ML20002C419

Application for Amend 4 to License DPR-6

ML20002C419
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/24/1972
From:	Lamley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20002C415	List: ML20002C414 ML20002C419
References
NUDOCS 8101100278
Download: ML20002C419 (21)

Text

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CONSUMERS POWER COMPANY Docket No 50-155 i

Application for Amendment No 4 to Operating License DPR-6 and Request for Change to' the Technical Specifications Change No 3h License DPR-6 For the reasons hereinafter set forth, the following changes to the Technical 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 are requested:

I.

Section 8 A.

Change the first paragrriph of Section 8.1 to read as follows:

"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 8.8.

Principal design features shall be essentially as on Table 8.1."

B.

Section 8 - Figures:

Add Figure 8.8 - Euclear Fuel Services Inc, Demonstration Assembly.

C.

Delete present Table 8.1 and insert the attached Table 8.1.

D.

Delete present Table 8.2 and insert the attached Table 8.2.

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TABLE 8.1 Research and Development Fuel Types New New gy Centermelt Centemelt NFS Centermelt Centermelt g,g General Intemediate Advanced

'kodified E-C" 2

2 Intermediate Advanced DA ceametry, Fuel Rod Array 8x8 7x7 9x9 9x9; 8x8 7x7-11 x 11 '

6

-Rod Pitch, Inch 0. 8 (77 0.921-0.707 0 707 0.807 0 921 0.577-M 1

Standard Fuel Rods per Buntle 36 29 52 0

60 N5 h0 I9)

III'

'4(1) 81 0' U k

28(3) po(3) 29(1, 2, M 81 Special Fuel Rods per Bundle Spacers per Bundle 5

5 3

3 5

5-5 Final Rod Cladding s

0)

Material Zr-2 Zr-2 Zr-2 With various Zr-2 Zr-2 Zr-2 '

Zr-2 Initial Mechanical Properties Zr-3Nb-lSn Standard Rod Tube '.'sil, Inch 0.035 0.0ho 0.040 0.035 0.0h0 0.034 r

• IUI Special Rod Tube Wall, Inch 0.035 0.040 0.Cho 0.040 0.031 0.031 Fuel Rods Standard Rod Diameter, Inch 0 570 0.700 0 5625 0.570 0.700 0.W9 Special Rod Diameter. Inch 0.57 0 0.700

0. % 25 0 5625 0.347(8) 0 347(8) 0.u9(H) '

Fuel Stacked Density, Percent 94 Pellet 94 Penet 94 Pr. net ($}

82 Powder 92-93 Penet '

92-93 Penet' 91.5%

i Theoretical 85 Powder 85 Powier Active Fuel length, Inches Standard Rod 66-67 3 65 66.3 70 70 67.3 66.3 70 Special Rod 64.9 Central.

68.6 Removable Fill Gas Be1.ium Helita Helium Helium Helium Helium Helita l.

w See attached page for footnotes.

j (RevisedwithChangeNo34 issued 7/24/72.)

FOOTNOTES 'IO TABLE 8.1 (1} Modified E-G and EEI UO -PuO and new centermelt fuel bundles may contain (in the corner regions of the bundle)~

2 2

four Zr-2 tubes having encapsulated cobalt targets sealed within.

(

Modified E-G and EEI UO -PuO fuel bundles have a special central fuel rod to which the bundle spacers are fixed.

2 2

In addition, two of the interior bundle fuel rod 1 5 r

Transient Minimum Burnout Ratio in Event 1.5 1.5 15

>1.5 of Loss of Recirculation Pumps From Rated Power Maximum Heat Flux at Overpower, Btu /h-Ft 500,000 402,000 Maximum Steady State Heat Flux, Btu /h-Ft 410,000 500,000 500,000 329,000 Maximum Fuel Rod Power at Overpower, kW/Ft 21.6 13.8 Maximum Steady State Fuel Rod Power, kW/Ft 17 7 21.8 26.8 11.3 Stability Criterion: Maximum Measured 20 20 Zero-to-Peak Flux A=plitude, Percent of Aversge Operating Flux Maximum Steady State Power Level, W 240 2h0 t

Maximum Value of Average Core Power h6 46 Density at 240 W, kW/L Maximum Reactor Pressure During Power 1,485 1,h85 Operation, Psig I

Minimum Recirculation Flow Rate, Lb/h 6 x 10 6 x 10 (Except During Pump Trip Tests or N:.tural Circulation Tests as Outlined

)

in Section 8)

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

3 2

Powder UO 1

2 2

Rate of Change of Reactor Power During Power Operation:

Control rod withdrawal during power operation shall be such that the average rc.te of change of reactor power is less than 50 Wt Per minute when power is less than 120 W, less than 20 Wt per minute when power is between 120 and t

200 W, and 10 W per minute when power. is between 200 and 240 W.

t cBassd upon critical heat flux correlation, APED-5286.

