ML20002C963
| ML20002C963 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/12/1969 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101150478 | |
| Download: ML20002C963 (18) | |
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]O P0Ylci Consumtrs
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C0mpBRy General Offecess 212 West Michigan Avenue. Jackson. M6ch8 gen 40201. Area Code 517 7tl8 OSSO
. February 12, 1969 T' < <
Dr. P. A. Morris, Director Re: Docket 50-155 Division of Reactor Licensing DPR-6 ZEK.
-United States Atomic Energy Commission Washington, DC 20515 4
Dear Dr. Morris:
Attention: Mr. D. J. Skovholt Transmitted herewith are three (3) executied and thirty-seven (37) conformed copies of a request for a change to the Tech-nical Specifications 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 18) will enable Consumers Power Company to insert into the reactor at Big Rock Point a fuel design, designated as " Modified E-G," which will permit the irradiation of four fuel bundles with up to 2)+ fuel rods with various mechanical properties, in removable rod positions in the fuel bundle. The purpose of the fuel rod irradiation tests, described herein, is to evaluate the irradiation-damage-induced mechanical property changes in zirconium-base alloy cladding as a function of alloy composition and fabrication method.
It is cur intent to insert " Modified E-G" fuel into the Big Rock Point Reactor during our next refueling outage which is currently scheduled for April 1969 We would, the efore, be most appreciative of an expeditious handling of this Request for a Technical Specification Change so that we might receive approval before April 1, 1969 Yours very truly,
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R. L. Hnueter
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Production Superintendent -
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CONSUMERS POWER COMPANY 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, Ibcket No 50-155, issued to Consumers Power Company on May 1, 1964 for the Big Rock Point Nu-clear Plant be changed as follows:
I.
Section 5 A.
In Section 5 1.1, change Structural Components to read as follows:
" Structural Components (fuel cladding vill, in addition to 304 SS and Incoloy 800, include Zr-2,.Inconel 600, and
.Zr-3Nb-1Sn."
B.
In Section 5.15, change "(c)" to read as follows:
(c) Fuel Bundles "The general dimensions and configuration of the seven types of fuel bundles shall be shown in Figures 5 2, 5 3, 5.h, 5 5, 5 6, 5 7, 5.8 and 8.1 of these specificittons. Prin-cipal design features shall be essentially as follows:"
C.
In Section 5 1 5, add Figure 5 8.
D. -In Section 5 1 5, replace the present table of fuel bundle parameters with the following table (next page).
E.
In.Section 5 2.1(b), in column titled " Reload
'E' Fuel," change to " Reload ' E, '
'E-G' and Modified E-G fuel."
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a...! ' t.ti) 71lE1, Bt'ND1ES Research and Development Original Reload lieload Feload Cente rzelt Centerselt General (A)
B4C E
E-G
- D" Fuel Intermediate
' Advanced WifM E4" "t. cow.etry, Fuel Rod Array 12 a 12 11 s 11 9x9 9m9 11 x 11 8m8 7m7' 9y9 Rod Pitch, Inches 0.533 0.577 0.707 0.707 0.580 0.807 0.921 o,707 109 36 29 p
Standard Fuel Rods per Bundle 132 109 74 70.5 28' 20" 3 7 Special Fuel Rods Per Bundle 123 122 73 gg1 12 Spacets Per Bundle 3
5 3
3 7
5 5
3 Fuel Rod Claditet Zr-2 with various Material 30455 Zr-2 Zr-2 Zr-2 30455, Zr-2 Zr-2 Zr-2
. initial sicchanical Incomel 600 and/or E#
P'*
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Incoloy 800 2r-32-1Sn StEndard Fod Tube Wall, In.
0.019 0.031 0.040 0.040 0.010 to 0.030 0.035 0.040 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 0.040 Inclusive Fuel Rods 0.5625 Standard Rod Diameter, In.
0.388 0.449 0.5625 0.5625 0.425 0.570 0.700 0.5625 94 Pellet' Spscial Rod Diameter, In.
