ML20002D875
| ML20002D875 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/15/1970 |
| From: | Walke G, Wall H CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101230254 | |
| Download: ML20002D875 (18) | |
Text
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Consumcrs l
Povici Company General Off ces: 212 West Mecnigme Avenue. Jackson. Michigan 49201. Area Code St7 788-05S0 December 15, 1970 Re~uld y File Cy.
Dr. P. A. Morris, Director Re: Docket 50-155 Division of Reactor Licensing DPR-6 ZEK United Ctetes Atomic Energy Commission Proposed Tech Spec Washington, DC 20545 Change 21
Dear Dr. Morris:
Attention:
Mr. D. J. Skovolt 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 1,196h, for the Big Rock Point Nuclear Plant.
This proposed change (No 21) will enable Consumers Power Company to insert into the reactor at Big Rock Point a fuel design designated as Reload-F. This fuel design is, in its performance characteristics, essentially identical to the Reload E-G fuel, currently the standard Big Rock Point reload fuel. The minor differences between the Reload-F and E-G fuels represent an attempt at optimizing the design.
It is our intention to insert Reload-F fuel into the Big Reck Point Reactor during our next refueling outage which is currently schad-uled for February 1971. We would, therefore, be most appreciative of an expeditious handling of this Request for a Technical Specifications Change so that we might receive approval before.Tanuary 15, 1971.
Yours very truly,
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GJW/ map erald J. Walke Nuclear Fuel Management C
Administrator
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. NN['I.O CONSUMERS POWER COf7ANY rr.. y Docket No. 50-155 Request for Authorization of Change in Technical Specifications Change No. 21 License No. DPR-6 For the reasons hereinafter cet forth, the following changes to the Technical Specifications of License No. DPR-6 issued to Consumers Power Company on May 1,1964, for the Big Rock Point Nuclear Plant are requested:
I.
Changes:
Section 5,_
A.
Delete present Figure 5 2, Original 4 fuel.
B.
Renumber present Figure 5 3, Reload-E fuel as Figure 5 2.
C.
Renumber present Figure 5.4, Reload-C fuel as Figure 5 3 D.
Transfer present Figures 5 5, 5 6, 5.8 and 5 9 to a new Section 8.
E.
Renumber present Figure 5 7, Reload-E and E-G fuel as Figure 5.h.
F.
Add new Figure 5 5, Reload-F fuel.
G.
In Section 5 1 5, change (c) to read as follows:
"(c) Fuel Bundles The general design and configuration of the five types of reload fuel bundles shall be ar. shown in Figures 5 2 through 5 5 (inclusively) of the speci-fications. Principal design features shall be as on Table 5 1."
H.
In Section 5 1 5, delete the present table and substitute the attached table, new designated Table 5 1.
I.
Transfer Sections 5 1 7, 5 1.8 and 5 1 9 to a new Section 8.
J.
In Section 5 2.1 (b), change the text to read:
"(b) Reactor Operation The reactor operation shall be so limited as to be consistent with the most conservative of parameters in Table 5.2 and Table 8.2."
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Designate the present. table as Table 5 2 and delete the column heading references to Original ("A") and," Modified
.' E-G Fuels" and ar" "F" fuel to Column 2.
II.
Changes:
Section 8 Delete the entire present Section 8 and insert,the attached new Section 8.
i A.
Renumber present Figure 5 5, 8 x 8 Fuel Lattice Centermelt Bundle as Figure 8.1.
i B.. Renumber present Figure.5 6, 7 x 7 Fuel Lattice Centermelt Bundle as Fis;ure 8.2.
C.
Renumber present Figure 5 8, Modified E-G Fuel as Figure 8 3 D.
Renumber present Figure 5 9, ESI-UO -Fu0 Fuel as Figure 8.h.
