ML18046A640

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Application for Amend to License DPR-20 to Revise Tech Spec Re Fuel Storage.Exxon Rept XN-NF-542, Palisades Nuclear Generating Station Spent Fuel Storage Pool Criticality Safety Reanalysis, Encl
ML18046A640
Person / Time
Site: Palisades Entergy icon.png
Issue date: 05/11/1981
From: Hoffman D
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Crutchfield D
Office of Nuclear Reactor Regulation
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NUDOCS 8105190288
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consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201 * (517) 788-0550 May 11, 1981 Dir.ector, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Bra.heh No 5 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 -

LICENSE DPR PALISADES PLANT ~ TECHNICAL SPECIFICATION CHANGE REQUEST - FUEL STORAGE Attached are three ( 3) original and thirty-seven ( 37) copies of a request for a change to the Palisades Technical Specifications.

The proposed changes define the upper boundary for enrichment of assemblies to be stored in the Palisades spent fuel pool.

Supplementary information is provided in Attachment A.

The requested changes involve a single issue.

Therefore, a check in the amount of $4,000 is attached pursuant tolCCFR 170.22, Class III change.

David P Hoffman Nuclear Licensing Administrator CC Director, Region III, USNRC NRC Resident Inspector-Palisades Plant

\\

8 1 15 1 9 0 2 B 8 Aool s

1/c.f o w It.I. E,c I,:

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  • ~.

CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in the Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972, for the Palisades Plant be changed as described in Section I below:

I.

Change A.

Change the second sentence of Section 5.4.2(c) to read:

"The high capacity spent fuel storage racks are designed such that an assembly comprised of 208 fuel rods with a maximum U-235 loading of 44.11 grams per axial centimeter or an assembly containing 216 fuel rods with a maximum U-235 loading of 45.80 grams per axial centimeter placed in the racks would result in a keff equivalent to S.95 when flooded with unborated water."

B.

Delete Section 5.4.2(b).

II.

Discussion Consumers Power requested that E'xxon determine the maximum average enrichment of Palisades Plant fuel bundles that could be stored in the Plant's spent fuel storage racks.

Future fuel designs were expected to exceed current Technical Specifications conditions stated in Sec-1 tion 5.4.2.

Specifically, the. increase*in.the number of fuel rods contained in 12.of the 68 assemblies included in the next fuel batch will increase the.U-235 linear density.

Subsequent analysis performed by Exxon Nuclear Company (XN-NF-542) has shown the addition of 8 rods in the 12 assemblies, has a negative effect on the resulting keff and, as a result, a Technical Specifications change was found not to be necessary for this particular batch, In the same analysis, Exxon also determined the maximum allowable enrichment of a fuel assembly corresponding to a keff of less than or equal to.95 when flooded with unborated water.

The enrichment obtained from this analysis was then used to calculate the appropriate linear density specified in the attached calculations.

It is requested that the calculated linear densities be incorporated into the Technical Specifications to provide an upper boundary of assembly enrichment.

Also pertaining to the spent fuel pool, Sec-tion 5.4.2(b) should be deleted since all of the original fuel racks have been removed and only the high capacity racks described in part (c) remain in the pool.

nu0581-0124a-43

III.

Conclusion Based.on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed these changes and find them acceptable.

CONSUMERS POWER COMPANY By-fiWJ~~/

R B DeWitt, Vice President Nuclear Operations Sworn and subscribed to before me this 11th day of May 1981.

2

~~di~

(SEAL)

Helen I Dempski,NOtacybiiC Jackson County, Michigan My conimiss.ion expires December 14, 1983.

., '\\ :**1'

  • -~;.....

~',-

I' nu0581-0124a-43

ATTACHMENT A to Technical Specification Change Request Fuel Storage

TYPED FROM ORIGINAL Consumers Power Company

. Nuc~ear Services O.ept~

Engineering Calculation Theoretical Density Total Dish Volume Subject CALCULATION Q}

MAXIMUM ALLOWABLE u235 LINEAR DENSITY By Date Chkd Adjusted Theoretical Density (l-.Ol)(0.94)xl00 Page_1_

Of Date 94%

1%

93,06%

Pellet Diameter 0.3500 inch Pellet Length Density of U02 Enrichment

. 2 Volume of Pellet= TI

(

0 3~

00

) (0.274)in 3 = 0.0264 in 3 3

0.0264 in3 x 16 : 39cm

=

o.4320 cm3 in 0.274 inch 3.8%

Dish volume is taken into consideration in the Adjusted Theoretical Density 235 j

Amount of U per Pellet 1

235 I 3

lo.96g 6 )(0.038)(235)+(0.962)(238)

(

8 )-

476.....

