ML20003A052

Safety Evaluation Supporting Proposed Change 24 to Tech Specs Re Insertion of Two Centermelt Fuel Bundles in Reactor During Feb 1971 Refueling Outage

ML20003A052
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	03/03/1971
From:	Ziemann D US ATOMIC ENERGY COMMISSION (AEC)
To:
References
NUDOCS 8101290347
Download: ML20003A052 (12)

Text

.

ATOMIC ENERGh COMMISSION N

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WA2 HI N CTC N. D.C.

20545 H;rch 3, 1971 4mo e

AA/t Files (Dfck't No.54-155) W M THRU:

D. L. Zic6snn, Chief, ORB #2, DRL EVALUATION OF TWO NEW CENTER'ELT FUEL ASSEMBLIES FOR BIC ROCK POINT NUCLEAR REACTOR (CONSUMERS POWER COMPAW)

INTRODUCTION Consumers Power Company, by letter dated Decenber 21, 1970, reouested cpproval of Proposed Change No. 24 to insert two new centernelt fuel r bruary 1971 bundles into the Big Rock Point Nuclear Reactor during the e

refuelinF outage. The Consumers Power Cemeanv and the General Electric Company Nuclear Energy Division, have jointiv undertaken the c'ontinuation of the centermelt irradiation testing program at Big Rock Point, a procram that was originally sponsored by EURATOM and the U. S. AEC, and terninated on June 30, 1969. Of the six original centermelt fuel assemblies, five were removed after the first 3-month cycle of irradiation pending destruc-gj tive evaluation of selected fuel rods. The sixth centermelt subassembly (D-50) was removed in May 1969 when suspected multiple fuel rod failures were confirmed by visual inspectic? of the bundle. The cause of premature rod failures was tentatively attributed to accelerated corrosion due to clad overheating as a result of excessive crud deposition, predominantly a copper oxide.

A supplemental report (6), based on hot cell examination of two fuel rods from the Intermediate Performance Centermelt Assembly that was irradiated for about one year achieving high power rod average exp.osures of 10,000 mwd /TU, indicates that the cause of severe clad deterio-ratien was accelerated corrosien on the cutside surface of the clad driven by local overheating of the clad. Grain growth in the zircaloy structure adj acent to the deterioration indicated temperatures of 1200-1300*F.

Failure was attributed to excessive crud depositien and high surface heat fluxcs.

DESCRIPTION The two new centermelt fuel assemblies (D-57, an intermediate performance fuel assembly consisting of an 8 x 8 array of 0.570 inch 0.D. fuel rods, 16 of which are hot enough at rated power conditions to have incipient centerline fuel melting; D-56, an advanced performance fuel asse=bly con-sisting of a 7 x 7 array of 0.700 inch 0.D. fuel rods, 16 of which have 2 melting at rated power conditions) dif fuclassembliesthatwereapprovedbyDRL{yr definite but moderate center UO C

from the original centermelt and inserted into the Big Rock Point reactor in March 1968 in that:

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2-

'M rch 3, 1971 Fil;s n-1.

stainless steel tubes with the same dimensions as fuel rod cla'dding are'used:

as corner structural posts for tying the upper and lower frames a.

together, b.-

to hold the five fuel rod spacers in each assembly at the proper elevation, and c.

to hold dummy rods containing cob' alt targets.

2.

there are only 16 high power rods in each bundle in contrast to 36 in the original 8 x 8 intermediate performance centermelt assemblies or 29 in the original 7 x 7 advanced performance centermelt assemblies.

k'e have prepared the following table from information provided by Consumers Power Company to identify the most important performance characteristicsofthenewcentermeltfug})assembliesincontrastto fuel assemblies we have already approved for the Big Rock Point core, and in this manner provide the basis for our evaluation.

,~

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For example - line 8 shows that the-weight of UO2 for each bundle type q,.

has decreased more than can be accounted for by the PE emission of fuel in the corner positions, i.e.,

7.2%

7d for the 8 x 8 instead of 6% and 10.5% for the 7 x 7

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instead of 8%.

[.;.

line 9_ shows a 16% decrease in U235 per bundle.

p line 1_0 shows 9.5% decrease in fuel enrichment for the 8 x 8 fuel assembly; 7.5% decrease for the 7 x 7 fuel assembly.

lines 7,__11 and 13 snow that 96% of the power was generated in 56% of the fuel (36 rods) in the original 8 x 8 in contrast to 44.5% of the pocer in the new E x 8 which is generated in 26.7% of the fuel (16 rods).