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Table 5 1,
.i REIDAD FU2L TfPES Gentral Reload (B and C)
Reload (E)
Reload (E-G and F)
Geometry, Fuel Pod Array 11 x 11 9x9 9x9 Rod Pitch, Inch 0 577 0 707 0 70Y 74 )
70(2,3)
Standard Fuel Reds per Bundle 109 )
7(2 11 12ll Special Fuel Rods per Bundle Spacers per Bundle 5
3 3
Fuel Rod Cladding Material Zr-2 Zr-2 Zr-2 Stsndard Rod Tube Wall, Inch 0.034 0.040 0.0ho Spee!al hod Tube Wall, Inch o.031 0.0ho 0.0ko Fuel. Rods Standard Rod Diameter, Inch 0.4h9 0 5625
- 0. %25 Special Rod Piameter, Inch 0 344 0 5625 0.5625 Fuel Stacked Density, Percent 94 ! 1 Pellet 90-95 Pellet (5) 94 "ellet(b> 5)
Theoretical 85 Powdered Active Fuel Length, Inches Standard Rod 70 69 75 70 Special Rod 64.6 Central 64 9 Centul Fill Gas Helk.m Helium Helium (1} Reload B, C, E and E-G fuel bundles may contain (in the corner regions of the bundle) four Zr-2 tubes having encapsulated cobalt targets sealed within.
(2) Reload E and E-G fuel bundles have a special central fuel rod to which the bundle spacerc +."e fixed.
In addition, two of the interior bundle fuel rods are removable and may contain UO -Pu0 fuel.
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(3)In eddition to special rods for Reload E, Reload E-G has four gadolinia-containing rods.
With 3% dishing on selected rods.
}UO -Pu0 2
2 fuel rod stack density will vary from 74 to 92% therletical by using annular, dished, or nondished pellets in selected rods.
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i III. rDiscussion: Organization of Technical 3pecifications The elimination of the present Section 8 is an attempt to remove a completely obsolete section from. the Technical Specifications..
Both phases of the oriEinal. Big Rock Point R&D Program.have been com-pleted and the material in tre present Section 8 no longer is relevant
.to present or future planr.ed
.-ograms.
The transfer of mnterial from Section 5 to the new Section 8 provides an opportunity to segregate material that is strictly R&D and I
. s' bject to frequent change, ' revision and additione from the material u
which changes little. It is our intention to continue to use the Big Rock Point reactor as a facility for the irradiation of various devel-opmental. fuel types. This organizational change will be helpful in that it will clearly segregate material that needs special attention by Tbchnical Specification users from material that constitutes the
. basic reactor fuel and is less changeable.
IV.
Discussion: Reload '7" Fuel 4
l.
'A.
Fuel Description The Reload "F" fc?1 1:: simcat identical to the Reload "E-G" The only 'iffererces in the '7" fuel are:
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-fuel.
1.
Gadolinia concentration is reduced to 1.25 weight percent, with no gadolinia in the upper or lower 7 78" of the active fuel length.
(See Figure 1.)
~2.
Four of the tie rods are moved one~ row toward the center of the assembly so they will not be at peak rod locations.
3 Enrichment distribution within the bundle is changed to improve power peakfug. Four fuel rods which were low enrichment rods in the "E-G" fuel are now middle enrichment rods.
i 4.
Middle enrichment rods are 3 3 v/o.U-235 instead of 3 4 w/o.
(Average enrichment of the fuel assembly remains exactly the same.)
1 Because the modifications to the reload fuel result in only very slight changes to the reactivity and its distribution within the bundle, there is no increase in the hazards acsociated with the Big Rock Point. Nuclear Plant.
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h B.
Thermal-Design A summary of nominal Reload "F' fuel data is given in Table 1, and a list of thermal performance characteristics is shown in Table 2.
Although the geometry of the "F" fuel is identical to the "E-G" fuel, the center line temperatures reported for '7" fuel are not the 4
same. In the license submittal for "E-G" fuel, the maximum temperature reported wasl50h0 F which occurred at a heat flux of 500,000 Btu /hr-ft at the end of the first cycle. Thble_2 indicates that the most limiting temperature is still at the end of Cycle 1, but the maximum temperature is now reported as 5203 F, which exceeds the melting temperature of UO '
2 The. reason for this reported difference is that calculations at peak heat _ flux formerly were me(e with the assumption that the pellet-clad gap was closed which improved the heat transfer across this gap. Present
- calculations do not take credit for improved gap conductance so, conse-4 quently, the fuel tet7erature is higher.