~u=----

(o.432cm )(

cm3 )(0.930 (0.038)(235)+(0.962)(238)+(2)(16) 0

  • 03 -O.l pelletl 235 gu2 35 208 pellets
  • in

)

I U

Linear Density= (0.1476pellet)(

bundle

) ((0.274in)(2.54cm)

I

=

.u2~

44. llg/ cm - bundle For a 208 rod array For a 216 rod array 235 6
  • gU 21 pellets)* (

. in

)

= (O.l47bpellet)(

bundle (0.274in)(2.54cmJ I

235

  • I

=

45.80g~ '-bundle.

  • For purposes of this calculation assume a plane 0.274 inches wide passing through the bundle.

The linear density of the plane will be the same as the bundle.

I i

I

XN*Nf *542 *.

PALISADES.NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS MAY 1980 RICHLAND, WA 99352 EJ!{ON NUCLEAR COMPANY, Inc.

8 1 t5 l *9 ~ 2 9 2

Prepared by:

XN-NF-542 Issue Date: 06/12/80 PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS C. O. Brown, Licensing Engineer Critical Hy Safety Concurred by: /?: ~

t,/lo/f'o L. E. Hahsen, Senior Sp~cialist Criticality Safety and Security Approved by:

R. N1 son, Manager Corporate Licensing and Compliance May 1980 E)){ON NUCLEAR COMPANY, Inc.

U.S. CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Exxon Nuclear Company's warranties and.representations concerning the subject matter of *this document are those set forth in the Agreement between Exxon Nuclear Company, Inc. and the Customer pursuant to which.

this document is issued.

Accordingly, except as* otherwise expressly provided in such Agreement, neither Exxon Nuclear Company, Inc. nor any_ person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed _any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method or process 0 diselosed in this document.

  • The information contained herein is for the sole use of Customer.'

In order to avoid impairment of rights of Exxon Nuclear Company, Inc.

in patents or inventions which may be included in the information con- '

tained in this document, the recipient, by its acceptance. of this document agrees not to publish or make public use (in the patent sense of the term) of suc::h information until so authorized in writing by Exxon Nuclear Company, Inc. or until after six (6) months foll()wing termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement.

No rights or licenses in or to any patents are implied by the furnishing of this document.

XN-NF-F00,765

TABLE OF CO_NTENTS PAL1SADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS INTRODUCTION.

SUMMARY

FUEL ASSEMBLY DESCRIPTION SPENT FUEL STORAGE POOL DESCRIPTION CALCULATIONAL METHODS RESULTS OF PREVIOUS STORAGE POOL CRITICALITY SAFETY EVALUATIONS RESULTS OF NEW STORAGE* POOL keff CALCULATIONS..........

STORAGE POOL ACCIDENT CONDITIONS CONCLUSIONS -.

REFERENCES..

XN-NF-542 Page No.

l 2

2 3

4 4

6 7

14

Table No.

I I I I I I XN-NF-542

  • LIST OF TABLES.

PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS PALISADES (EXXON NUCLEAR BATCHES HAND I)

FUEL ASSEMBLY PARAMETERS.......

CALCULATIONAL RESULTS OF THE PALISADES SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS BY P. SOONG (NUS) IN JULY 1978................

PALISADES NOMINAL MAIN POOL STORAGE ARRAY keff VERSUS ENRICHMENT CALCULATlONS.....

Page t~o.

8 9

10

Figure No.

2 3

LIST OF FIGURES PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS PALISADES (EXXON NUCLEAR BATCHES HAND I)

FUEL ASSEMBLY DESIGNS.........

MArn POOL STORAGE CELL-NOMINAL GEOMETRY ARRANGEMENT..............

P/\\LISADES MAIN POOL STORAGE ARR/\\Y HORST CASE keff VERSUS ENRICHMENT......

XN-NF-G42 Page No.

11 12 13

INTRODUCTION PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS XN-NF-542 I~ November 1976 Consumers Power Co. submitted to the USNRC a request I

t~ install high density spent fuel storage racks at the Palisades NJclear Generating Station.

The subsequent safety analysis (l) showed the pool to be adequately subcritical for fuel assemblies enriched to 3.05 wt. % 235u.

