ML20002E037

Approves Util 701215 & 17 Proposed Change 21.Change Permits Operation W/Reload F & Type J-1 Fuel Assemblies in Core. Drawings Encl

ML20002E037
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/09/1971
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
Shared Package
ML20002E038	List: ML20002E037 ML20002E042
References
NUDOCS 8101260168
Download: ML20002E037 (11)

Text

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UNITED STATES ATOMIC ENERGY COMMISSION

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j February 9, 1971 Docket No. 50-155 Consumers Power Company ATTN:

Mr. Gerald J. Walke Nuclear Fuel Management Administrator 212 West Michigan Avenue Jackson, Michigan 49201 Change No. 21 Gentlemen:

License No. DPR-6 By letters dated December 15 and 17, 1970, Consumers Power Company requested changes to the Technical Specifications of Facility License No. DPR-6 to permit reactor operation with Reload-F (Proposed Change No. 21) and " Type J-1" (Proposed Change No. 22) fuel assemblies in the core. Both types of fuel assemblies are essentially identical to the Reload E-G fuel assemblies which were approved'for the Big Rcck Point core by Change No. 16 issued on February 27, 1969. We have consolidated the requests as Proposed Change No. 21.

According to the proposed changes, the differences between Reload-F and E-G fuel are small variations in the fuel and poison distribution and repositioning of the four tic rods in the Reload-F fuel assemblies.

The difference between the " Type J-1" fuel assemblies provided by Jersey Nuclear Company and the approved E-G fuel assemblies fabricated by General Electric is the use of short -

dif fusers in the " Type J-1" fuel at the tie plate inlet to streamline the flow after the coolant leaves the flow distribution orifices.

Based on our evaluation of the changes that have been described by the Consumers Power Company, we have concluded that use of the proposed fuel assemblies in the Big Rock Point core does not present a change in the hazards considerations described or implicit in the safety analysis report and there is reasonable assurance that the health and safety of the public will not be endangered by the operation of the Big Rock Point Nuclear Reactor with Reload-F and " Type J-1" fuel bundles in the core.

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• donsumers Power Company-

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. February 9, 1971 -

l Accordingly, pursuant to Section 50.59 of JLO CFR Part 50, the. Technical Specifications of Facility License No. DPR-6~ are hereby changed as indicsted in Attachment A to this letter.. For clarification and correctness, r.inor.

revisions of'the propo.ed changes have been made as agreed by your representative.

Sincerely, sf) 'Y l

g h Vt Peter A. Morris, Director Division of Reactor Licensing -

Enclosure:

Attachment A - Changes to Technical Specifications cc: George F. Trowbridge, Esquire Shaw, Pittman,.Potts, Trowbridge & Madden 910 - 17th Street, N. W.

. Washington, D. C.

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i ATTACHMENT A CHANGE NO. 21 TO TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-6 CONSUMERS POWER COMPANY DOCKET NO. 50-155 1.

.Section 5 Figures

• A.' Delete Figure 5.2, " Original A fuel" B.

Transfer existing Figures 5.5 (8 x 8 Fuel Lattice Centermelt Bundle),

5.6 (7 x 7 Fuel Lattice Centermelt Bundle), 5.8 (Modified E-G Fuel),

and 5.9 (EEI-0U,-pug Fuel), to a new section 8 and renumber'as 7

Figures 8.1, 8.2, 8.3, and 8.4, respectively.

C.

Renumber the following Figures:

Figure 5.3, " Reload-B fuel", to Figure 5.2 Figure 5.4, " Reload-C fuel", to Figure 5.3 Figure 5.7, " Reload-E and E-G fuel", to Figure 5.4 D.

Add the attached Figure 5., " Reload-F fuel" 5

E.

Add the attached Figure 5.6, " Type

'J-l' fuel" 2.

Section 5.1.5.(c) - Change to read:

"(c)

Fuel Bundles The general design and configuration of the six types of 1

reload fuel bundles should be as shown in Figures 5.2 thrbugh 5.6 (inclusively) of the specifications. Principal design features shall be as shown in Table 5.1."

3.