0.3}0 0.3}4 0.5625 0.5625 0.320 0.570 0.700 tJo Stacked Density, Percent 94 - 1 94 - 1 Pe!!at 90-95 Pellet 94 Pellet.7 90 95 gogg,,g,,
94 p,gg,g 94 p,gg,g 6
85 Powder 85 Powder 2
85 Powdered n eoretical 70; Active Fuel Length, Inches Standard Rod 70 70 69.75 70 68 to 70, inclusive 66-67.3 65-66.3 6k.9 central, 68.6 Removable
}!elium Special Rod 59 (Corner) 64.6 Central 64.9 Central Fill Cas Helium Helium He!!um Belium Helium Helium Helium 8 Four special f uel rods at bundle corners are segmented.
Reload B.C.E. and EG fuel bundles may contain (in the corner regions of the bundle) four 2r-2 tubes having encapsulated cobalt targets 2
ses, led within.
In addition, two of the interior T Reload E and EC fuel bundles have a special central fuel rod to which the bundle spacers are fixed.
fuel.
bundle fuel rods are removable and may contain t'op Puoy Special rods have depleted uranium.
O In addition to special rods for reload E, reload E-G has four gadolinia containing rods.
i With 3% dishing on selected rods, f uel rod stack density will vary f rom 82 - 92 percent theoretical by using snnular, dished, or nondished pellets in selected rods.
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Discussion
" Modified E-G" Reload Fuel A.
Program Description Four (k) fuel bundles designated " Modified E-G" bundles are proposed as carriers for test fuel rods.
The purpose of the fuel bundle irradiation tests is to evaluate the irradiation damage induced mechanical property changes in zirconium base alloy cladding as a function of alloy compo-sition and fabrication method.
The initial mechanical properties of the Zr-2 tubing to be used for fuel cladding vill be varied by thermomechanical treatments.
Four types.of zirconium alloys (three are of basic Zr-2 with variations in thermomechanical processing; the fourth type being the Zr-3Nb-ISn alloy) would be incorporated into these fuel bundles along with normal Zr-2 fuel rod cladding.
Figure 1 summarizes the proposed test matrix and position of the test fuel rods within the bundle.
4 The Zr-3Nb-ISn alloy has the best combination of corrosion resistanee 'and high-temperature strength of the potential alternate zirconium base water reactor cladding alloys and has been successfully tested as cladding in the KAHL-RWE boiling water reactor by Meta 11gesellschaft A.G., Frankfurt Am Main.
B.
Fuel Description The four demonstration fuel bundles (Figure 5 outline draw-ing) are physically the same as the Reload "E" and "E-G" fuel.
Differences are summarized below:
- Modification of the Reload "E-G" mechanical design to include:
Addition of 16 removable peripheral fuel rod positions; Use of four removable interior fuel rod positions; Use of four corner removable thermal neutron absorber rod positions for nonfuel samples.
- The enrichment distribution and local peaking factors have been changed to enable as many of the 20 test fuel rods per assembly as possible to operate up to, but not exceed-ing, a maximum stgady state Pie.fr I;yal of 17.7 kv/ft or h10,000 Btu /hr-ft surface heat flux, vnidi is the steady state limit set for the Reload"E" and "E-G" fuel.
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- The position within the fuel bundle of the gadolinia con-taining fue.1, rods has been changed. The fuel rod design and gadolinia distribution, however, are the same as the Reload "E-G" fuel.
Figure 2 shows the tentative standard and test fuel rod _ positions and enrichment distribution. The configuration of the test fuel rods will be identical to the Reload "E-G" fuel rods with the exception of the removable rod feature. UO2 pellet fuel will be utilized with no deviations from the "E-G" except for fuel enrichments. Table 1 summarizes the fuel data for all rods. The cladding operating stress criter'. used for the "E-G" fuel rod design will be the same for the " Modified E-G" test rods.