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l II.- Discussion - NFS D,emonstration Assemblies A.

Fuel Description d '

The NFS-DA fuel has been designed within the concept of direct interchangeability with incumbent fuel st that no change in operations, instructions or limits vill'be required.

Most of the features of the NFS-DA fuel have been employed previously in Big Rock Point fuel assemblies. Table 1 (attached) compares the principal characteristics with Types B&C, EEI, F and J-2 fuel assem-blies. Like the Type B&C fuel, the NFS-DA employs an 11 x 21 array of rods, utilizes a fuel rod outer diameter of 0.449 inch and a cladding thickness of 0.034 inch. It has cobalt target rods in the four corners i

of the assembly, is designed with sufficient reactivity to achieve an aver-rage burnup of 15,000 mwd /MTM, and does not use gadolinia burnable poison.

Like the EEI and J-2 fuel, it uses UO and mixed oxide (Pu0 -UO ) 81"-

2 2

2 tered pellets of approximately the same density, with the nixed oxide rods arranged in the inner regions of the assembly and the UO fuel r ds on the 2

' outer rows. Nonfueled water rods, similar to that of the EEI fuel, are positioned within the assembly.

The principal differences between the NFS-DA fuel an3 the previous

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fuel are in mechanical construction as described below.

1.

Mechanical Design The NFS-DA fuel fits the same physical envelope as other big Rock Point fuel assemblies and is fully interchangeable with them.

However, the NFS-DA utilizes a mechanical design that is unique (see Figure 8.8).

The tubes in the corner positions serve as the structural members to tie the assembly together, as the spacer grid capture rods, and as the cobalt target rod housings.

The fuel rods are positioned by five spacer grids. Previous Big Rock Point fuel designs accomplished this by providing three intermediate spacer grids and employing fuel rods with end plug shanks that fit into 1

holes in the upper and lower end fittings.

The lower end plugs of the NFS-DA fuel rods rest on the lower end fitting and the upper end plugs stop short of the upper end fitting. Each rod is free to expand axially, independent of othar rods ip the assembly.

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The lower end fitting is similar in appearance to those of earlier Pig Rock Point fuel designs (ie, Type F), except that it has its coolant flow openings graded in size from the center outward to redis-tribute the jet formed by the existing channel support tube orifice located beneath the fuel and is not drilled to position fuel rod end shanks. The coolant flow openings in this lower end fitting were designed to be compat-ible with the existing support tube orifices as well as with the new ori-fices which arc being used to replace a portion of the existing orifices.

The upper end fitting is completely removabic so that any fuel rod may be removed from the assembly and examined or replaced, if desired.

Four nonfueled tubes are positioned in the inner regions of the fuel assembly. They are made of the same material, diameter, wall thick-ness and overall length as the fuel rod cladding. The upper end of these tubes is fully open but the lower end has a special end plug with side openings to aisit coolant to the inside of the tube. These end plugs rest on the lower end fitting like those of the fuel rods.

i 2.

Nuclear Design

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The NFS-DA fuel is designed to achieve an average burnup of 15,000 mwd /MIM during three (3) cycles in the reactor.

The assembly average enrich-j ment is 3.31 w/o ficsile (Pu-239, Pu-2hl and U-235), distributed as shown in Figure 1 (attached) and in accordance with the detailed mass balance given i

in Table 2 (attached).

Beginning-of-life (BOL), normalized rod power dis-tributions within a fuel assembly are shown in Figures 2 through h (attached)

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for uncontrolled, hot operating conditions at 0%, 25% and 40% void planes.

q The lack of rod power symmetry in the assemblies is due to nonuniform burnups assumed for three surrounding assemblies in a two-by-two unit j

assembly group. This arrangement of fuel, with varying degrees of deple-tion, around a fresh fuel assembly, was employed for the rod power factor evaluctions rather than one that assumes the assembly to be inserted in an I

infinite array of identical fresh fuel because the former represents a re-alistic evaluation of the environment these bundles will experience.