Section 5.1.5 - Delete the present table and substitute the attached table, now designated Table 5.1.

4.

Sections 5.1.7, 5.1.8 and 5.1.9 are transferred to a new Section 8.

5.

Section 5.2.1. (b) - Change 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."

6.

Section 5 - Add the attached Table 5.2.

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s TABLE 5.1 RELOAD FUEL TYPES General Reload (B & C)

Reload (E)

Reload (E-G & F)

J-1 Geometry, Fuel Rod Array 11 x 11 9x9 9x9 9x9 Rod Pitch, Inch 0.577 0.707 0.707 0.707 Standard Fuel Rods per Bundle 109 74 70 67 Special Fuel Rods per Bundle 12(1) 7(le 2) 11(le 2. 3) 14(6)

Spacers per Bundle 5

3 3

3 Fuel Rod Cladding i

Material Zr-2 Zr-2 Zr-2 Zr-2 Standard Rod Tube Wall, Inch 0.034 0.040 0.040 0.040

[

Special Rod Tube Wall, Inch 0.031 0.040 0.040 0.040 Fuel Rods j

_ Standard Rod Diameter, Inch 0.449 0.5625 0.5625 0.5625 Special Rod Diameter, Inch 0.344 0.5625 0.5625 0.5625 Fuel Stacked Density, Percent 94 + 1 Pellet 90-95 Pellet (5) 94 Pellet (4, 5) 91(7)

Theore tical 85 Powdered Active Fuel Length, Inches Standard Rod 70 69.75 70 68 Special Rod 64.6 Central 64.9 Central 62.2 Central Fill Gas Helium Helium ~

Helium Helium 295%

(1) Reload B, C, E, E-G and F fuel bundles may contain (in the corner regions of the bundle) four Zr-2 tubes having encapsulated cobalt targets sealed within.

(2) Fuel bundles have a special central fuel rod to which the bundle spacers are fixed.

In addition, two of the interior bundle fuel rods are removable and may 2

2 f"*1' (3)contain UO -Pu0 In addition to special rods for Reload E, Reload E-G and F have four gadolinia-containing rods.

(4)With 30 dishing on selected rods.

(5)UO -Pu0 fuel rod stack density will vary from 74 to 92% theoretical by using 2

2 annular, dished, or nondished pellets in selected rods.

(6)This includes: 4 gadolinia-containing roc, one of which is removable, 4 removable cobalt bearing corner rods, 5 removable rods and one central spacer capture rod.

(7) Pellets are dished }% of undished volume.

2/9/71 l

l.

l TABLE 5.2 RELOAD "B"

& "C"

'E', 'E-G',

Fuel

'F'

& 'J-l'.

• Minimum Core Burnout Ratio at Overpower 1.5 1.5 Transient Minimum Burnout Ratio in Event of Loss of Recirculation Pumps from 530,0b0 500,0b0 Ma m m Hea Flux at Overpower, Btu /hr-ft2 Maximum Steady-State Heat Flux, Btu /hr-ft2 434,000 410,000 Maximum Fuel Rod Pover at overpower, kW/f t 17.2 21.6 Maximum Steady-State Fuel Rod Power, kW/f t 14.2 17.7 Stability Criterion: Maximum Measured 4

Zero-to-Peak Flux Amplitude, Percent of Average Operating Flux 20 20 Ma ;tmum Steady-State Power Level, MWt 240 260 Maximum Value of Average Core Power l

Density @ 240 MWe, kW/L 46 46 Maximum Reactor Pressure During Power Operation, psig 1485 1485 Minimum Recirculation Flow Rate, Ib/hr (Except During Pump Trip Tests or Natural Circulation Tests as Outlined in Section 8) 6 x 106 6 x 106 Maximum mwd /T of Contained Uranium for an Individual Bundle 23,500 23,500 Rate of Change of Reactor Power During Power Operation:

Control rod withdrawal during power operation shall be such that the average rate of change of reactor power is less than 50 MWt per minute when power _is less than 120 MWt, less than 20 MWt per minute when power is between 120 and 200 MWt, and 10 MWt per minute when power is between 200 and 240 MWt.

