ML18043A403

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Forwards Documents Updating 781218 Submittal Requesting Tech Spec Changes to Allow Storage of Highly Enriched Fuels in New Spent Fuel Storage Locations
ML18043A403
Person / Time
Site: Palisades Entergy icon.png
Issue date: 01/12/1979
From: Hoffman D
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Ziemann D
Office of Nuclear Reactor Regulation
Shared Package
ML18043A404 List:
References
TASK-09-06, TASK-9-6, TASK-RR NUDOCS 7901160256
Download: ML18043A403 (37)
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Text

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  • Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201
  • Area Code 517 788-0550 January 12, 1979 Director, Nuclear Reactor Regulation Att Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR-20.

PALISADES PLANT - PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST ~ FUEL STORAGE The following documents are provided to update our submittal dated December 18, 1978 requesting Technical Specifications changes to allow the storage of higher enriched fuel in both new and spent fuel storage locations. -

1. Page 18 of the NUS Report entitled "Criticality Analysis for 3.27 Weight Percent or W/O Enriched *Fuel Palisades High Density Fuel Rack. 11
2. Exxon Report XN-309 - "Palisades New Fuel Storage Array - Criticality Safety Analysis."

235 1.

3. Examp1 e
  • c al cula t ions or u inear d ensiuy. .....
4. Corrected Technical Specifications P.age changes incorporati.ng final Batch H design data. These changes have been discussed with the NRC staff.

David P Hoffman (Signed)

  • David P Hoffman Assistant Nuclear Licensing Administrator cc JGKeppler, USNRC r.:1 i\ F

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  • ATTACHMENT 2

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XN-309

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JULY, 1975

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    • . ,* RICHLAND. WA 99352 J '

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  • E>}"(ON NUCLEAR COMPANY, Inc.

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  • XN-309 PALISADES NEW RJEL STOAAGE ARRAY .

CRITICALI1Y SAFE1Y ANALYSIS

  • July, 1975 EXXON NUCLEAR COMP.ANY, INC .

XN-309 PALISADES NEW RJEL STORAGE ARRAY CRITICALI1Y SAFE1Y ANALYSIS i' '

Prepared by: 2//Ylzr Date '

Criticality Safety Approved by:

~~NeChodom,ger Licensing and Compliance W. E. Niemuth, Manager Contract Performance

~-r------------

- Quality Assurance and

_ Licensing

XN-309 TABLE OF CONI'ENI'S Page No.

INTRODUCTION 1 SlM-1ARY 1 FUEL ASSEi\1BLY DESCRIPTION 2 STORAGE ARRAY DESCRIPTION 5 CALCULATIONAL ME1HODS 7 RESULTS 11 CONCLUSIONS 16 APPENDIX A (Second-Party Review) 18

. i i

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  • XN-309
  • INrRODUCTION American Nuclear Society Standard Ai~S-Nl8.2 stipulates that new fuel storage arrays be designed to limit the reactivity of such aTTays to a value of < 0.98 for all credible conditions of moderation. Further, it is suggested. that such conditions of moderation include.

full flooding or the envelopment of the array in a tmifonn low density aqueous foam as could credibly exist as a result of fire fighting.

As originally designed, the Palisades new fuel storageaTTay location and design was felt to preclude the addition of moderation. The 3 x 24 aTTay of fuel assemblies, therefore, was not designed to remain sub-critical in the event of water flooding or the addition of low density hydrogeneous materials within and between the stored fuel assemblies.

To comply with suggested limits established in ANS-Nl8.2; the ConslilTiers Power Company has proposed a reduction in capacity of the array from 72 to 36 fuel assemblies located in a checkerboard array with alternate locations occupied by structural steel box beams. This doclUTient de-I scribes the criticality safety analysis o~ that proposed storage array and demonstrates compliance with the suggested lirriits established in ANS-Nl8. 2.

sm.-MARY Criticality safety analyses of the proposed new fuel storage array con-taining steel box beams in alternate storage locations demonstrate the safety of the aTTay for all credible conditions of moderation. Array

- I -

XN-309 reactivities were computed using the KEN0-2 lvbnte Carlo code assuming fully flooded* conditions and the presence of low density aqueous mater-ials uniformly distributed both within and between the stored fuel assemblies. Calculated reactivities are well below the suggested limit established in ANS-Nl8.2 for new fuel storage arrays. Tbe highest reactivity (keff = 0.896 _:'.:. .007) occurs when the array is fully flooded by water.