With the trend in fuel management toward higher burnup, average reload fuel assembly enrichments have increased to as high as 3.267 w/o, (Exxon Nuclear Batch H).

Thus, in July 1978 a reanalysis( 2) of the high density storage racks was per-formed by NUS Corp. for the average Batch H reload fuel enrichment of 3.267 w/o.

The intent of this reanalysis is to. define a maximum average fuel assembly enrichment (i.e. maximum axial 235u loading), which will continue to meet applicable criticality safety criteria.

SUMMARY

The criticality safety reanalyiis of th~ Palisades spent fuel storage pool, as described in Reference 1, and this report demon-strates the pool to be adequately subcritical, i.e. keff ~ 0.95 at the 95% CL(confidence level), for fuel assembly axial 235u loadings up to and including 44.11 g/cm (3.80 w/o). The worst case keff of

.;. 1 -

XN-NF-542 the racks is estimated to be 0.946 at the 95% CL for fuel assemblies enriched to 3.80 w/o.

FUEL ASSEMBLY DESCRIPTION

  • The Exx6n Nuclear Bate~ H and I fuel assembly designs are depicted in Figure l. As indicated, the 15xl5 lattice arrangement includes a single instrument tube, eight guide bars, and eight locations for removable poison rods, gadolinia-bearing fuel pins or water rods.

Current plans call for a Batch I reload of 68 fuel assemblies.

Of the:>~ fuel assemblies 48 wili contain eight water rods, twelve will

-contain eight gadolinia-bearing fuel pins* and eight will provide

  • ~ight poise~ rod.locations.

The fuel assembly parameters assumed in this evaluation are given in Table I.

From this information bundle-averaged cell parameters were calculated by: including*the zirconium associated with the instrument tube, guide tubes and the guide bars in the.zirconium clad of each fuel rod.

~Jater associated with each guide bar, instrument and guide tube*was included by,_increasing the unit cell dimension (rod lattice pitch).

For producing cell-averaged cross section data the above assumptions permit a conservative estimation of the.effect on reactivity

. - of the extra zirconium an.d water within the fuel assem.bly by maintaining the correct assembly water-to-fuel volume ratio..

  • SPENT FUEL STORAGE POOL DESCRIPTION The Palisades spent fuel sto~age pool consists of racks comprisi~g.the

~main pool and tilt pit pooL The main pool storage racks have 8.56"

_;. 2 -

I I*

XN-NF-542 s9uare ID storage cells and a 10.25" center-to-center spacing, see F~gure 2.

In the tilt pit pool the rack is designed to ~tore control rods as well as fuel assemblies.

Thjs rack has a 9.0" storage cell ID and cells are located on 10.69"xll.-25" centers.

f\\s d~monstrated in Reference 2 the reactivity of the main pool *Storage racks is -v2.0% tik/k higher than th.e tilt pit pool rack.

Hence, the I

main pool rackdesign is analyzed with respect to criticality safety I

a~ the limiting case.

I T~e neutron absorber plat~ is composed of s4c bonded in a carbon m~trix~ The plate is 0.21". thick and 8.26" wide with a minimum 10s I

16ading of 0.0959 g/cm2.

As shown in Figure 2 each plate is centered wjdth-wise in each storage cell wall.

CALCULATIONAL METHODS T~e KEJJO,IV Monte Carlo code( 3) was utilized to calculate the r~activity (keff) of the Palisades.spent fuel storage pool.. '1ulti-g~oup cross section data from the.XSDRN 123 group data library-were I

g~nerated for input into KENO IV using the NITAWL( 4) and XSDRNPM( 4)

I

' cbdes.

Specifically, the NITAWL code was utilized to obtain cross s~ction data adjusted to account for resonance self-shielding using i

the Nordheim Integral Method.

The XSDRNPM code~.~ discrete ordinates I

I i

one-dimensional transport t.heory code, was then used to prepare I

I s~atially cell-weighted cross section data repres~ntative of the fuel assembly for imput into KENO IV.

I XN-NF-542 RESULTS OF PREVIOUS STORAGE POOL CRITICALITY SAFETY EVALUATIONS*

In order to show the storage pool to be adequately subcritical for Exxon Nuclear Batch H fuel enriched to 3.267 w/o, a*reanalysis( 2) of the pool was performed by P. Soong of NUS* in July 1978.