Similarly for the 8 x 8, 95.7% of the power was generated in 59.0% of the fuel (29 rods) compared with 55.60%

power generated in 35.60% of the fuel (16 rods).

_line 14 shows that the percentage of bundle power generated in the hottest rod of each bundle is about the same or up slightly.

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v line 18 shows that the ratio of pow 2r in cdjacent rods is noticeably lower for the new fuel in contrast to the original centermelt assemblies and line 22 shows that there a*a fewer low power rods adja-cent to the highest power generation rod.

line 21 shows that power generation in the original hot rods was slightly higher than anticipated -nd line 20 shows the ratio of old to new fuel rod peaking factors line 1_2_ shevs that the Technical Specifications MCHrR 1.5 is satisfied at the 122% steady-state p,gwer Icvel.

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On

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1 1

1

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Comparinon of New and Original Centermelt Fuel i

1 2

3 4

7 x 7 Advanced 8 x 8 Intermediate Performance Centermelt Performance-Centermelt-Original New Original New

_ Ref. 1)

Prop. Ch. 24-(

(Ref. 1)

Prop. Ch. 24

1) Rod Dic.acter inches 0.570 0.570 0.700 0.700

2) Fuel Rods / Bundle 64 60 49 45

3) Cobalt Targets in 0

4 0

4 corner positions

4) Number Depleted i.s..

rods per bundle (low' 20 0

28 0

power)

5) Number Natural UO2 rods per bundle (low power) 0 28 0

12

6) NumSt:c Intermediate

\\

Pcrer (2% U235) I d8 0

16 0

17 per bundle

7) Number High Power rods per bundle 36 16 29 16

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i 1

2 3

4 B) Weir,ht UO / Bundle 140 167.

kg 136 cale 126 cale 161 cale 144 calc 2

9) Weight U 0 / Bundle-kg 3.89 calc 3.26 calc 4.75 calc

'3.98 calc.

235 2

10) Avarage bundle 3.89 or 2.86 3.26 or 2.58.

4.75 or 2.95~

3.98 or 2.76' enrichment %

136 126

'161 144 36(100)S6.3 16(100)26.7 29(100)59.2 16(100)35.6

11) Percent total UO in high power rods IDI 62i #

IUI ##

45

12) MCliFR at 122% Power Multi Channel Model 1.56 (Ref 4c)*-

1.54' (Ref 4c)*

!!ulti Rod Correlation 1.53 1.58~

(Ref S a)

(Ref Sa)

13) Percent Power Generated (Ref 2b)

(Ref 4 a&b)

(Ref 2a)

(Ref 4 a&b) in Low power rods / bundle 4.0 11.1 4.3 15.85 Intermediate rods / bundle 0

44.4 0-28.55 liigh power rods / bundle 96.0 44.5 95.7 55.60

14) Percent bundle power generated in highest 2.8 2.8 3.43 3.55 power rod s

• MCilFRs are reported for the intermediate performance assembly only. New thermal. hydraulic correlations have been used to calculate the MCHFRs'and therefore a direct comparison of MCHFRs.

is not valid.

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e 1

l' 2

3 4

- 15) Variation of power in high power generation 4

rods within each bundle 8.5 2.0 7.5.

3.0

16) liigh Power Rods 12 0'4.3%'

1264.3%

qi:

Enrichment - ea bundle 16 0 5.0%

1210 5.0%'

8 0 5.6%

5 @ 5.6%

{ 8 @ 4.5%

/

' 8 @ 5.0%

' 4 0;5.6%

8 @ 6.5%

4 0 15.5%

~~,

17) liighest rod power 1.81-1.83 1.688 1.71-1.89 1.596 factor / bundle Lowest Rod Power Factor 0.10-0.21 0.29 0.09-0.24 0.289' Bundle

'1.688 or 5.83 1.596 or 5.5

18) Ratio liighest Rod Power 0.289.