A comparison of rod power peaking factors for the NFS-DA fuel with Types F, EEI and J-2 fuel is included in Table h (attached).

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The infinite multiplication factor, the reactivity deficit from cold to hot standby, the reactivity responses and the control rod worth for the NFS-DA fuel' are compared 'to those of previous fuels in Tcble 3-(attached).

l The use of mixed oxide in the NFS-DA produces a slightly reduced l

' control rod worth (cold). However, with the presence of four DA and the lower k. (cold) of the fuel, shutdown and hold-down capability of the core i

with one control rod adjacent to an NFS-DA fuel assembly fully withdrawn is assured.

Lattice characteristics of the NFS-DA fuel produce a large cold-to-hot reactivity sving. The result is a more negative moderator temper-ature coefficient than that of the Type J-2 fuel.

Similarly, the void response of the NFS-DA fuel is more negative than that of the Type J-2 fuel but somewhat less negative than that of the Type F fuel. Finally, the increased resonance absorption of the Pu-240 creates a Doppler re-sponse for the NFS-DA fuel that is more negative than any other fuel shown in Table 3 (attached). The combination of the moderator temperature, the void, and the Doppler characteristics produces a power response to re-activity insertion events equivalent to or less severe than that of pre-viously licensed fuel.

I 3.

Thermal'and Hydraulic Design The thermc1 performance design of the NFS-DA is such that they may be considered operationally interchangeable with the present principal resident fuel (Type F). The hydraulic characteristics of the upper and j

lower end fittings and the spacer grids have been carefully adjusted to produce an overall assembly hydraulic characteristic similar to that of

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the Type F fuel. The putpose of this tailoring of the hydraulic charac-teristics is to permit the assemblies to pass the same flow as that of Type F fuel. This assures that they will not detract from the hydraulic performance of the Type F or other resident fuel. By matching flow of the NFS-DA, the critical heat flux characteristic is affected mainly by the differences of heat transfer surface area and P eal peaking factors.

a Table h (attached) compares the principal thermal-hydraulic character-istics of the Types B, E, E-G, F, EEI, J-2 and NFS-DA fuel (

Individual e

i.

9 NFS-DA have 17% more heat transfer surface area than a Type F fuel assem-bly and slightly more than this when compared to the other mixed oxide (UO -Pu0 ) fuel assemblies in the core.

2 2

The local peaking factor depends upon the interactien of the effects of fissile material distribution in the fuel rods, the assembly internal water to fuel ratioc, the void fraction within the assembly and the exposure. Figures 2, 3 and h (attached) show the local power distri-butions at 0%, 25% and 40% void planes at beginning of life. These are comparable to or slightly below what would be expected from previously licensed fuel in the BRP reactor. Accordingly, with comparable fuel assembly pin power peaking factors, the sw=e coolant flow rate,175 more heat transfer surface area, and an equivalent or less severe power re-1 sponse to renetivity insertion events, the critical heat flux margins J

in the NFS-DA fuel will be greater than that of 9 x 9 array fuel under both steady state and transient operation.

As stated earlier, the NFS-DA fuel is designed to be operationally interchangenble with the principal resident fuel (Type F) at the time of its insertien into the reactor. Accordingly, it will be operated at the

)

i same assembly power limits applied to the existing fuels. In this mode the NFS-DA linear and surface flux limits are less than those of pre-viously licensed 9 x 9 array fuel.

If an NFS-DA was operated at that j

power level and power distribution that would produce a aurface heat flux in a Type F fuel assembly of 500,000 Btu /h-ft and a linear heat flux of 21.6 kW/ft, the corresponding values in the NFS-DA fuel would be approxi-mately h02,000 Btu /h-ft and 13.8 kW/ft. At this condition the NFS-DA fuel central temperature would only be 3650 F.

The thermal conductivity characteristics of UO were used in these calculations. Because the con-2 tent of Pu0 in the mixed oxide rods is low, the thermal conductivity of 2

the fuel material will be essentially that of UO.

Furthermore, the 2

greater neutron self-shielding of the Pu0 will ause a greater fraction 2

of the power to be generated away from the pellet center line, with the result that temperature calculations using UO characteristics will indi-2 cate higher than actual values.

l

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t 10 h.