In this analysis storage array keff values under nominal conditions were ca lcu 1 a ted using the 11 KENO code in conjunct ion with 123-group /l.MPX averaged. cross sections 11.(2) *Th~ PDQ~? code with NUMICE (NUS version of LEOPARD) cross sections was then used to calculate the reactivity changes resulting from variations in storage rack.conditions. The reactivity for "worst case" conditions was then calculated_ by summing the KENO calculated keff for nominal conditions and the

~keff values c_alculated using the PDQ-7 code.

  • Table II summarizes the results of these calculations for both the main pool and.the tilt pit pool.

l.

  • RESULTS OF NEW STORAGE POOL keff CALCULATIONS In order to demonstrate the criticality safety of the storage pool for higher fuel assembly enrichments, additional keff calculations were performed using *the KENO IV code; *using a geometric model of the nominal main pool storage arrangement as described in Reference 2 and depicted in Figure 2, keff was calculated for E~~on N~clear Batch H fuel (See Table I) enriched to.3.267 w/o.*. The resulting keff value of 0.875 +/-.. 005 represents an infinite array and fs within two standard deviations of the ~ENO keff reported by P. S_oong in Reference 2 and shown in Table JI of this report.
  • Hence, based on this result it is concluded that the calculational model used for this calcu-lation gives conservative results for the main storage pool.

XN-NF-542

The above duplication calculation for Batch H fuel represented a fuel
  • assembly design of 208 active fuel rods.-

To determine the effect on

!array keff of a higher fuel a;sembly axial 235u* loading based on the*

i inumber of fuel rods, the above case was rerun with 216 fuel rods.

I

For this case keff was calculated to be 0.365 +.005.

This result 1indicates a slightly greater reactivity worth of the water holes in I

!the 208 fuel rod arrangement relative to having those positions

filled with eight additional fuel rods.

Since the fuel assembly

'design with fewer active rods gives a higher array keff' all I

'subsequent calculations assume 208 fuel pins.

I i Having established a representative calculational model, additional I

I \\reactivity calculations were performed for increased fuel assembly I i enrichments in the nominal main pool storage arrangement.

These i

! results are summarized in Table III and are shown graphically in 1 Figure 3.

I I

Also shown in Table III are final worst case keff values estimated
at the 95% confidence level for the storage pool based on fuel 1

b l

. l 2 3 5u 1. d.

a s s em y ax 1 a.

o a rn g. These values were calculated from the I

I 1 following expression:

I I

I I

'. where, keff (WC)= keff(N6m) +*2a + W + T k ff(Nom) = the nominal storage array keff ae-one standard deviation W = Root-mean square of the worst case

. tolerance.variations, see Table IL T =Maximized moderator temperature variation, see Table II.

XN-NF-542 No reactivity penalty was assessed for either s4c particle self-shielding or calculational bias defined by benchmark calculations.

In the KENO IV calC:ulational model B4c was conservatively modele.d to account for the eff~cts of self-shielding (i.e. the boron was lumped in the center of the neutrbn absorber plate). With regard to calculational bias Exxon Nuclear Co. has had extensive experience using the above computer code calculational model and ben~hmark calcula-tions(5) show no significant bias using these methods as described.

STORAGE POOL ACCIDENT CONDITIONS Since the storage pool will under normal operating conditions

. ( 1 ).

contain ~2,000 ppm soluble boron in the water, the actual k~ff of the storage array, based on the de~c~ibed pool conditions, will be ~20% lo~er(

2

) than calculated. Thus, for the accident conditions of a fuel assem~ly lying either across the racks or up against the outs.ide of the racks, the storage array reactivity will remain well below the limiting value of 0.95.

In the event of a single failure in the storage pool cooling system, based on the assumed conditions presented in Section 6 of Refererice l, the bulk pool temperature woul.d not exceed ll8°F for a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> normal off load.

For this condition the maximum surface tempe~ature of a fuel rod is less than 230°F providing greater than 9°F margin to local boiling*.(l).* Both initial (2200 MWt) and stretch (2650 MWt) power cores and their applicable design peaking factors have been considered in establi~hing limiting thermal conditions. (l) From

  • XN-NF-542 the standpoint of neutronics, any localized fuel rod surface boiling inside the fuel assembly will have a negative effect on array re-activity (i.e. keff of the array will decrease).