The Big Rock Point Reactor is scheduled to continue to operate at-a _ coolant pressuge of 1,350 psi, giving a saturated water tem-perature of N585 F.
The reactor will be operating on 10-month to 1-year cycles starting in April 1969 The " Modified E-G" bundles will be designed to operate for three cycles and should achieve exposures of 20,000 Mwd /T average by the first quarter of 1972.
Interim test fuel rod removals and inspec-tions during outages are planned.
The design of the four " Modified E-G" bundles and the Reload "E" and "E-G" bundles includes corner positions for nonfueled rods containing thermal flux suppressors.
In the " Modified E-G" bundles, these corner positions have been adapted for removable nonfueled' test rods and will be used for the eval-uation of Zr-2 mechanical property changes due to irradiation.
These rods will include a thermal neutron absorber equivalent to that in the Reload "E-G" bundles (35 g/ft cobalt). The rods will have the same dimensions and configuration as the Reload "E-G" corner rods. The design of the nonfueled test rods will-utilize thinner Zr-2 cladding (0.022-in min wall); however, the design will insure that no gross collapse or dimensional changes occur that would alter the nuclear or thermal hydraulic characteristics of the bundle.
C.
Nuclear Design The principal nuclear characteristics of the " Modified E-G" bun-dies have been calculated and compared to Reload "E" and "E-G" fuel and are summarized in Table 2.
The reactivity values for the " Modi 11ed E-G" fuel at all conditions are lower than for "E" fuel, resulting in ample core shutdown margin. The temperature and void coefficients of the " Modified E-G" fuel are more negative than for the "E" fuel. The_ Doppler coef-ficient of the " Modified E-G" fuel is essentially the same as the "E" or "E-G" fuel at all conditions.
The tentative cal-culated local power distribution of the test rods is shown in
. 6 Figure 3.. It should be noted that all interior rods are of relative power less than average (1.0).
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D.
Thermal Hydraulic Data The thermal hydraulic characteristics of the " Modified E-G" bundles are essentially identical to the Reload "E-G" fuel..
The only differences are the _ differences in local power dis-tribution which will be limited to a local peak-to-average ratio of less than approximately 1.h.
These bundles will be placed in core positions having lower radial power factors to compensate for the higher local peaking. Consequently, the bundle steam quality will be reduced, resulting in more thermal margin to the MCHFR (1 5 at 122% overpower). Core thermal hydraulic analyses have been performed on predicted core configurations and indicate all license limits will be met.
During the refueling outage, these analyses will be performed on the finally selected core configuration.
E.
Accident Analyses 1.
Reactivity Excursion Analysis a.
Postulated Reactivity Accidents The Big Rock Point Reactor operatec with one specified control rod withdrawal pattern. The control rods are grouped in banks of two or more; all the control rods in a bank are withdrawn together, with a procedural limit of one notch between any two control rods in a bank. This sequencing prevents large control rod worths; however, an operator error or series of er-rors can result in larger worths. The possible con-trol 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 Ak for drop from 1'ull-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 full-out.
Case 4: Maximum theoretical worst case of about 0.0k5 Ak, Case 1 required the following equipment malfunctions and operator error:
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7 (1) ' Control rod becomes uncoupled from control rod drive; (2) - Control rod drive is withdrawn (in-sequence), but control rod hangs up temporarily. Operator does not notice that control rod is not following; (3) Control rod then unexpectedly releases and drops from full-in to position of the drive due to gravity.
LN*e 2 requires an additional operator error of with-drawing the control rod drive 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 tot the hot standby condition was used for this analysis. These are the same conditions used by D for their analysis of the previous fuel submittals. 1)
[
At the present time, the core is licensed to contain six centermelt fuel bundles. Analysis is performed for a core of "E/E-G" fuel with the centermelt bundles and
" Modified E-C" bundles included.
j To prevent a large amount of centermelt fuel from being in the peak neutron flux during a 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 42 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:
(1) Quantity of reactivity insertion; (2) Rate of reactivity insertion; (3) Specific power distribution; (h) Doppler coefficient; r
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~(5) Resonance neutron flux distribution; (6) Initial power.