Developmental Of the 113 fuel rods in each assembly,12 rods are test or developmental rods. Theca iods are identical to the standard fuel rods in all respects except that their cladding material will be Zr-k instead of Zr-2 and 6 of them vill not be prefilmed (autoclaved). The use of Zr-k cladding material in 12 rods and the nonprefilming of 6 rods in each assembly is not expected to interfere with the performance of the assemblies in any way.

5 Pu02 Particle Size Fanufacturing specifications require that 95% of the Pu0 2 particles in the mixed oxide pellets be less than 100 microns in diameter at a 95% confidence level. At this level, 95% of all particles measured by the alpha-autoradiographic technique must be 60 microns or less. Man-ufacturing procedures will assure that only a very few particles will exceed 100 microns in diameter.

The significance of particle size is amply discusaed in the

" Additional Information for Proposed Technical Specification Change No 19,"

dated January 28, 1970. It'is shown that the presence of 460 micron plutonium particles does not adversely affect the negative Doppler reac-tivity coefficient because of prompt energy transfer from the particles

)

to the surrounding UO matrix during a transient. Also, Change No 19 2

i shows that if 15% of the plutonium is in particles greater than 50 microns, the fraction of energy that is available for conversion to mechanical energy for isolated particles is less than 0.3% of the total energy avail-able, and is therefore rather insignificant.

Accordingly, the discussion of the significance of particle size presented in Change No 19, including results and conclusions stated therein, applies to the NFS-DA fuel.

B.

Accident Analysis 1.

Reactivity Insertion Accident j

The reactivity response data of Table 3 (attached) indicate that the characteristics of the NFS-DA fuel are all acceptable. The table reveals that the Doppler coefficient is more negative than either the J-2 l

l or the EEI-Pu fuel and that the void response is also more 6egative than l

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11 that of either the J-2 or the EEI-Pu fuel and that the void response is also more negative than that of either the J-2 or EEI-Pu fuel. The de-layed neutron fraction is comparable to that of the J-2 fuel and greater than that of the EEI-Pu fuel. The rod worth is less. Accordingly, the consequences of a rod drop accident would be expected to ?? less for the NFS-DA fuel than for the J-2 fuel. Therefore, the reactivity insertion accident discussion presented in the application for Change No 19, dated December 27, 1969 and additional information dated January 20, 1970 and February 13, 1970 vill conservatively apply to t'le NFS-DA fuel.

2.

Loss-of Coolant Accident i

Analysis of the Design Basis Accident for the NFS Demonstration Assemblies has been performed in accordance with the Interim Acceptance Criteria. The results of this analysis were submitted to the Directorate of Licensing by letter dated May 18, 1972. The analysis yields a maximum clad temperature of 2129 F and a core average metal-to-water reaction of 0.3%.

The reanalysis was performed assuming that the NFS-DA fuel was operating at a bundle power equivalent to tL3 licensed limit of Type F fuel. The limits proposed for Section 8.2 of the specifications are con-sistent with values used in the reanalysis.

Reanalysis of small bresk losc-of-coolant accidents are in progress. The results of this reanalysic will be submitted to the Directorate of Licensing when they are completed.

3 Primary System Integrity The reactivity insertion associated with a rod drop accident will be less than that for the J-2 fuel, ris discussed earlier. There fore,

1 the discussion on Primary System Integrity presented in the application for Change No 27 vill conservatively apply to the NFS-DA fuel.

14 Fuel Handling and Criticality The NFS-DA fuel vill be shipped to Big Rock Point in a fresh fuel shipping container, licensed separately for mixed oxide fuel ship-ment.

Inside the Big Rock Point containment vessel they will be handled in the same fashion as previous fuel.

I

i 12 The NFS-DA fuel exhibits reactivity characteristics which are similar to other BRP fuel. Therefore, they may be handled and stored under the criticality and control limits which apply to BRP fuels.

III. Conclusions Based on the data and discussion presented above, 't is concluded that:

1.

The NFS-DA fuel may be operated interchangeably with present resident fuel. All present limits and operating philosophy may be applied.

2.

The fuel rod power peaking factors of the NFS-DA fuel are compar-able to or lower than those of previously licensed fuels. In addition, the relative or radial power of these assemblies will be lower _ then those of the Types E, E-G, F and EEI-Pu fuels due to lower k..

3.

The greater heat trahsfer surface area and the same overall hydraulic characteristics as the Types E, E-G and F fuels, together with co= parable or lower power peaking factors, assure a greater critical heat flux ratio for equivalent assembly power perfomance under both steady state and transient conditions.