0.29 Lowest Adjacent Rod (Ref 2a)

_(Ref 2b)

.1.71_ or 19.0 1.81 or 18.1

'0.09 0.10 (Ref 3a)

(Ref 3a) 1.89orh.9 1.83 or 8.7 0.24 0.21

19) Moderator /UO 2

2.6 1.98 i

W/F ratio f

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20) Ratio llot Rod Old 1.81'or 1.83 or 1.07-1.08 1.71 or 1.89

- or 1.07 or 1.18 Ilot Rod New 1.688 1.596

21) Ratio Design llot R_od (Ref 3a) 1.83_or 1.01 1.89 or l.11 i

Actual llot Rod 1.81 1.71 (Ref 2a, 2b) 9

22) Max. No. Iow power 6

1 low 6

4

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rods adjacent to 3 intermediate hottest rod

23) Fuel Depletion av.

15,000 10,000 15,000-

'10,000 Bundle FNd/T liigh Power rods 20,000 15,000 cale 20,000 15,000 cale (Ref 2c)

(Ref 4d)

(Ref 2c)

(Ref 4d) d e

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. March 3,fl971

[ EVALUATION m

-The ilntermediate Performance' Fuel assembly, D-50 -(an 8 x 8 array _of fuell

rods irradiated ; for a ;three-month period with five other centermelt fuel

~

x assemblies ini the Eig: Rock Point. reactor followed:by an additional irradi-y stion period of,about eight months after the other five centermelt' assemblies

~

ha'd been removed from the reactor to await destructive evaluation of-fuel-

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rod performance)jwill.-nut be re-inserted into'.the core.

This-fuel. assembly

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with-. hot. rod average 1 exposures of 10,000 mwd /TU has been permanently removed

-~from the. center =elt fuel irradiation program because of failure of many rods.

TheLfailures are attributed to severe clad deterioration caused by-accel-( }

~

xac erated corrosion o'n:the' clad surface where clad overheating had occurred.

Similar failures were. observed in normal reload fuel that was irradiated-during the same period..-Excessive' crud deposition and:high heet fluxes werefreported to be the main ~ factors involved in creating the high temper-

~

fatu're condition. Neutrographs of one, entire rod with; incipient failure

?[

ahowed'no hydriding of the cladding at the inside surfaces. The absence

~GR; of such hydriding is.an indication-that there was no problem with contam-gig;7 ination'of;the fuel with hydrogenous impurities.

In'other words, examin-~

i;

- ation of: fuel rods'from the. failed centermelt fuel assembly has revealed l?"

that the failures were not: caused by impurities in the fuel, that clad

~ ~ " = ~

temperatures were exce.ssive and that there was an unusual accumulation of crud on the centermelt fuel rods and other reload fuel rods which

~

also failed. We have conclud2d that :these failures are~ not due to' the-Thigh centerline fuel'te=peratures in the.centermelt fuel assembly and we agree that the resultant higher than normal heat fluxes can cause signi-ficantincreases-incladtemperatureyggysstherate:ofcrudaccumulation is reduced.

The licensee has stated that-the new centermelt fuel

)

assemblies.uill'not be inserted into the Big Rock Point reacter until there

'is. reasonable assurance that crud deposition on fuel rod surfaces has been-

,significantly reduced to prevent local clad heating.

Based on the evidence presented, we. agree ~that this is prudent.

The. licensee plans to insert the five centermelt' assemblies remaining from the original centermelt

~

irradiation program some time in the future af ter 1) chemical cleaning of j.

the fuel' rods to remove the' crud accumulated during three months of in-

~

core irradiation and 2) substitution of low power rods-for the high power density, rods in the outside rows of fuel rods. The net effect of 4-

-item 2 will be to reduce the_ number of high power rods from 188 in the i

six original, assemblies to 170 in the 7 assemblies to be retained in the l

centermelt irradiation program. We have concluded that this reduction in i.

the nu=ber of high power rods and co=pliance with the previously approved

' restriction that centermelt fuel assemblies be no closer than 16.5 inches center-to-center reduces the accident consequences belev.those that were reviewed and accepted for the original six centermelt assemblies. The licensee also~has reported that the reactivity value for the new assemblies t-t O

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March 3, 1971 v

is lower than the core average thereby reducing fuel rod reactivity worths and increasing shutdown margins.

Based on line 9 of the table,' we

.here is a significant reduction in U235 and therefore a cor-concur that t

rssponding reduction in reactivity.