• Based on correlation given in " Design Basis for Critical Heat Flux Condition in Boiling Water Reactors'- by J. M. Healzer, J. E. Hench, E. Janssen, and S. Levy, September 1966 (APED 5286 and APED 5286, Part 2).

J 2/9/71

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Section 8 - Change to read:

"8.0 RESEARCH AND DEVELOPMENT From time to time, various developmental types of fuel will be irradiated in the Big Rock Point reactor.

This section will describe these fuel types and any operating limitations asso-ciated therewith.

8.1 Developmental 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 1

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 rods) clad with annealed special Zr-2.

The i

remaining fuel rods in the bundle (maximum of 95) shall be clad with cold-worked standard Zr-2.

8.1.2 Thin Clad Powder Fuel Bundle Two the the Reload 'C' fuel bundles 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.

f 8.2 Principal Developmental Fuel 3perating' 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.

j 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 -

2 2/9/71 i

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The fuel shall be specially designed for this oper-ation and shall be permitted to exceed the general core operating limitations of Section 5.2.1.(b) but shall be limited to the most 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 increase between 170 MWt and 240 MWt shall be limited to 1/2 MWt average per minute per notch of control rod with-drawal when any of the following conditions exist:

(1) A centermelt fuel bundle is being brought to power for the first time, or (2) a scram recovery is being made at a time in the xenon transient when the peak of the axial power distribution is lower in the core than the peak existing at the time of the last shutdown.

(c) Fuel Examinstions Nondestructive examinations of each fuel rod in the centermelt fuel bundles shall be perfora ed-during each core refueling period. Any rods displaying unexpected increases in diameter shall not be returned to the Core.

Selected fuel rods shall be removed during each refueling period for destructive examinations. When the first rods are removed for destructive examination at about 15% expscted lifetime, the 4 advanced per-formance bundles shall be removed from the core.

These essemblies shall not be returned to the core I

until the results of the destructive examinations have been evaluated and it is confirmed that the design performance 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 independent of the core spray system."

(

l Table 8.1 RESEARCH AND DEVEIDPMEFF FUEL TYPES Centermelt Centermelt EEI UO -PuOp General Intermediate Advanced

" Modified E-G" 2

Geometry, Fuel Rod Array 8x8 7x7 9x9

't x 9 Rod Pitch, Inch.

0.807 0 921 0.YO7 0 707 Special Fuel Rods per Bundle 20(3) 2O(3) 29(1 E* b)

O (6, 7)

Otandard Fuel Ibde por Bundle 36 29

$2 81 8

Spacaro per Bundle 5

5 3

3 Fuel Hai cladding Material Zr-2 Zr-2 Zr-2 With Various Zr-2 Initial Mechanical Properties Standard Pod 'ibbe Wall, Inch O.035

0. Oho Zr-3Nb-lSn

0. oho Special Rod Tube Wall, Inch 0.035 0.0h0 0.040 0.040

~

Fuel Rods Standard Rod Diameter, Inch O.570 0 700 0 5625 Fuel Stacked Density, Percent 94 Pellet 94 Pellet 9h Pellet (,'.

0 5625 Special Rod Diameter, Inch O.570 0 700 0 5625 82 Theoretical 85 Powder 85 Powder Active Fuel Length, Inches Standard Rod 66-67 3 65-66 3 To 70 Special Rod 64 9 Central, 68.6 Removable Fill Gas Helium Helium Helium Helium Modified E-G and EEI UO -PuO2 fuel bundles may contain (in the corner regions of the bundle) four Zr-2 tube 2

having encapsulated cobalt targets sealed within.

(2} Modified E.G and EEI UOp-PuO 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 UO -PuO2 I"*1*

2 (3}Special rods have depleted uranium.

b)Also has four gadolina-containing rods.

(5)With 3$ dishing on selected rods.

(6)UOp-PuOp fuel rod stnek density will vary from 74 to 92% theoretical by using annular, dished, or nondished pellets in selected rods.

Sixty-four UO *I"02 rods similar to standard UO2 rods, four removable PuO2 rods, eight gado11nia-containing 2

rods, four cobalt corner rods and one empty (water-filled during operation) spacer rod.