RJEL ASSEMBLY DESCRIPTION A typical Palisades fuel assembly design (Reload E(l)) is depicted in Figure 1. As indicated, this arrangement.includes a single zirconium instn..nnent guide tube located in the center of the assembly and eight zirconium guide bars positioned on the exterior of the assembly. lX)Z fuel rods within the assembly contain uranium of three different 235u enrichments.

Actual "Reload E" fuel assembly specifications and two different assumed conditions evaluated as part of this analysis are given in Table I.

These conditions include existing and bt.mdle averaged cell parameters.

The bundle averaged cell parameters were calculated by including the zirconium associated with the guide bars and instn..nnent guide tube in the zirconium clad thickness of each rod. Water within the assembly was included by increasing the unit cell dimensions (lattice pitch). Such assumptions permit an estimation of the effect of the extra zirconium and water within the fuel assembly .

  • It shoult.! he noted that the analysis discussed herein assumed a single ennc. Iunent o 1* .)~ . 2 wt "o 235 tJ f or a 11 ro <l s.

'l11c nominal "Reload E"

XN-309 FIGURE 1 RELOAD E ASSEMBLY rt_ WG fl WG_* !_____________ ,_______ ,_

I I

I L

I -1 L

~-~~~ G M M M M M G M

+--

M L IL I L M M I M M M M M M M M M M M L I

I M M M H H H H H H H H H M M M I

I -

M M H H H H H H H H H H H M M I

I.

I G M H H H H H H H H H H H M G I

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

I

- M M

M M. H H H H

H H H

H H"

H H

I H

H H"

H H

H H

H

--- - - 1 H

H M

M M

M II I M -;-r~;--- If H H H H H H H H H M JM I

M M H H H H H H H H H H H M M

-- f-- - --

G M H H H H H H H H H H H M G M M H I!

H H H 1-1 H H H H H M Ir*1 M M M H H H H H H

H H H

M M

M I

L M M M M M M M M M M M M M L I

  • I L L M M G M M M M M G M M L L I

Ii_ NGI I

--------- --+-*--

L = 2.40 w/o U-235 Fuel  !

M = 2 . fll wIo lJ -2 J.5 Fuel II ] . t'll w/o lJ-?..15 Fuel r;

111*.11*111111*111 l1i1J1*

XN-309 TABLE. I PALISADES RJEL ASSEMBLY r .. PARAMETERS i

I

\ Blll1dle I Lattice Averaged Actual(l)

  • Cell Cell (Nominal) Parameters Parameters Lattice Pitch 0.5500" 0.5500" 0.5~13" i

Clad OD 0.4150" 0.4150" 0.4241" Clad Thickness 0.0285" 0.0285" 0.0331" uo 2 Pellet Diameter 0.3505" 0.3505" 0.3505" Pellet' Density (%TD) 94+1. 5 96% 96%

Percent Dish 1.0 0% 0%

Avg. Enrichment (wt % U-235) 3.04 3.20* 3.20*

Rod Array 15xl5 15xl5 15xl5 *

  • Design base enrichment specified by Consumers Power Company( 2) .

---* XN-309

  • bundle averaged enrichment, however, is 3.04 wt % 235u. Consequently, for the water to fuel volume ratio of these assemblies, this analysis is valid for bundle averaged enrichments of ~ 3.2 wt % 235u.

S1DRAGE ARRAY DESCRIPTION The Palisades new fuel storage rack has been measured to determine actual "as built" dimensions. Figure 2 is an arrangement drawing giving those measured dimensions. This information was supplied by Consumers Power Cornpany(Z)_ Subsequent inforrnation( 3) established that the dimen-sions indicated are actually measured center-to-center distances between adjacent top bands. Such dimensions, therefore, represent nominal i center-to-center spacings between adjacent storage locations. It should I

be noted that the nominal center-to-center separation between assemblies was designed to be 9.5 inches. Measurements of the installed rack, however, show that a maximum negative tolerance of 1/8" exists on the design value (i.e., a minimum nominal center-to-center separation of I 9-3/8" was measured). It should also be noted that the plates con-i, -

necting adjacent angle irons are 5" wide and 3/16" thick. These plates establish a minimum edge-to-edge separation between adjacent storage 1ocat ions

. c*i.e., 3/16").

Concrete walls are adjacen~ to three sides of the storage array and are separated from the fuel by 0.5 to 1.5 inches. For the purpose of this analysis, a 16" thick concrete reflector were assumed to be touching three edges of the storage rack. The fourth side of the array was assumed to be reflected by 4" of water, which is effectively an in-finitely thjck water reflector.