The significance of this new temperature calculation is actually very small and in no way places the integrity of the fuel in jeopardy.
In actual operation to date, the current license heat flux limit for "E" and "E-G"_ fuel has not been approached (nor is it anticipated that it will be approached in the future). Actual peak heat flux has been about 375,000 Btu /hr-ft compared to the license limit of 410,000 Btu /hr-ft.
Assuming a 122% overpower were to occur (although there is very low probability of this) from a peak heat flux of 375,000 Btu /hr-ft, there would be no melting of the fuel. The maximum center line temperature would be approximately h830 F, or about 200 F below the melting tempera-
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ture of UO after exposure for one cycle.
2 However, since the license limit for heat flux at overpower is 500,000 Btu /hr-ft, this condition is still evaluated.
It is now recog-nized that some melting 'of fuel could occur, but the volume of molten fuel would be so small that the integrity of the cladding still would not be endangered. At the maximum heat flux position, in the highest power rod, approximately. h% of the cross-sectional area of the fuel would be molten. Because of the axial power distribution, there would be molten fuel along less than 15% of the rod's length and a maximum of 0 5% of all s
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the fuel in the peak fuel rod would be' molten. The 10% increase in volume of this 'small amount of fuel, due to phase change, can easily be accommo-dated by the. porosity available in the fuel pellets which are dished 3%
in all peak power rods. These thermal performance limits for the "F" fuel are almost identical to the performance limits reported in licensing the low-density "E" fuel. The molten volumes at a heat flux of 500,000 Btu /hr-ft are both reported as %.5% and, because of the 3% disl.ing in i
the "F" fuel, the amount of porosity available to accommodate an increase in fuel volume is almost identical in both the low-density "E" and the "F" fuels. Therefore, the perfomance limit reported for the "F" fuel does not exceed the maximum limit of fuel previously licensed in the Big Rock Point Nuclear Plant.
Thermal conductivity of the fuel is the same as referenced far a
licensing of the "E" and "E-G" fuels.
C.
Fuel Physics Data The principal nuclear characteristics of Reload "F" fuel have been calculated and are compared with "E-G" fuel in Table 3 The uncon-trolled k. for "F" fuel is' reported separately for the 54.44' section of fuel containing gadolinium and for the two end sections not containing gadolinium. The combined effect indicates that k. is almost identical to "E-G" fuel. The fully controlled k. in the gadolinium section is
.883 which demonstrates ample shutdown margin. The reactivity for "F" fuel is lower at 68 F than at 572 F, whereas the "E-G" fuel showed a slightly higher reactivity at 68 F.
This slight difference can be pri-me.rily accounted for by the difference between "E-G" and "F" fuels in the axial distribution of gadolinium. The "F" fuel temperature coefficient reported in Table 3 is based on a just-critical combination of controlled and uncontrolled fuel. The uncontrolled temperature coefficient for "F" fuel would be approximately the same as previously reported for "E-G" fuel.
The void coefficient is also verr comparable to "E-G" fuel, showing a slightly more negative value at 25% voids. The Doppler coefficient is reported the same as for "E-G" fuel. The magnitude of the Doppler coeffi-cient is inherent in the fuel design and does not vary significantly for a light water reactor having low enrichment fuel. The buildup of Pu-2hO
'will increase the magnitude of the Doppler coefficient by 10% to 15% over the lifetime of the "F" fuel.
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D.
Thermal Hydraulic Data The ther=al hydraulic perfcr=ance of the "F" fuel vill be essen-tially identical to that of the "E-G" fuel. The local and axial power distribution vill be about the same er slightly lower peaking for the "F"
fuel. Upper and lower tie plates and spacers are essentially the same so there is no change of loss coefficients. Thermal hydraulic calculations for the present core show that there is ample critical heat flux margin for all types of fuel presently in core.