2/9/71

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Table 8.2 "EEI UO -Pu0,

2 2

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

-of Loss of Recirculation Pumps From Rated Power MaximumHeatFluxatOverpower, Btu /Hr-Ft 506,000 MaximumSteadyStateHeatFlux, Btu /Hr-Ft 410,000 500,000 500,000 Maximum Fuel Rod Power at Overpower, kW/Ft 21.6 Maximum Steady State Fuel Rod Power, 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 Ievel., MW 240 t

Maximum Value of Average Core Power 46 Density at 240 MW ' kW/L t

Maximum Reactor Pressure During Power 1,h85 Operation, Psig 6

MinimumRecirculationFlowRate,Lb/Hr 6 x 10 (Dccept During Pump Trip 'Ibsts or Natural Circulation Tests as Outlined in Section 8)

Maximum mwd /T of Contained Uranium for 23,500 an Individual Bundle Number of Bundles Pellet UO2 1

2 1

2 Powder UO2 Rate of Change of Reactor Power During Power Operation:

Control rod withdrawal during power operation shall be such that the average rate of change of reactor power is less than 50 MWg per minute when power is less than 120 MW, less than 20 MWg per minute when power is between 120 and t

200 MW, and 10 MWt per minute when power is between 200 and 240 MW

• t t

cBased upon critical heat flux correlation, APED-5286.

2/9/71

Distribution W. Dooly, DR R. Engelken, CO (2)

H. Shapar, OCC FEE o G71 N. Dube, DRL (5)

J. R. Buchanan, ORNL T. W. Laughlin, DTIE Document Room r0ccket File DR Reading

' Docket No. 50 155 DRL Reading ACRS (3)

Branch Reading R. DeYoung, DRL R. S. Bovd, DRL D. J. Skovholt, DRL Consur ers Power Company ATTN.

Mr. Gerald J. Walke R. H. Vollmer, DRL Nuclear Fuel llanagement D. L. Ziemann, DRL J. J. Shea, DRL Administrator 212 West Micnigan Avenue R. M. Diggs, DRL Jackson. Michigan 49201 Change No. 21 License No. DPR-6 Gentlenen.

By letters dated December 15 and 17,1970, Consumers Power Company requested changes to the Tecanical Specifications of Facility License No. DPR-6 to permit reactor operation with Reload-F (Proposed Change No. 21) and " Type J-1" (Proposed Cnange No. 22) fuel assemblies in the core. Both types of fuel asseublics are essentially identical to the Reload E-G fuel assemblics which were approved for the Big Rock Point core by Change No.16 issued on February 27, 1969. We have consolidated the requests as Proposed Change No. 21.

According to the proposed changes, the dif ferences between Reload-F and E-C fuel are small variations in the fuel and poison distribution and repositioning of the four tic rods in the Reload-F fuel assemblies. The difference between the '! Type J-1" fuel assemblies provided by Jersey Nuclear Company and the approved E-G fuel assemblies f abricated by General Electric is the use of shoi-t dif fusers in the " Type J-1" fuel at the tie plate inlet to streamline the flow after the coolant leaves the flow distribution orifices.

Based on our evaluation of the changes that have been described by the Consumers Power Company, we have concluded that use of the proposed fuel assemblies in the Big Rock Point core does not present a change in the hazards considerations described or implicit in the safety analysis report and there is reasonable i

assurance that the beklth and safety of the public will aqt be endangered by the operation of the Big Rock Point Nuclear Reactor with Reload-F and " Type J-1" fuel bundles in the core.

I E/d/96b/4f

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. FEB 9 E7I Accordingly, pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifichtions of Facility License No. DPR-6 are hereby changed as indicated in Attachment A to tais letter.

For clarification and correctness, minor revisions of tne proposed changes have been made as agreed by your representative.

Sincerely, Peter A. Morris, Director Division of Reactor Licensing

Enclosure:

Attaciment A - Changes to Technical Specifications George F. Trowbridge, Esquire cc; Shaw, Pittu:an, Potta, Trowbridge & !!adden 910 - 17th Street, N. W.

Washington, D. C.

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