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XN-309

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SG5Sc~: I PT B ii 9.50" :COilC.ii.S :11.C.S BU!LTi  !)[. c:*:s. :mrn-!ISOUTH PE~ ... ELL AT aor1a:*: jf: ~~CK .

s.o* *.r 9.44" Cl/I&" TOO s:-:ALU ~ B G.K. EUILT ;:..~ S:-JOUL:l FE

-1/8 3. 36" C1/8 TQO S:lALU -11~ B cJ. 33" - mo s:*;.Lu PALISADES NEW FUEL STORAGE MEASUREMENTS OF INSTALLED RACK FIGURE 2

  • XN-309
  • 1he proposed arrangement for storing Palisades fuel assemblies is in-dicated in Figure 3. Calculations for this arrangement result in the assumption of a 3 x 25 array of storage locations. 1his "checkerboard" array contains fuel bundles with alternate positions occupied by 8"x8" r structural steel box beams having a nominal wall thickness of 5/16 11 ( 2).

A minimum wall thickness of 0.25" was established for this analysis( 4).

In addition to the nominally spaced array of fuel assemblies shown in Figure 3, it is quite possible that the fuel assemblies will not be centered in each storage location. As previously noted, however, each storage location is bounded by a 3/16" plate. Hence, a minimum edge-to-edge separation between adjacent storage locations is assured. A

  • I

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section of the array considered to evaluate the effect of abnormal arrangements of fuel assemblies is indicated in Figure 4.

CALCUIATIONAL METIIODS

.( .

1be KENO-II Monte Carlo code( 5) was utilized to calculate the reactivity I

,_ of the Palisades ne\v fuel storage array. Multigroup cross section data (18 energy groups) utilized in these calculations were averaged using the CCELL( 6), BRT-1(?), and GAMfEC-II(S) codes. Specifically, the cross section data for various regions within the storage array were averaged as follows:

CCELL - Utilized to obtain cell averaged multigroup cross section data for fuel rod-water lattices. Such calculations included both the bundle averaged cell parameters and the

  • actual lattice cell parameters (See Table I) .

PALISADES NEW FUEL STORAGE ARRAY (NOMINAL SPACING)

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Concrete ~ I/\ -../\llm.lo

= 1. 0 Concrete D Fuel Bundle Location (8.25"x8.25")

~ Steel Box Beam (8"x8" Outside Dimension)

Center-to-Center Spacing of Units = 9-3/8" FIGURE 3

,- I XN-309 FIGURE 4 A3NORMAL ARRAY ARRANGEMENT 4" Water

(

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16" Concrete f .

f Fuel Blll1dle (8. 25"x8. 25")

Steel Box Beam (8"x8" Outside Dimension)

All.lllrinum Straps Resulting in Minirm.nn Edge-to-Edge Separation of 3/16" (Represented as Void Filled with Appropriate Density H o) 2 Minurm..nn Center-to-Center Distance Between J\<ljaccnt Aluminum Straps = 9-3/8" XN-309 BRT Thennal group (.::_ 0.683 ev) cross section data for the structural steel box beams were averaged using the Battelle Revised IBERMJS code. Such data were averaged assuming a 0.25" thick region of iron separated from the rod-water lattice by the thickness of interspersed water (at varying densities) that \vould exist in the nominally spaced storage array (0.4375"). Epithennal multigroup data were averaged over a slowing down neutron energy di~tribu-

! tion in water.

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GAMTEC-II - Multigroup cross ~ection data for water and con-i I

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crete were averaged over neutron energy spectra character-istic of infinite media of these respective* materials.

concrete was asstnned to have constituents as specified by The the Constnners Power Company( 9). (Due to the fact that cross section data for sulfur were not readily available, and since the atom density of sulfur represents only 0.06% of the total constituents, the effects of sulfur in the con-crete were ignored.)

l.

In addition to the codes identified above, the XM:C ~nte Carlo code was utilized to verify the accuracy of CCELL calculations for undennoderated

('.

rod-water lattices. The XMC code is a puesdo-point energy M:>nte Carlo code c~ 90 energy groups) which pennits the discrete representation of the entire fuel assembly .

XN-309 RESULTS Values of k were computed using the CCELL code for 3.2 wt % 23 5u rods 00 as a function of water density within the unit cell.

  • Both lattice cell and bundle averaged cell parameters (See Table I) were assumed to gain an insight into the reactivity effects of the zirconium guide bars and I -

the water associated with guide bars and the instrtm1ent guide tube. The results of those calculations are given in Table II.

Comparison of the calculated values of k indicates that the uncertainty 00 associated with the zirconium and water extraneous to the actual lattice cells is quite small. - Indeed a maximum effect of 5 rnk was calculated

  • I I.

assuming these materials were unifonnly distributed throughout the bundle.