Because of the difficulty mf predicting core configurations in Big Rock Point, specific core analyses must be performed during the re-fueling cutages, after fuel inspection and prior to start-up. These analyses assure that all license limits are met by the selected core confi guration.
V.
Conclusions Based upon the above analyses and comparisons with previous fuel, the following conclusions may be dravn:
1.
Mechanical design of the 9P" fuel is essentially the same as "E" and "E-G" fuels which have demonstrated very satisfactory per-formance to date. in the Big Rock Point reactor.
2.
The limiting condition for the thermal design of the "F" fuel occurs at the end of the first cycle that the fuel is in the core.
Although it is highly unlikely that a heat flux of 500,000 Btu /hr-ft should be reached at that time, such a heat flux would result in a peak center temperature of 5203 F and a maximum molten volume of less than 0 5% of the fuel. The vcPtte change associated with melting such a small acount of fuel can easily be accommodated without ecmpromising the in-tegrity of the cladding.
3 Nuclear characteristics of "F" reload fuel are similar to "E-G" fuel, with some increase in shutdown margin and slight improvement in reactivity coefficients, both of which enhance control of the reactor.
Safe performance of "E-G" fuel has been demonstrated in a previous license submittal.
h.
The thermal hydraulic characteristics of the 9P" fuel are essentially the same as "E-G" fuel. The only differences are due to minor changes in power distribution within the bundle. Since there is ample critical heat flux margin for "E-G" fuel, there vill also be ample margin for "F" fuc'
s.
7:
Based upon the above considerations, we have coacluded that the use of Reload "F" fuel in the Big Rock Point reactor does not present a significant change in the hazards considerations described or in:plicit in the Final Hazards Summary-Report.
-CONSUMERS POWER COMPANY res ent l
Date: December 15, 1970 Sworn and subscribed to before me this 15th day of D3cember 1970.
W e)
\\ cuvvu >t I Notary Public, Jackson County, Michigan My Con: mission Expires January 15, 1972 1
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Axial Distribution of Gadolinitn for "F" Fuel (Four each per assembly).
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TABLE 1 Reload "F" Fuel Data Fuel Rod Description Zr-2
~ Cladding _ material Cladding thickness, inches 0.040 o
Cladding outside diameter, inches 0.5625 helium Fill gas Fue) material sintered, ground UO2 pellets Fuel pellet diameter, inches 0.471 Active fuel length, inches
!70.0 Central rod 64.9 Fuel density, % theoretical 95 Enrichment, wt. percent U-235 Low 2.5 Middle 3.3 High 4.5 Dishing (on selected rods), 1 3.0 Fuel Bundle Description Fuel md array 9x9 77 Number of fuel rods Low enrichment 16 Middle enrichment 36 (4 with 1.251, gadolinia)
High enrichment 25 Ntm >er of cobalt mds (35 gm/ft) 4
- ch, inches 0.707 h..,;t -T 30 Per bundle, pounds 346 2
Moderator :-fuel volume ratio 2.39 Number of Spacers 3
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7 TABLE'2 Reload "f 7hbcmAl Performance Beginning End of of Life (BOL)
Cycle 1 Fuel pellet diameter, inches 0,471 0.471 Cladding thickness, inches 0.040 0.040 Cladding outside diameter, inches 0.5625 0.5625 Fuel density, % theoretical 95 95 Fuel conductivityl
, Watts /cm 93 93 k
Fuel centerline temperature,*F 2
Rated-power (410,000 Btu /hr-ft )
2 122% Overpower 500,000 Btu /hr-ft )
5051 5203 Incipient melting temperature of 00, 'F 5080 5020 2
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TABLE 3 COMPARISON OF PFINCIPAL CALCULATED NUCLEAR CilARACTERISTICS OF "EG" AND "F" FUEL AT BEGINNING OF LIFE
..pn NO Gadolinium Cadolinium "E-G" Section Section
. Reactivity (Uncontrolled Ka.)