The only significant difference between the *"Demonstra-
' tion" core and **"E" core is in the specific power distribution. The " Modified EJ3" fuel bundle local power factor is about 16% higher than "E-G" fuel (Figure 3). For a given reactivity excursion, this would increase the peak energy density in that bundle as well as yield more fuel mass above some stated energy levels.
0.021 ok Rod Drop at Hot Standby "E" Core
" Demonstration" Core Peak Enthalpy (Cal /Gm) 450 h50 Mass of Fuel (Kg) Above:
h25 Cal /Gm 1.0-1.0 330 Cal /Gm 26 26 265 Cal /Gm 37 37 230 Cal /Gm 40 58 As can be seen, there has been an increase in the mass of fuel above 230 cal /gm but no increase in the mass at higher energy levels. It should be noted that the in-crease in mass above 230 cal /gm will occur only if a demonstration bundle is located immediately adjacent to a centermelt bundle. Howevet, even if all four demon-stration bundles are next to a centermelt bundle, it would not change the values shown above because of the transient power distribution.
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 follow-ing factors:
"" Demonstration" core would contain four " Modified E-G" bundles in the currently licensed core.
- "E" core is the currently licensed core.
9 (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 explosion will result only if there is a significant quantity of promptly dispersed fuel in the moderator.
For material to be promptly dispersed, it must attain an energy density of h25 cal /gm or more. The above table demonstrates there is little, if any, promptly dispersed material in all the considered conditions. It is also seen that the " Demonstration" core is identical to the "E" core in this respect.
i A large quantity of new transient test data has been o -
tained recently in the SPER1' IV Capsule Driver Core. 2-5)
These data, and earlier data, indicate that fuel sub-jected to a transient energy deposition of 275 cal /gm or less remaifa intact (is not dispersed) after the transient. Thic is cons stent with the most recent cal-orimetric data for UO (6{1 which indicate incipient melt-2 ingoccursatanenergylevelofabout269 cal /gm. Even if one promptly dispersed all of the fuel above 265 cal /gm, the energy contained in the dispersed fuel would only contain h7 Mw-sec. This is well below the 64 Mw-sec used by DRL in evaluation of other fuel submittalc.(1) 2.
Lons-of-Coolant Accident The " Modified E-G" bundles have fuel rods placed such that most of the highly peaked rods are on the periphery adjacent to the bundle channel, as shown in Figure 3 The fuel rod po-sitions that are adjacent to the bundle channel valls are the most efficiently cooled positions in the bundle in the event of a loss-of-coolant accident since the spray cooling water that enters the bundle cools the channel in a period of approximately 200 seconds, providing an excellent heat sink for the rods that are adjacent to,the channel wall. To demonstrate this, an analysis of the bundle thermal perform-ance during a loss-of-coolant accident transient of a standard "E" type fuel bundle, and a " Modified E-G" fuel bundle was made. The " Modified E-G" fuel maximum temperatures during the transient following a loss-of-coolant accident involving
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the maximum break' size (3 5 sq ft bottom break) was the same as that of the "E" fuel. However, the percent of "E" fuel rods at the maximum temperature was 55%, while the
" Modified E-G" fuel had only 5% of the rodc at the maximum
~ temperature. This type.of relationship between "E" fuel behavior and " Modified E-G" fuel behavior will be consistent through the range of break sizes.
III. Conclusions Based upon the above analyses and comparisons with "E" and "E-G" fuel, the foll.owing conclusions per;;ain to the " Modified E-G" fuel:
A.
The mechanical design of the " Modified E-G" fuel is essentially identical to the "E" and "E-G" fuel which is a well-proven concept and has_ proven very satisfactory based on experience with the "E" fuel to date in the Big Rock Point Reactor. The addition of the removable rod positions is'a minor mechanical modification.
B.