4.

Peak fuel temperatures of the NFS-DA fuel vill be considerably lower than those of previously licensed 9 x 9 array fuel due primarily to the lower linear heat flux that occurs as a result of an 11 x 11 array design.

5 The consequences of a rod drop reactivity insertion a6 9.h will be less severe than for the J-2 fuel because of lower rod worth.

6.

The NFS-DA fuel meets the Interim Acceptance Criteria for a Design Basis Accident.

t

I 13 Based on the above considerationn, we have concluded that the use of NFS-DA fuel 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.

-CONSUMERS POWER COMPAIN

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By

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Vice President]

Date: July 24, 1972 j

Sworn and subscribed to before me this 24th day of July 1972.

G nu u

G Notary Public,.Jj[ckson County, Michigan My con: mission e2pires October 14, 1974.

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i TABLE 1

^

B&C_

EEI

_F_

J-2 NFS-DA Gen ral Geometry, Fuel Rod Array 11x11 9x9 9x9 9x9 llxil R:d Pitch, Inch 0.577 0.707 0.707 0.707 0.5 7 40 11)

Standard Fuel Rods per Bundle 109) 81{2) 70(1,3,4) 4'(6) 1241 jj 32 81 8,11)

Special Fuel Rc.ds per Bundle Spacers per Bundle 5

3 3

3 3

Fuel Rod Cladding I9)

Material Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 0.040 (7) 0.034(10)

Standard Rod Tube Wall, Inch 0.034 Special Rod Tube Wall, Inch 0.031 0.040 0.040 (7) 0.034 Fusi Rods 0.5625 0.5625 0.449(10)

Standard Rod Diameter, Inch 0.449 0.449 0.56?g)

Sp;cial Rod Diameter, Inch 0.344 0.5625 0.5625 90.71 91.5 Fu21 Stacked Density, % Theoretical 94 + 1 82 94 pel-pelTet 85 powdered Active Fuel Length, Inches 70 68 70 Standard Rod 70 70 64.9 62.2 70 Special Rod Central Central Fill Gas Helium Helium Helium Helium Helium

> 95%

> 90%

(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) 64 UO -Pu02 rods similar to standard U02 rods, 4 removable Pu02 rods, 8 Gadolinia 2

containing rods 4 cobalt corner rods and 1 empty (water filled during operation) spacer rod.

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

(4 In addition to special rods there are 4 gadolinia containing rods.

(5 With 3% dishing on selected rods.

This includes 24 mixed oxide (Pu0 -UO ) rods, 4 cobalt bearing corner rods and (6

2 2

4 gadolinia bearing rods.

(7).This includes 44 fuel rods of 0.050", 33 fuel rods of 0.040"; and 4 cobalt rods of 0.035".

(8) The 81 special rods are comprised of 4 tobalt bearing rods,12 experimental rods, 4 empty rods (water filled during operation), with the remainder U0 -Pu02 rods 2

similar to the standard U02 rods.

(9) The 12 experimental rods and cobalt bearing have Zr-4 cladding, the empty rods are of Zr-2.

(10)The corner tubes for loading the cobalt target rods are 0.537" OD and have a tube wall 0.081".

(ll)Some of the 12' experimental rods may replace standard fuel rods, f

TABLE 2 tiFS-DA FUEL MASS BALAt;CE (AT BOL)

Inner Region

. Intermediate Region Outer Region Assembly Average

• 2.45 Pu-2.4 U l.03 Pu-l.56 U 2.4 U l.18 Pu-2.16 U (grams / rod)

(grams / rod)

(grams / rod)

(Kgs/Assy)

~

1.

U-235 25.767 17.043 26.545 2.663 2.

U-236 0.000 1.747 0.000 0.056 3.

Uranium Total 1073.679 1092.508 1106.098 123.225 4

Pu-233 0.134 0.078 0.000 0.010 S.

Pu-239 24.864 10.501 0.000 1.355 6.

Pu-240 6.062 2.560 0.000 0.330 7.

Pu-241 2.251 0.951 0.000 0.123 3.

Pu-242 0.554 0.234 0.000 0.031 9.

Fissile Plutonium (5 + 7) 27.115 11.452 O.000 1.478 10.

Plutonium Total 33.915 14.324

• 0.000 l.849 11.

Total Fissile (1 + 9) 52.882 28.495 26.545

4. l f.1 12.

(U + Pu) Total 1107.594 1106.832 1106.093 125.074' 13.