Similarly, the reactivity coefficients melt'fuelassemblies:(geslightlymorenegative,thantheoriginalcenter-for the new' assemblies

~ and therefore acceptable since the severity of teccidents will not be increased beyond the values calculated for the original-p cix centermelt fuel assemblies. We concur that 1) the hot cell' examinations ofirradiatedcenter=eltfuelhavebeencompletedaj) reported (3)(6)in accoidance with the requirements.of Amendment No. 1

, 2) the accumulation-cf crud on fuel rods should decrease with time, 3) the rate of crud accumula-tion on the center =elt fuel rods should be measured -(at each refueling out-cge),.4) the instrumented Reload-F assembly rod may give insight into the crud deposition prablem, and 5) the operational experience with centermelt fuel so far warrants a continuation of the centermelt fuel irradiation program.

Wa have noted the following inconsistencies in the application.

The MCHFRs-at the 122% overpower condition are increased although water enthalpy must have increased because the rods adjacent to the high power rods generate significant power in contrast to the depleted fuel rods in the original cantermelt assemblies. The Advanced Perfomance high power rods may not achieve'the objective of molten fuel at the center because,) based on a

.=

2 at 122% power (48, the rated reported heat flux of 535,000 Btu /hr. ft hastfluxisexpectedtobe440g0 Btu /hr.ft.2,toolowtocausemelting at the center of the f g rod (

~

at the start of life when such heat fluxes are attainable {

We also note that clad failure has been attri-buted to a very localized clad overtem ture condition resulting in accelerated corrosien of the cladding of centermelt and reload fuel rods, and that a centributory factor, crud accumulation, has been identified but the precise cause of local failures has not been determined. We are satisfied, however, that the failures were not caused by molten fuel conditions.

Additional infor=atien_has been requested from the licensee to obtain greater accuracy in the comparative evaluation of the original and new centemelt performance characteristics but this infomatiori is not required x

to' complete our safety evaluation. The general impression of some backing down from the original objective of definite center melting is evident. We have concluded that the hazards of operation with the two new centermelt essemblies and five of the original, chemically cleaned, and reconstituted assemblies are no greater than those considered in our previous evaluation of centemelt fuel assemblies for the Big Rock Point Nuclear reactor (1) and that the Technical Specifications may be changed to permit reactor operation with the two new centermelt fucl assemblies and five of the original centermelt fuel assemblies in the c.anner proposed by the licensee.

1

~

- Files ~

- 10,

March 3,:1971~

a

' CONCLUSION

'The two new;centerme tl ' fuel' assemblies and five cleaned ~and~ reconstituted

~

centermelt. fuel assemblies together~will include 172 fuel rods that will

operate'with.centermelting or.near centermelting. temperatures, 16 less

~

- than were contained in the original six centermelt fuel assemblies.;' The total energy in the high performance fuel rods-as a result.of this changc is lower than the value considered in our original: evaluation-The mechanical

~

design and fuel distribution of the new elements have been-improved. We have-

's therefore concluded that operation of the' Big Rock Point. Nuclear Reactor in' 3

the manner proposed:by Consumers' Power Company kill not increase the proba--

bility of or change the consequences of the design basis accident nor does it

. involve.significant hazards. considerations not descr.ibed or implicit in

. the Safety Analysis Report for Amendment No. I to the operating license or.

impair the effectiveness of engineered safety systems. (Core Spiay).

There is reasonable assurance that the health and safety-of the public will not-be endangered by. operation of the Big-Rock Point Nuclear Reactor with~two new and five previously irradiated centermelt fuel assemblies in

~

,the core and therefore, pursuant to Section 50.59 of 10.CFR Part 50,-the-

.g e 2

'. Technical Specificaticas ~ of Facility License No. DPR-6 should be changed.

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1 as proposed.-

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'V m Q

1mes J.. S ea.

Operating Reactors Branch #2 r--

Division of Reactor Licensing i

P.nclosure :

References cc:

.D.'J. Skovholt, DRL

~

.R.

H. Vollmer, DRL t

3

.D.

L. Ziemann, DRL J. J. Shea, DRL R. M. Diggs, DRL Mary Jinks (2) k

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6 i

~-

ja References r

k

-1.: DRL; approval' to operate the Big Rock Point reactor with six centermalt fuel bundles in the core.

Amendnent No. 1 to Facility Operating License No. DPR-6 date'd

}brch 12,1968.

Page'3.- Comparison with " Type C" Fuel.

gg a.

2.. Consumers Power Company Proposed Change No. 13 dated May 26, 1967.

a.