The thermal hydraulic performance of the " Modified E-G" fuel-will be within the limits set for the "E" and "E-G" fuel. The local power peaking will be higher,1.4 compared to 1.2, than for the_"E-G" fuel. Ther=al hydraulic calculations show that there is ample critical heat flux margin.
C.
The consequences of a loss-of-coolant accident are no more severe with " Modified E-G" fuel than-with "E" or "E-G" fuel. Safe performance of "E" reload fuel was demonstrated in the "E" license submittal.
D.
The consequences of a postulated reactivity accident are no more severe with " Modified E-G" fuel than with "E" or "E-G" fuel. The " Modified E-G" fuel bundles can be placed adjacent to each other or adjacent to a centermelt bundle with no ad-ditional risk.
It is also concluded that there is no danger of breaching the primary system due to a credible reactivity accident with either "E,'? "E-G" or " Modified E-G" fuel bundles in the core.
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Based upon the above considerations, we have concluded that
.the use of " Modified-E-G" fuel'in the Big Rock Point Reactor does not present n significant change in the hazards considerations described or implicit in the Final Hazards Summary Report.
CONSUMERS POWER COMPANY
~
By Uenior Vice President Date: February 12, 1969 Sworn and subscribed to before me this 12th day of February 1969.-
Notary Public, Jackson County, Michigan My commission expires January 15, 1972
~
TABLE I DEMONSTRATION FUEL DATA (Modified "E-G")
f Test f
Test l
Test Test Corner Standard Test Fuel Rods Fuel Rods Fuel Rods ! Fuel Rods ! Fuel Rods Fuel Rods Rods 1
Fuml Pellet Diameter 0.471 0.471 0.471 0.471 0.471 0.471
' O.707 0.707 0.707 Rod Pitch, Incher 0.707 0.707 i 0.707 0.707 Citdding Thickness, Inches 0.040 0.040 0.040 0.040 0.040
- 0. 0'e0 0.022 to j
l 0.040 L
l C' :d Outside Diameter, Inches 0.5625 0.5625 0.5625 0.5525 0.5625 0.5625 0.5625 Active Fuel Length, Inches 70.0; Central 68.6 68.6
,68.6
' 68.6 68.6 Rod, 64.9
' UO UO U0 UO 2
2 2
2 2
2
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Fu21 Material UO UO l
I Futl Density, % of Theoretical 95 95 95 95 95 l 95 I Zr-3Nb-lSn Zr-2 Zr-2 Cirdding Material Zr-2 Zr-2 Zr-2 Zr-2 1
B C
,D
,E A
Alloy Code (See Figure 1)
Number of Rods per Bundle 57 4
4 4
f4
!4 4
q Enrichment (See Figure 2)
Low
- 2.5%
4.0 4.0 36 36 l h.5 8
Middle - 3.4%
High - 4.5%
Fill Gas
!!elium IIelium llelium llelium
,llelium lllelium Fu?l Bundle Fuel Rod Array 9x9 Weight UO Per Bundle, Founds
-346 i
2 Moderator-to-Frel Volume Rati: 2.39 Number of Spacers 3
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' TABLE'2
~ COMPARISON OF PRINCIPAL CALCULATED NUCLEAR CHARACTERISTICS OF MODIFIED E-G FUEL WITH COBALT AND NOMINAL GADOLINIA "E"
"E-G" Modified E-G Reactivity (k.)