Average w/o Fissile 4.77 2.57 2.40 3.31 T

C Assembly average data is based upon the following combinations:

g 41 rods of 2.45 Pu - 2.4 U (Fresh)

Inner Region 32 rods of 1.03 ?u - 1.56 U (Reprocessed)

Q Intermediate Region 40 rods of 2.4 U (Fresh)

N Outer Region M

mm r-

TABLE 3 E

E-G F

EEI-Pu J-2 Frs-DA Fuel loading No. of Gd 023 Rods 4

4 8

4 Kg of Uranium D6.4 G8.4 138.2 117.8 123.3 123.2

'c..

Kg of Plutonium 5.33 1.47 1.85 Kg of U + Pu 136.4 138.4 138.2 123.1 124.8 125.1 Total Fissile Content (w/o fiss!1e) 2.979 3.55 3.233 4.84 3.522 3.3' Relative Fissile Content in Pu 88.9 82.06 79.s5 Reactivity (Uncontrolled L)

Cold (68'F) 1.268 1.208 1.179 1.160 1.148 1.199 Hot Standby NA NA NA NA 1.144 1.184 Full Power, 01 Void 1.280 1.203 1.183 1.168 1.138 1.176 Full Power, 25% Void 1.262 1.183 1.171 1.158 1.118 1.144 Temperature Deficit 68* - 77*F (a b/ L per 'F)

NA NA NA NA NA

+0.32x104 Start of Cycle (aKeff/K ffper'F)

+0.38x1C-4

+0.27x10-4

+0.31x10-4

+0.30x10-4 NA NA Cold - Hot Standby (ab,)

NA NA NA NA

-0.0036

-0.0146 Vaid Response 0 - 251 Vold (a L)

NA NA

-0.012*

-0.010*

-0.0202

-0.0325 Full Power, 25% Vold (at/K per unit void)

-0.11

-0.12

-0.133

-0.084

-C.;;710*

-0.1107 Doppler Response Hot Standby (aL/L per "F)

NA NA NA NA NA

-1,33x10-5 Full Power, 25% Void (AK, f

-1.2x10-5

-1.2x'.0-5

-1.2x10-5

-1.25x10 5

-0.90x10-5

-1.48x10-5 5

Hot Standby - Full Power fa/Keff per 'F) gg gg b/ L per 'F)

NA NA NA

-1.05x10-C'ntrol Rod Worth Cold (aL,oneadjacentrod)

NA NA NA NA 0.7126 0.1048 021ay Neutron fraction M

NA NA 0.00385 0.0057

~0.0050

• Infered Values NA = Not Available

THE h B

E E-G F

EET-Pn J-2 NFS-DA Number of Fuel Rods per Bundle 117 77 77 77 76 77 113 Fu21 Rod Diameter qin) 0.449 0.5625 0.5625 0.5625 0.5625 0.5625 0.449 Active Fuel Length Lin) 70 69.751 702 702 70 688 70 Heat Transfer Area we Bundle (sq.ft) 7P.94 65.84 66.08 66.08 65.29 64.18 77.48 Coolant Flow Area pe Bundle tsq.in) 44.23 22.47 22.47 22.47 22.47 22.47 23.17 Average Surface Heat flux" (Stu/hr-sq. ft) 119,830 144.000 143.140 143.140 144.870 151.915 122.080 Maximum Surface Heat Flux 5 5 LBtu/hr-sq.,ft) 530.000 500.000 500.000 500.000 500.000 500.000 500.000 Average Linear Heat Flux *

(Kw/ft) 4.06 6.26 6.18 6.18 6.25 6.55 4.20 Maximum Linear Heat Flux 5 8 IKw/ft) 17.2 21.6 21.6 21.6 21.6 21.6 21.6 Maximune Fuel Temperature' d*F)

NA NA 5040 5203 4606 5060 4600 Minimum Critical Heat Flux Ratio'

> 1.5

> 1.5

> 1.5

> 1.5

> 1.5

> 1.5

> 1.5 7

1.53 1.21 1.3

" 1.3 1.287' 1.121' 1.163' Pin Power Factor of Highest Power Rod 1 ne spacer capture rod of 64.6 inches active fuel langth 0

20ne spacer capture rod of 64.9 inches active fuel length 3 ne spacer capture rod of 62.2 inches active fuel length 0

"At rated reactor power i

5 icense limit L

8At 122% of rated reactor power Eeginning4f life, uncontrolled At 25 Void

FIGURE 1 DISTRIBUTION OF FUEL R0D TYPES NFS MIXED OXIDE (Pu0 -UO ) DEMONSTRATION ASSEMBLY 2

2 FOR BIG ROCK POINT

~

s I

Water Gap i

)

l l

L Co l

j

)

}

l 4.

