higher than 'E" ' fuel. For a given reactivity excursion, this would increase the peak energy density in that. assembly as well as yield more fuel mass above some energy levels.
L Also, the heterogeneity of the plutonium fuel might slightly affect the Doppler coefficient under rapid transient conditions. The best estimate -
of this effect indicates-a ' reduction in the Doppler coefficient of 0.01%.
Even extremely conservative considerations would only indicate a reduc-tion of.O.5%. Both of these errors are within the uncertainties associated-with the Doppler coefficient. Therefore, this effect was not considered in this analysis, c.
Primary System Integrity As discussed 'at length in previous license applications for this plant, the integrity of the primary system depends upon the severity of any steam explosion. The severity of a steam explosion depends upon the following~ factors:
- 1) Time of fuel failure.
- 2) -Mechanism of fuel failure.
- 3) Amount of fuel failed.
- 4) Energy in the failed fuel.
- 5) Heat transfer rate to coolant.
- 6) system geometry.
As has been shown in previous applications, a severe steam ex -
plosion will result only_if there is a significant quantity of promptly dispersed fuel in the moderator.
There is little or no information avail-able on the effects of plutonium heterogeneity on prompt fuel failure.
Because of this lack of information, the most conservative assumption is that all of the plutonium fuel is promptly dispersed. This would lead to a maximum energy release of 16 5 Mw-see in the core. This energy would be quite diluted as it is contained in 15 bundles scattered about the core.
Choosing a more reasonable but still conservative failure threshold such as 150 cal /sn vould lead to an energy release of 10 5 Mw-sec in the core.
- " Plutonium"_ core contains the currently licensed core with 30 plutonium fuel rods distributed with two rods in each of 15 standard bundles.
- "E" core is the ' currently licensed core.
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8-t Taking into account the conservative. figures for energy release from the "E" core, or the "E" core with demonstration bundles, (calculated to be 47 Mw-see for an 0.021 Ak rod' drop at hot standby with all fuel' above 265 cal /sm being promptly dispersed in the moderator) and the energy in the plutonium fuel, the total prompt energy release in the core would
>be 63 5 Mw-sec.
g is slightly less than the 6h Mw-see used by DRL in previous analyses 1
d.
Conclusionn e
It is concluded that'the results of a postulated reactivity accident are slightly more severe _in the " plutonium" core than in the "E" However, the results are still within an envelope considered ac-core.
ceptable in granting the license for the "E" fuel.
It is also concluded that there is no. danger of breaching the primary system due to a credible reactivity accident with either' core loading.
2.
Loss of Coolant The loss-of-coolant accident was discussed at length in conjunc-tion with Change 14 which allowed insertion of Beload "E" fuel. The addi-tion of 30 UO2-Pu02 rods to the core will not increase the severity of the postulated accident. As mentioned above, in discussion of core thermal 4
hydraulics, these assemblies will be placed in core locations with lower power factors in order to readily meet thermal limits. Thus, the result.
of any postulated LOC accident-will be less severe because of the reduced
. bundle stored energy. For equal bundle powers in an "E-G" bundle with' and without the UO2-Pu02 rods, the peak clad temperatures are only slightly For example, for an average bundle (ie,. thermal pwer of. 2.68 '
different.
0 Mwt), the peak clad temperature changes from 1755 F.to 1790 F.
III.
Conclusions Based on the above. analyses and comparisons with "E/E-G" fuel, i
_ the following conclusions concerning the UO -Pu02 fuel rods are made:
2 1.
Fuel rod mechanical design is identical to "E/E-G."
r 2.
The local power factor is slightly higher for the UO -Pu0 '
2 2
rods than the UO2 rods in the _"E/E-G" design but the plutonium bundes will be located in radial positions so that the peak rod power will not exceed the' design peak poser for the "E/E-G" fuel.
(_
3 The local power coefficients for the UO2 rods are essentially unchanged.
h.
The data available for 1cw enrichment UO -Pu02 fuel indicate 2
that the performance is essentially identical to UO2 fuel; therefore, the
. eak fuel temperatures in the solid pellet UO -PuO2 rods are identical to J
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the UO2 fuel rods.
In the case of annular _ pellet UO -Pu02 fuel rods, the
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2 peak fuel temperature is lower than the.UO2 rods.
(See Table II.)
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5 Annular' pellets have shown good structural integrity during
. tests. Annular pellets also are capable of higher thermal ratings without the fuel becoming molten.
16.