'68 F 1.268
-1.208 1.229 572 F, O Voids 1.280 1.203
-1.225 572 F, 25% Voida 1.262 1.183 1.206 Temperature Coefficient ak,ff/k,ff per-F @ 77 F
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Start of. Cycle
+ v.38 x 10
+ 0.27 x 10"
+ 0.13 x 10 Void Coefficient Ak/k per Unit Void Within Channel Cold (68 F)
- 0.07
. - 0.08
- 0.07 Hot (572 F)
- 0.11
- 0.12
- 0.11 Doppler Coefficient ok,7f/k per F ff Fuel Temp Moderator
~5
~5 5
68 F 68 F, 0 Voids
- 1.3 x 10
- 1.3 x 10
- 1.3 x 10~5
~5
~5
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1323 F 572 F, O Voids
- 1.0 x 10
- 1.0 x 10
- 1.0 x 10
~5
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~b 1323 F 572 F, 25% Voids
--1.2 x 10
- 1.2 x 10
- 1.2 x 10 t
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1 1h Figure 1 Modified E-G Bundle Incation of Test Fuel Rods a
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A A
D 2
E E
3 C
C 4
B B
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B 7
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9 D
A A
D STRENGTH LEVELS 70*F ALLOY RELATIVE 0.2% Y.S.
U.T.S.
CODE ALLOY STRENGTH PSI x 1000 PSI x 1000 A
Zr-2 Low 40 - 50 60 - 70 B
Zr-2 Low 60 - 70 80 - 90 C
Zr-2 High 80 - 90 100 -110 D
Zr-3Nb-1Sn High
, 90 -105 115 -130 E
Zr-2 Normal 70 - 80 90 -100
~
15 Figure 2 Modified E-G Enrichments Test Rod Positions 1
7 6
2 6
1 4
1 3
3 2
2 2
3 3
1 7
3 3
2 2
2 3
3 2
2 1
1 1
2 2
2 2
2 1
1 1
2 2
2 6
2 2
1 1
1 2
2 6
7 3
3 2
2 2
3 3
7 1
3 3
2 2
2 3
3 1
4 1
7 6
2 6
1 4
g No.
Enrichment (wt %)
1 17 2.5 2
28 3.4 3
16 4.5 4
4 test locations containing absorber 6
8 k.0 36 7
8
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16 Figure 3 Modified E-G Fuel j
Local Fuel Hod Relative Power Distribution 300*C - 25% voids 1
- Test Rods 0
1.17 1.4 1.19 1.17 1.3 1.17
.86 1.17 1.02
.74
.41
.86
.74 1.19
t IT References l.
" Safety Evaluation by the Division of-Reactor Licensing, Docket No.
50-155, Consumers Power Company, Proposed Ar.endment No. 1."
~
2.
ID0-ITR-100, " Transient Irradiation of 1/4 Inch 0.D. Stainless steel Clad Oxide Fuel Rods to 570 cal /g UO," October 1968.
2 3.
IDC-ITR-1-1, " Transient Irradiation of 0.466-Inch 0.D. Stainless Steel Clad Oxide Fuel Rods to 300 cal /g UO," Nov. 1968.
2 4.
IDO-ITR-102, " Transient Irradiation of 1/4 Inch 0.D. Zircaloy-2 Clad Oxide Fuel Rods to 590 cal /g UO," November, 1968.
2 5.-
IDO-ITR-103, " Transient Irtsdiation of.3125 Inch 0.D. Zircaloy Clad Oxide Fuel Rods to 45t, cal /g UO," To be Published.
2 and Tungsten R. A. Hein, P. N. Flagella; "Enthalpy Measurement of UO 6.
to 3250*K," Annual Meeting of Am. Cer. Soc., April 20 225, 1968.
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1 37 add'1 cys rec'd 3-24-69 Dr. Peter A. Morris ACTices NECESSARY O
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NO ACTION NECESSARY O CouuE=T O
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POST OFFICE F 4 LE CODEa CLA55 eF.
50-155 REG. Not g
REFERRED 70 OATE RECEIVED BY DATE MCAtFTION; (Mwet Untingsfed) gg request for Changs No.18 to the Ziemasun 3-25 tcch specs of Lis D m-es w/9 cys for ACTION
'"'afi$ Fast for Change to the Technical DISTRIstrTION:
Sp:cifications Lie DPR-6 to perinit M. Price b Staff CPC to insert " Modified E-G" inte Boyd the Big Rock Re ac ter......
Skovholt Dg_be h ias (3 Signed 37 coef'd cys rec'd)
Saltsunam R,_Thospeau
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