CONCLUSIONS This analysis conservatively demonstrates* the reactivity of the Palisades spent fuel storage pool for fuel assembly axial 235u loadings of 2. 44. 11 g/cm (3.80 w/o for Batch I fuel with 208 active fuel rods) to be less than 0.95 under existing assumptions of worst credible storage array conditions described in this report.

Hence, at the 95% confidence level the keff of the storage pool will be

< 0.946.

The analytical efforts, the results of which are presented in Table Ill, were reviewed by a second party knowledgeable in the performa*nce of criticality safety evaluations. I.

XN-NF-542.

TABLE I

Palisades (Exxon Nuclear B~tches H and I)

Fuel Assembly Parameter~

Lattice Pitch, i~.

  • clad oo,* in.

Clad Material Clad Thickness, in.

U02 Pellet Diameter, in.

Pellet Density, % Pr Percent Dish *.

No. Active Fuel Rods Ave. Enrichment Rod Array Eff. Array Dimensions, in.

No.. Guide Bars (Solid Zr)

No~ Guide Tubes

  • GT OD, in.

GT TK, in.

No. Instrument tubes IT OD, in.

IT TK, in.

- 8,;,

Nominal 0.550 0.417 Zr-4 0.028 0.350 94 + 1.5 1.0-208 (Batch H) 208, 216 (Batch I) 3.267 (Batch H ~ 208 acti~e rods) 3.260 (Batch I - 208 active rods) 3.232 (Batch I - 216 active rods) l 5xl 5 8.25 x 8.25 8

8 0.416 0.011 l

0.415 0.029

Case 1

2 3

4 5

6 7

8 9.

. 10.

. 11 12 13 14 15 16

  • XN-NF-542 TABLE II Calculational Results of the Palisades Spent Fuel 2,torage Pool Criticality Safety Reanalysis by P. Soong (NUS) in July 1978 Batch H Fuel (3.267 w/o)

Description Main Pool Tilt Pit Pool Enrichment Variation U02 Density Variation

  • B c Slab Width B~C thickness and loading Variation in Spacing

. Storage Can Dim. Variation

  • B4C Slot thickness B4C Slab Missing (l)

Bow and Twist Storage Can thickness Temperature Variation TOTAL*

Two Standard Deviations Particle Self Shielding _(B4C)

Benchmark Bi as

  • Nominal Storage Lattice Celi koo PDQ-7 0.8569 0.8397 KENO 0.8693*+.0042 0.8516 +.0034 W6rst.tas~ Par~metrics, 6k

.0037

.0006

.0008

.0038

. 0081

.0097

.0045

.0018

.0165

.0061

.0228* (Root-mean square sum)

. 0028*

Other, 6k

.0084*

.0040*

.0086*

Total Sum of (*) Values, k00 17 Maximum Credible Worst Case 0.9309

  • These Values summed to establish maximum "worst case" storage pool keff'

\\

XN-NF-542 TABLE III Palisades Nominal Main Poo1 Storage Array keff Versus Enrichment Calculations.

Fuel Assembly Parameters - See Table I

    • Storage Array Description - Main Pool (Nominal)

See also Figure 2 Nominal Axial 235u KENO IV Enrichment,

( 123 group)

Case w/o Loading, g/cm keff :_ a 1

3.267 37.92 0.875 +.005 2

3.50 40.63 0.892 +.005 3

3.70 42.95 0.906 +.004 4

3.80

44. 11 5

3.90 45.27

. o. 918 +.004 For Cases 1, 2, 3, and 5 worst.case values are calculated as follows:

keff(WC)_ = keff(Nom) + 2cr + W + T.

where W = 0.0228 (Root-Mean square sum of tol~rante variations, see* Table II).

T = 0.0028 (Maximized moderator temperature variations).

Worst Case*

at 95% CL 0.9106 0.9276 0.93.96 0.9460 0.9516 Case 4 worst case keff value taken from graph, see Figure 3. keff (est. )
  • XN"".NF-542 FIGURE l' I **

PALISADES (EXXON NUCLEAR BATCHES &'AND H) FUEL ASSEMBLY DESIGNS LL L L.G LL LL LG LL LL L L L H H H H H H H H H L L L L L L H H P H H H P H H L L L L H H H H H H H H H H H H H L G H H H H H H H H H H H H*H G L H P*H H H H H H H H HP H L L H H H H H H H H H H H H H L L H H H H H H I H H H H H H L L H H H H H H H H H H H H H L L H P H H H H H H H H H P H L G H H*H H H H H H H H H H HG L H H H H H H H H H H H H H L L L L H H P H H H P H H L L L

  • L L L H.H H H H H H H H L L L L L L L G L L L L L G L L L L L M M M G M M M M M G M M M L M M M H H H H H H H H H M M M M M M H H P H H H P H H M M M M H H H H H H H H H H H H H M G H H H H H H H H K H H H H G M H P H H H H H H H H H P H M M H H H H H H H H _H H H H H M.