The results of a postulated reactivity accident are slightly more severe in' the plutonium" core than the "E/E-G" reload. However,
' there is no danger of breaching the primary system due to a credible acci -
dent with either core loading. The severity.of a loss-of-coolant accident is essentially unchanged from "E/E-G" fuel vithout UCg-PuO2 rods.
Based upon the above considerations, we have concluded that.the -
- ' use of Reload "E" or "E-G" fuel bundles containing one or two plutonium
- bearing rods in the Big Rock Point reactor does not present a significant change in the hazards considerations described or implicit in the Final Hazards. Summary Report.
4 CONSUMERS POWER COMPMiY By:
Senior Vice President a
Date: February 11, 1969-Sworn and subscribed to before me this lith day of February 1969 Etary Public, Jackson County, Michigan My commission expires January 15, 1972 l
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Figure 1 Bundle Array E-G FUEL ENRICHMENTS + 2 Pu RODS
- In the 9 x 9 fuel array the following distribution is to be used:
Type 1 18 - 2.5 wt %.
2 32 - 3.4 wt %
3 25 - 4.5 wt %
4 4 - 35 g/ft cobalt Pu 2 - Pu Rods (Nat'l Uranium, plutonium) 4 1
1 2
2 2
1 1
4 1
Pu 2
2 2
2 2
1 1
1 2
3 3
3 3
3 2
1 2
2 3
3 3
3
-3 2
2 2
2 3
3 3
3 3
2 2
2 2
3 3
3 3
3 2
2 1
2 3
3 3
3 3
2 1
1 1
2 2
2 2
2 Pu 1
4 1
1 2
2 2
1 1
4 4
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i TABLE I-RELOAD "E" AND "E-G" FUEL DATA Cobalt UO -Pu0 "EG" Fuel Rods "E" Fuel Rods Rods-Robs 0.471 Fuel Pellet Diameter.
0.471 0.471 Rod Pitch', Inches.
0.707 0.707 0.707 0.707 Cladding Thickness, Inches 0.040 0.040 0.040 0.040 Clad Outside Diameter,-Inches 0.5625 0.5625 0.5625 0.5625 Active Fuel Length, Inches 70.0; Central 69.75; Central 68.62 Rod 64.9' Rod, 64.62 UO -Puo Fuel Material UO UO 2
2 2
2 92-95 Fuel Density, % of-Theoretical 95 90-95 Cladding Material Zr-2 Zr-2 Zr-2 Zr-2 Number of Rocs per Bundle 77 77 4
2 Enrichment (See Figures 5.7 & 5.8)*
Low-2.5%
Low-2.35%
Nondished-1 30 Middle-3.4%
Middle-2.93%
Dished-1 30 Annular High-4.5%
High-3.55%
0.1" hole 1 36' Annular 7
0.2 ' hole-1 1 59 Helium Fill Gas Helium Helium Fuel Bundle Fuel Rod Array 9x9 Weight UO and UO -Pu0 Per Bundle, Pounds
- 346, 2
2 2
Moderator-to-Fuel Volume Ratio 2.39 4
Number of Spacers 3
- UO -Pu0 Enrichment is percent fissile Pu (Pu-239 and Pu 241) in normal (0.7 w/o) UO
- 2 2
2 4
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TABLE II THERMAL PERFORMANCE CHARACTERISTICS OF RELOAD "E" 6 "E-G" FUEL M -PuO, FUEL 2
Fuel Pellet Diameter, Inches 0.471 0.471 0.471 0.471 0.471 Fuel Pellet Inside Diameter, Inches 0.0 0.0 0.0 0.1 0.2 Cladding Thickness, Inches 0.040 0.040 0.040 0.040 0.040 Cladding Outside Diameter, Inches 0.5625 0.5625 0.5625 0.5625 0.5625 Incipient Melting Temperature of UO, F 5080 5080 5080-5080 5080 2
Fuel Density, % Theoretical 95 90-92 92 95 92 92 Fuel Center Line Tegperature at 500,000 Btu /Hr-ft
'F 5040
>5080
>5080 5040 4600 3800 Fuel Center Line Tegperature at 410,000 Btu /Hr-Ft
'F 4250 4400 4400 4250 3900 3300 Heat Flux fgr Incipient Melting, Btu /Hr-Ft
>500,000 477,000 477,000
>500,000 540,000
'680,000~
Area Fraction Molten at Peak Heat Flux 0
0.04 0.04 0
0 0
4
Table III Plutonium Isotopics-Pu-239 89.0 A/0 Pu-241 1.0 A/O Pu-240
'10.0 A/O Local P/A-Hot (25% In-Channel Voids)-
1 2
3 4
5 10 1.214 1.050 1.217 1.148 '
E-G Fuel 1.217 1.059 1.238 1.178 '
E-G Fuel + 2 Pu Rods
.901
.969
.837 1.297
.968
.852 1.034 3
1.043 1:
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' TABLE IV - MIXED-0XIDE Pu RECYCLE EXPERIENCE - SINTERED PELLET FUEL Reactor Heat Peak Type Thickness-0.D.
den Active No. of Rating Burnup (mils)
Inches
%T.D.