68*F 1.208 1.259 1.179 572*F, 0 Voids 1.203 1.273 1.183 572*F, 25t. Voids 1.183 1.257 1.171 4
Temperature Coefficient-(akeff/keff per *F) at 77'F Start of Cycle
.31 x 10 Void Coefficient (ak/k per Unit Void Within Channel) 68*F
.08
.08 572*F, 25% Voide
.12
.133 i
Doppler Coefficient akeff/keff per *F Fuci Temp.
Moderator 68'F.
68'F, 0 Voids '
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-1.3 x 10'
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1323*F 572*F, 0 Voids
-1.0 x 10
-1.0 x 10
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-l'.2 x 10
-1.2 x 10-5
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"8.0 RESEARCH AND DEVEIOPMENT From time to time, various developmental types of fuel will be irradiated in the Big Rock Point reactor. This section will deceribe these fuel types and any operating limitations associated therewith.
8.1 DEVEIOPMENTAL FUEL DESIGN FEATURES The general dimensions and configurations of the developmental fuel designs shall be as shown in Figures 8.1 through 8.4.
Principal design features shall be essentially as on Table 8.1.
8.1.1 Zr-Cr Alloy Test Bundle One of the reload fuel bundles may contain up to 18 rods (2 dummy and 16 fuel rods) clad with an annealed Zr + 1.15 w/o Cr alloy and up to 8 rods (2 dummy and 6 fuel reds) clad with annealed special Zr-2.
The remaining fuel rods in the bundle (maximum of 95) shall be clad with cold-worked standard Lr-2.
8.1.2 Thin Clad Powder Fuel Bendle Two of the Reload "C" fuel bundlea may contain standard rods with Zr-2 cladding of 0.025" thickness; otherwise, they will be the same as the remaining Reload "C" fuel bundles.
3 8.2 PRINCIPAL DEVEIOPMENTAL FUEL OPERATING LIMITATIONS The reactor operation shall be.so limited as to be consistent with the most conservative of parameters in Table 5 2 and Table 8.2.
8.2.1 Centermelt Test Fuel Bundles (a) Operating Limitations Six fuel bundles may be operated at increased thermal output with various amounts of centermelting of the UO. The fuel shall be 2
specially designed for this operation and shall be permitted to exceed
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the general ecr9 operating limitations of Section 5 2.l(b) but shall be limited to the noct conservative of the parameters listed on Table 8.2.
(b) Rate of Power Level Change Control rod withdrawal shall be limited as in Section 5 2.1.
In addition, when centermelt fuel is in the core, the rate of power in-crease between 170 MW and 240 MWf shall be limited to 1/2 MWt "#*#"8' t
per minute per notch of control rod withdrawal when any of the following conditions exist:
(1) A centermelt fuel bundle is being brought to l
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. power for the' first time.. or (2) a. scram recovery is being made ct a 1
time in the xenon. transient when the peak of the axial power distribu-
' tionfis lower in the core tha'n the peak existing at the time of the last shutdown.
(c) Puel Examinations Nondestructive examinations o ' each fuel rod in the center-melt fuel bundles shall be performed duri.ig each core refueling period.
Any rods displaying unexpected increases in diameter shall not be re-turned to the core.
Selected fuel rods shall be removed during each refueling period for destructive examinations. W1 an the first rods are removed for destructive examination at about 150 expected lifetime, the h ad-vanced performance bundles shall be rec.aved from the core. These asse=blies shall not be. returned to the core until the results of the destructive examinations have been evaluated and it is confirmed that the design per-formance of the fuel has been met and continued irradiation can be safely accomplished.
(d) Supplemental Core Cooling During irradiation of centermelt fuel bundles, a supplemental system for core cooling shall be provided. This system shall provide a means of introducing fire water into the reactor pressure vessel inde-pendent of the core spray system."