M H H H H H H I H H H H H H M M H H H H H H H H H H H H H M.

M H P H H H H H H H H H P H M.

G H H H H H H H H H H H H H G M H H-H H H H H H H H H H H M M M M H H P H H H P H H M M M M M M H H H H H H H H H M M M L M M M G M M M M M G M M M L Exxon Nuclear Batch H L - 2.90 w/o Fuel H = 3.43 w/o Fuel G = Guide Bar I = Instrument Tube P = Poison Rod Location Exxon Nuclear Batch I L = 2.52 w/o Fuel M = 2.90 w/o Fuel H = 3.43 w/o Fuel G = Guide Bar.

I - Instrument Tube P = 2.52 w/o Fuel with 4.0 w/o Gadolinia*

or, Water Hole; or, Poison Rod Location

  • Gadolinia-bearing pin loc-ations may vary slightly from r locations as shown.
    • Correction of tY})ographical error.
., 11 -

Ins Wat Ga (0.

ide er p

155")

_J I

Poison Plate:

B4C in Carbon Matrix*

W1dth - 8. 26" Thickness - 0.21" 10s Ldg. - 0.0959 g/cm 2 (mi~.)

--t-~----

I I

ll:Si~l&il --... &-

~~~?!'::~

XN-NF-542

+

I I

~*

I Poison p late 25")

I 1r..

I Slot (0.

~-

I

. :~

I:

I Outside I

Gap ( 0.

Fuel Assembly I

I 8.25" x 8.25 11 I

Stainles I

I

  • l
0. 125" I

I I

I I

I-!

I Stainles (0. 25" a;;g:;;:;p; _...... #!?

rs;==*ew~ -

I

. + -_*.....:._ -- - - - - -.,.. I Storage Cell:

10.25" Center-to-center Can ID-8. 56" Can 00-9.56" Inside Water Gap-0.155" Outside Water Gap-0.69" Note:

With the poison plate in place a gap of O.d4 (total) is allowed. in the poison plate slot.

Figure

  • 2 Palisades ~1ain Pool Storage Cell Nominal Arrangement

. Hater 69")

s Steel thick s Steel thick)

e.
  • X'l-'ff-5~2 0.96 (Plotted values represent keff ! 2cr 0.95
0. 946 +-------

0.94 Array kef f

. 0. 92

0. 91 0.90 ----+-------'-+-----+------+------+--~

3.0

3. l 3.3 3.5 wt

~~- 235u 3.7 3.9 4.0 i

Figure 3

.Palisades Horst Case ['1ain Pool keff Versus Enrichment y

  • XN-NF-542 REFERENCES
1.

"Spent Fuel Pooi Modification Description and Safety Analysis,"

Consumers Power Company, Palisades Nuclear Generating Station, Docket No. 50-255 (Nov. 1976).

2.

Soong, P., "Criticality Analysis for 3.27 w/o Enriched Fuel Palisades High Density Fuel Rack,"

NUS Corp. (July 1978).

3.

L.M. Petrie and N.F. Cross, "KENO IV:

An Improved ~1onte Carlo Criticality Program," ORNL-4938, Oak Ridge National Laboratory (November 1975).

4.

N.M. Greene, et al., "AMPX - A Modular Code System for Generating Coupled Multigroups Neutron-Gamma Libraries from ENDF/B, "ORNL-TM-3706, Oak Ridge Nationa 1 Laboratory (March 1976).

5.

C.O. Brown, "Criticality Safety Benchmark Calculations for Low-Enriched Uranium Metal and Uranium Oxide Rod-Water Lattices", XN-NF;..499, Exxon Nuclear Co.

(April 1979).

- 14..,

PALISADES NUCLEAR GENERATING STATION SPENT FUEL STORAGE POOL CRITICALITY SAFETY REANALYSIS Distribution C. 0. Brown ( 2 )

L. E. Hansen R. Nil son H. G. Shaw Consumers Power Co. (10)/HG Shaw

  • Document Control (5)

XN*NF-542 ISSUE DATE: 06/12/8,

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