Enrichment Fuel Length Rods Kw/ft MWD /T Reference.
PRTR Zr-2 30 0.566
-91.5 1.94% Puo 75.7 3
18 3,600 1, 2 2
Saxton Zr-4 23.3 0.391 94 6.6% Pu0 36.6 470 9-16
-22,200 3,4,5 2
SENN Zr-2 37 0.593 91.5 2.0-3.2% Pu 104.3 60
<17 6'
SENN Zr-2 37 0.593 94.5 1.4-2.8% Pu 106.2 504
<17 7
- All fuel, mechanically blended powders of UO & Pu0
- 2 2
- Irradiation started in October, 1968.
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2 4
' Table V 4
ANNULAR UO PELLET TESTS 2
0.D.
V/o No.
Peak Heat Peak Burnup Reactor Coolant Type In.
% T.D.
In Hole of Rods Rating, KW/ft MWD /T
.Ref.
ORR NaK Stainless Steel 0.755 95
' 25 4
9 1500.
10 ETR 2,000 psi Zr-4 0.28 &-
95
- 15 4
12 40000
_11 water
- 0.56 GETR 1,000 psi Zr-2 0.566 95 8
13 56 12'A)0 12 water I
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References 1.
BNWL-SA-1204 " Operating Experience with Plutonium Fuels in PRTR" by M. D.
Freshley & S. Goldsmith 8/25/67.
2.
BNWL-739 " Plutonium Utilization Progran Technical Activities Quarterly Report 12/67, 1/58, 2/68" 4/68.
3.
WCAP-3385-52 Saxton Plutonium Program Mechanical, Thermal and Hydraulic Design of Saxton Partial Plutonium Core" by E. A. Bassler,. et al 12 '65.
4.
WACP-3385-8 "Saxton Plutonium Program Semi-Annual Progress Report Ending 6/30/66" by N. R. Nelson 7/61.
5.
WCAP-3385-12 "Saxton Plutonium Program Semi-Annual Progress Report for the Period Ending 6/30/67" by R. S. Miller & J. B. Roll 8/67.
6.
Cooperative Program on Pu Recycle between GE and ENEL.
7.
Cooperative Program on Pu Recycle between U.K., GE, and ENEL.
8.
J. Simmons, et 1, " Visit of AEC Fast Reactor Fuels Team to Installations in the U.K. and West Germany," May 23 - 31, 1968; August 20, 1968.
9.
K. W. Swanson, J. K. Butler & J. A. L. Robertson, " Mark II Subassembly Report on Examination of DFR-ll4 at 7.3% Maximum Burnup," TRG-Memo-4073(D); July,1967.
10.
R. F. Boyle, " Post-Irradiation Examination of ORNL Group II ORR Capsules,"
GEAP-3813, Oct. 1961.
11.
E. Duncombe, et al," Comparisons with Experiment of Calculated Dimensional Changes and Failure Analysis of Irradiated Bulk Oxide Fuel Test Rods Using the CYGRO-1 Computer Program," WAPD-TM-583, Sept. 1966.
12.
M. F. Lyons, et al, " Molten Fuel Rod Operation to High Burnup," GEAP-5100-2.
13.
" Safety Evaluation by the Division of Reactor Licensing, Docket No. 50-155, Consumers Power Company, Proposed Amendment No.
1."
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Consumers Puser Osmpany Feb 11,1968 Feb 13, 1999
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JOckson, Nishiges (R. L. Haueter)
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Dr. Peter A. Merris 3 Signed 37 esaf'd cys Actiose NECESSARY CCNCl.PRENCE G
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No ACT$oM SECESSARY O C<*+ <=r O
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30-155 4'Lt7dy.'"cMgA"M'.17 to the tech autmacoTo oArt atCrivro av oArc cpa s to permit licensee te insert Zieman 2-13 icto the reactor feel humales having 1 or 2 removabis fuel rods ceassiming w/9 cys for ACT105
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Into Coples to I,_Pr__1cs 6 Utaff Boyd Skovbc it Dube/Ievine Saltnr D. Thcapsoa awa s:
DISTRIBUTION:'-21-reg file 1-AEC PDE 2-Complimace 1-occ
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