. Figure 14 Individual Fuel Rod.

Relative Power at b.

.. Figure 15 Beginning of Life o

Page 8 - Centermelt fuel exposure.

c.

3.

Special Report No. 10 dated April 7, 1969

" Performance Evaluation of Centermelt Fuel af ter the First Period of irradiation in the Big Rock Point Reactor":

Figure 2 - Page 16 - Rod Local Power Factors

~

a.

4.. Consumers Power Company Proposed Change No. 24 dated December 21, 1970 -

"Information on New Centermelt Fuel Assemblies":

c __,.

a.

Fjjare 1, Rod Power Factors.

k; b.

Table 1, Enrich =ent Distribution.

Table 5, Therncl-Hydraulics at 122% Power Level.

c.

d.

Bundle Average Exposure - reactivity and rod power, page 14.

Reduction in cru'd deposition, page 4.

c.

f.

Centercelt fuel asse=bly reactivity, page 14.

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g.

MCHFR, page 21.

h.

Figure 4.

k.

Cause of clad failure,-page 23.

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2' -

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-it bundle power, 'page 17.

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3;

.ar Company -~ Answers to DRL questions

. dated August 15,-1967.

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'MCHFRs, page 10.

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- 6. 'E; fr -L February: 1971 - " Failure Analysis and Performance Eval-Fl5!tc[

- e.

.~iermediate Performance-Centermelt Fuel after 10 000 Wd/T

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no;hydriding of the cladding 'at the inside surfaces.

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y;g 3 1971 Docket i;o. 5J-155 Consumers Power Company ATTN:

Mr. Gerald J. Walke l

Nuclear Tuel Management Ad=inistrator 212 West Michigan Avenue Jackson, Michigan 49201 Change No. 24 Gentlemen:

License No. DPR-6 We have reviewed your Proposed Change No. 24 dated December 21, 1970, to the Technical Specifications of Facility License No. DPR-6 for reactor operation with two new centermelt fuel bundles and five of the original six centermelt fuel bundles.

Tne new centermelt fuel bundles are different from the original center-melt assemblies in the following respects:

1) sheet metal corner angles t~

have been eliminated, 2) removable cobalt targets have been placed in new stainless steel corner tubes, 3) there are fewer (16 compared with n

29 and 36) high power fuel rods, and 4) rod-to-rod power gradients have been reduced. Before the five original center = cit fuel assemblics that were irradiated during April, May and June 1967 are returned to the Lig Roch Point nuclear Reactor, the crud accunulated during that irrcdiction period will be rc=oved by che=ical cicaning and cight of tne hign power rods in die outer rows will be rcplaced by low power reds. With two new centerncit fuel bundles in the core, there will be a total of 32 higa powcr rods at or near center cit conditions at

. ro ced power. When the five original centermelt fuel assemblics have beca reconstituted and reinserted into the Eig Rock Point core, the totc1 nu=ber of high power rods will be 172 compared with 188 for the i.

j.

six centermelt fuel bundles as originally fabricated and irradiated in the Big Rock Point core in March 1967.

Uo have concluded that the proposed chcnge does not present significant hazards considerations not described or implicit in Consumers' Safety Analysis Report and Proposed Change No.13 dated !by 26, 1967, and crproved by D11 Amendment No. I to the Fccility Opercting License No. DPR-6 dated March 12, 1968.

There is recsonable accurence that the health and safety of the public will not be endangered by operation of tnc Dig Rock Point Nuc1ccr Reactor in the canner described by Consuncrs f.L [ L M b I

Consuncrs Power Coupcny. MA.

3 1971 Power Corpany with two ncv centerccit fuel bundles or with the two nca centernelt fuel bundles and five of the original centerecit fuel bundles as proposed.'

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

Sincerely,

=

-.E-

.=

Peter A. Morris, Director j;+

Division of Reactor Licensing

~

Enclosure:

Attachment A - Changes to

=

Technical Spe.:ifications cc: George F. Trowbridge, Esquire DistiAbution W. Dooly, DR ACRS (3)

R. Engelken, CO (2)

R. DeYoung, DRL H. Shapar, OGC R. S. Boyd, DRL N. Dube, DRL (5)

D. J. Skovholt, DRL J. R. Buchanan, ORNL R. H. Vollmer, DRL T. W. Laughlin, DTIE D. L. Ziemann, DRL Document Room J. J. Shea, DRL Leocket File R. M. Diggs, DRL DR Reading DRL Reading Branch Reading

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ATIACHMENT A CHANGE NO. 24 TO TECHNICAL SPECIFICATIONS

- FACILITY LICENSE EO. DPR-6 CONSUMERS POWER COMPANY

=

DOCKET NO. 50-155

'l.