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Table 8.1 RESEARCH AND DEVEIDPMENT FUEL TYPE 3 Contermelt Centemelt EEI UO -PuCt General Intermediate Advanced
" Modified E-G" 2
I Geometry, Fuel Rod Array 8x8 7x7 9x9 9x9 Rod Pitch, Inch 0.807 o.921 0 707 0 707 Standard Fuel Rods per Bundle
'46 29(3) 29(1, 2, b)
O (6, 7) 52 Special Fuel Rods per Bundle 28(3) 2o et Spacers per Bundle 5
5 3
3 Fuel Rod Cladding Material Zr-2 Zr-2 Zr-2 With Various Zr-2 Initial Mechanical Properties Standard Rod Tube Wall, Inch 0.033 0.0h0 Zr-3Nb-1Sn 0.0bO Special Rod Tube Vall, Inch 0.035 0.0h0 0.Oko 0.0h0 Fuel Rods Standard Rod Diameter, Inch 0 570 0 700 0 5625 Special Rod Diameter, Inch O.570 0 700 0 5625 0 5625 Fuel Stacked Density, Parcent 94 Pellet 94 Pellet 94 PelletIS) 82 Theoretical 85 Powder 85 Powder Active Fuel Length, Inches Standard Rod 66-67 3 65-66 3 70 To Special Rod 64 9 Central, 68.6 Removable Fill Gas Helium Helium Helium Helium lodified E-G and EEI UO -Pu02 fuel bundles may contain (in the Trner regions of the bundle) four Zr-2 tube:
2 having encapsulated cotalt targets sealed within.
(2) Modified E-G and EEI UO -Pu0p fuel bundles have a special central fuel rod to which the bundle spacers are 2
fixed. In addition, two of the interior bundle fuel rods are removable and may contain UC -Puo2 fuel.
2 (3)Special rods have depleted uranium.
Also has four gadolina-containing rods.
With 31 dishing on selected rods.
fuel rod stack deasity will vary from 74 to 92% theoretical by using annular, dished, or nondished UO -PuO2 2
pellets in selected rods.
rods, four cobalt corner rods and one empty (2 rods, four removable Puo2 rods, eight gadolinia-con Sixty-four UO -Pu02 rods similar to standard UO 2
sater-filled during operation) spacer rod.
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L Table 8.2 "EEI UO -Pu0,
2 2
a and ' Modified Centermelt E-G' Fuels"
-Intermediate Advanced
- Minimum Core Burnout Ratio at Overpower 15 15 15 Transient Minimum Burnout Ratio-in Event 15 15 15 4
of Loss of Recirculation Pumps From Rated Power MaximumHeatFluxatOverpower, Btu /Hr-Ft 500,000-MaximumSteadyStateHeatFlux, Btu /Hr-Ft 410,000 500,000 500,000 Maximum Fuel Rod Power at Overpower, kW/Ft 21.6 MaximumSteadyStateFuelRodPower,kW/Ft 17 7 21.8 26.8 Stability Criterion: Maximum Measured 20 Zero-to-Peak Flux Amplitude, Percent of Average Operating Flux Maximum Steady State Power Level, MW 240 t
Maximum Value of Average Core Power h6 Density at 240 MW, kW/L i
t Maximum Reactor Pressure During Power 1,485 Operation, Psig A
MinimumRecirculationFlowRate,Lb/Hr 6 x 10 (Except During Pump Trip Tests or i
Natural Circulation Tests as Outlined-in Section 8)
MaximumMWd/TofContainedUraniumfor 23,500 an Individual Bundle i
Number of Bundles l
Pellet UO2 1
2
. Powder UO2 1
2 Rate of' Change of Reactor Power During Power Operation:
Control rod withdrawal during power operation shs11 be such that the average rate of change of reactor power is less than 50 MWt per minute.when power is
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less than 120 MW, less than 20 MWt per minute when pewer is between 120 and t
200 MW, and'10 MWt per minute when power is between 200 and 240 MW
- t t
GBased upon crit 1<al heat flux correlation, APED-5286.
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