W; 3

Ol

(

M t

L i

L M

M M

{

i e

s r

L i

M i

M l

H I

H H

I s

G ln n

m

'n a

l H

P

\\

Water L

M H

H I

g Hole

'O"J

~

O O

O O

O O C Q G G L l

FUEL ROD TYPE NO. OF RODS w/o FISSILE j

L-2.40 w/o U-235 (Fresh) 40 2.40 M-1.03 w/o Puf + 1.56 w/o U-235 32 2.57 H-2.45 w/o Puf + 2.40 w/o U-235 41 4.77 (Fresh) 113 3.31 (Average)

I P.00R.0Rll;INAL

(

FIGURE 2 LOCAL POWER DISTRIBUTION * (NORM T0 113 RODS) l NFS MIXED OXIDE (P 0 -UO ) DEMONSTRATION ASSEMBLY u2 2

FOR BIG ROCK POINT W

ter Gap L

L L

L L

i..Co l

l 1.098 1.01 6 0.965 j

0.942 1

0.936 i

s L

L M

M

' *M M

t 1.087 0.861 1.106 0.986 0.941 0.930 l

e s

h

[

M M

H H

r i 0.997 l

1.096 j

0.829 1.171 I

i 1.086 1.070 I

s G

L M

H H

H H

0.942 0.969 1.162 0.975 0.980 0.952 1

s Water L

H H

g l

N I *'

I 0.91 6 I 0.921 1.073 0.976 l.046 s

N L

M H

H H

H 0.908 0.910 1.056 0.948 --

1.045

- 1.006 N

I

\\

9 HOT, OPERATING CONDITION, BOL (

0% VOID - NO CONTROL) 0.0 ASSEMBLY AVERAGE BURNUP

=

i I

4 200R OR GINAL

3, FIGURE 3 LOCAL POWER DISTRIBUTION * (NORM T0 113 RODS)

NFS MIXED OXIDE (P 0 -UO ) DEMONSTRATION ASSEMBLY u2 2

FOR BIG ROCK POINT I

Water Gap h

L L

L

.Co l

l 1.136 l

1.038 0.

~

l 0.946 1

0.938 I

s a

~\\

C1.125 L

M M

M M

L l

0.884 l

1.128 I

0.995 0.943 l

1 0.930 I

lt v

e s

H H

r M

M H

l 069 1.0 I

1.118 1

0.843 1.163 i

~

G fm n

m L

M T H

H H

T H

1.155}

a 0.954 0.981 l

0.954 0.943 1

0.91 5 i

P

\\

M H

H Water H

I I 0.921 l

0.924 1.058 0.939 Hole 0.987 x

s L

M H

H H

H

~~~

0.912 0.911 1.038 0.91 2 0.986 0.950 N

I HOT, OPERATING CONDITION (25% VOID - NO CONTROL)

ASSEMBLY AVERAGE BURNUP 0.0

=

P00R ORIGINAL

4 I

FIGUREJL LOCAL POWER DISTRIBUTION * (NORM T0 113 RODS)

NFS MIXED OXIDE (P 0 -UO ) DEMO.'ISTRATION ASSEMBLI u2 2

FOR BIG ROCK POINT rumns I

Water Gap W

L L

I l

1.054 I

I l

l O.94 L

L M

h M

M e

r 1.153 1 0.899 l

1.141 l

1.003 I

0.944 l

I 0.930 I

i s

M M

H H

H G

l 1.036 1.132 l

l 0.853 1.158 l

1.058 1.036 l

l j

a

(

\\

l L

l l

M l

H H

H l

H l

1.149 0.942 0.920 0.8 M

H H

Water H

l 0.926 1

0.926 i

i 1.047 0.91 6 I

Hole 0.948 I

s L

M H

H H

H 0.915 0.911 1.025 0.888 0.947 0.913 ----

/

N I

i 0 HOT, OPERATING CONDITION, BOL ( 40% VOID - NO CONTROL)

ASSEMBLY AVERAGE BURNUP 0.0

=

I P00R ORIGINAL

-