Change the first paragraph of Section 8.1 to read:

"8.1 The general' dimensions and configuration of'the developmental

. fuel de:Igns_ shall be as shown in Figures 8.1 through 8.7.

Principal design features -shall be essentially as shown in Table 8.1."

JE 2.

Section 8 - Figures:

LAdd Figure 8.6 - New Intermediate Performance Fuel Centermelt Assembly (8 x 8 Array) and Figure 8.7 - New Advanced Performance Fuel Centermelt

~~

.=

Assembly (7 x 7 Array).

3. ' Delete Table-8.1 and insert the revised Table 8.1.

~

4.

Table 8.2 - Change the number of centermelt fuel bundles to read:

Centermelt

" Number of Bundles Intermediate

. Advanced 1

3 Pellet UO2 Powder UO2 1

2 5.

Change Section 8.2.1(c) to read:

"(c) Fuel Examinations _

Nondestructive examinations of selected fuel rods in the centermelt fuel bundles shall be performed 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. The bundles shall be reconstituted with replacement fuel rods and may be returned to the core f or continued irr~adiation."

Tablo 8.1 RESEARC11 AND DEVELOPMENT FUEL TYPES New New Centermelt Centermelt EEI

Centerme lt -

Centermell Cencral Intermediate Advanced _

" Modified E-G" 00 -pug Intermediate Advanced 7_

Geometry, Fuel Rod Array 8.x 8 7x7 9x9 9x9 8x8 7x7 Rod Pitch, Inch 0.807 0.921 0.707-0.707 0.807-0.921.'

Standard Fuel Rods per Bundle 36(3) 29(3) 29(1' 2, 4) 0 (6' 7) 60(1)-

45 7)

Special Fuel Rods per Bundle 28 20 81 4

4

-Spacers per Bundle 5

5 3

3

.5 5

Fuel Rod Cladding if Material Zr-2 Zr-2 Zr-2 With Various Zr-2

'Zr-2 Zr-2 Initial Mechanical 4

Properties Zr-3Nb-lSn Standard Rod Tube Wall, Inch 0.035 0.040 0.040 0.035 0.040-Special Rod Tube Wall, Inch 0.035 0.040 0.040 0.040 0.031 0.031 Fuel Rods Standard Rod Diameter, Inch 0.570 0.700 0.5625 0.570(8)>

0.700(8)

Special Rod Diameter, Inch 0.570 0.700 0.5625 0.5625 0.347 0.347 Fuel Stacked Density, Percent 94 Pellet 94 Pellet 94 Pellet (5) 82 Powder 92-93 Pellet 92-93 Pel Theoretical 85 Powder 85 Powder Active Fuel Length, Inches Standard Rod 67.3 65-66.3 70 70 67.3 66.3 Special Rod 64.9 Central, 68.6 R movable Fill Gas IIelium Helium Helium Helium Helium-Helium See attached page for footnotes.

(Revised-with Change No. 24 iosued 3/3/71.) '

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' Footnotes to Table 8.1 O )Ifodified E-G and EEI UO -Pu0 and new centermelt fuel bundles may contain (in the corner regions.of.the bundic):

2 2

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

(

fuel bund]cs have a special central fuel rod la which the bundle. spacers are' ff.idifiedE-GandEE1UO-Puo$heinteriorbundlefuelrodsareremovabicandmay'containUO-Puo 2

I""I'

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fixed.

In addition, two of 2

2 (3)Special rods have depleted uranium.

(4) Also has four gadolinia-containing rods.

,a (5)With 3% dishing on selected rods.

(0 fu 1 r d stack density will vary from 74 to 92% theoretical by using annular, dished, or nondished UO -Puo2 pe$letsinselectedrods.

(

baltcornerrodsandoneempty(waferda udduringoperatio!) rods,eightgadolinia-containing r ds similar to standard UO ro sour removable-Puo Sixty-four UO -Pu0 2 spacer rod.

rods, fou*

.4

( } Diameter of cobalt targets. inside SS corner tubes.

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