The CCELL code has been used extensively at Exxon Nuclear for averaging rnultigroup constants in rod-water lattice configurations. Theory-experirnent correlations(lO) indicate that the reactivity of measured critical lJ0 2 rod water lattices can be calculated to within 18 rnk.

Biases associated with such calculations appear to be conservative in nature.

No experimental data are known to exist at water densities consistent

\. with those assumed in this analysis. k5 a consequence, in an effort to detennine the validity of the CCELL code for water densities of< 0.6 g/on3 , the ::-0.fC Monte Carlo code was used to calculate koo of the "Reload E" bundle with all rods assumed to contain uranium enriched to 3.2 wt %

235 u. Water interspersed between the rods was assumed to have a density XN-309 TABLE II Water Infinite Media MultiElication Factors Densi~y Lattice Cell Blilldle Averaged

! * (g/an )

Parameters Cell Parameters 1.00 1.401 1.402 0.75 1.369 1.371

o. 50 1.299 1.302 l 0.30 1.187 1.190 0.20 1.090 1.093 0.15 1.023 1.023 i

I

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0.10 o_.05 0.936 0.816 0.934 0.811

~ .

XN-309 3

of 0.2 g/cm . The value of k calculated for this case was 1.079 00 +

.005. This value compares favorably to that calculated by the CCELL code using bundle averaged cell parameters (1.093) or lattice cell parameters (1.090). These calculations indicate that the CCELL code calculates, with reasonable accuracy, the spatial and energy distri-bution of neutrons within the lattice cells at water densities far below the values where experimental data have been utilized to confinn the I validity of this calculational method. It appears, therefore, that the calculational biases will result in the calculation of array reactivi-I I ties which are conservative.

I.

For the nominally spaced "checkerboard" array shown in Figure 3, re-activities were computed as a function of water density with such water uniformly distributed.both within and between the fuel assemblies.

results of those calculations are tabulated in Table III and shown The graphically in Figure 5. These reactivities were calculated asstiming bundle averaged cell parameters. Figure 5 also includes bundle-averaged values of k calculated using the CCELL code and,. for comparison, in-00 dicates the single check calculation performed using the XMC Monte Carlo

. code.

The highest array reactivity occurs in the fully flooded condition. For all water densities, however, the array reactivity is well below the limiting value of 0.98 established in .ANS-Nl8.2. The highest calculated reactivity (cited at the 95% confidence level) is 0.910. All available checks of the validity of this calculated value indicate a probable conservative bias of 1 to 2% k.

XN-309 TABLE III

\ ....

Fractional Water KENO-II Calculated Density Array Reactivity i .

  • i i 1. 00 0.896 + .007 0.75 0.825 + *.007 a.so o. 783 +*
  • 007 0.30 0.792 + .006 0.20 0.786 + .006

.. - 0.15 0.789 + .006 0.10 0.755 + .006 0.05 0.669 + .006 I

P..\LlSADES NEW HmL STORAGE ARHAY CALCULATED REACTIVITIES 1.4 Vl I-<

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µ 1.3 u Cll

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µ Cll 0.95 1. 2 u

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µ r-i

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H 0.85 1.0 "d Q)

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e 0.75 0.8 2 KENO Calculated Array Reactivity A CCELL Calculated Bundle k"° ~ I 0.70 y XMC Calculated Bundle km VI

.0 t.O 0.651--~~~~~~~~~~~~~~~~~~~~~~~...-~~--~..~.~--~.~.,,-,...,~,.._.........~~~~~~~-'

0.0 0.1 0.2 0.3 () * '1 0:5 0.6 0.7 0.8 0.9 1.0 i:ractional Water Density 1: IC!JRE 5

XN-309 In addition to the nominally spaced array, the array of fuel assemblies spaced as indicated in Figure 4 was also evaluated. For this abnormal arrangement, which assumes the minimum possible spacing between bundles, water densities of 1.0 and 0.2 g/cm3 were assumed and the resultant calculated reactivities for such conditions were 0.914 .+ .007 and 0.812

~ .005, respectively. Hence, even for abnormal conditions, the re-activity of the array remains well below established limits.

CONCLUSIONS This analysis conservatively demonstrates that the proposed Palisades

( . new fuel storage array will remain subcritical under all conditions of

  • ~oderation as could credibly exist through the uniform addition of water at any density both with1n and between the stored fuel assemblies.

is important to note, however, that the array might not be critically It safe if moderators are added to individual stored fuel assemblies with-out the addition of such materials between the assemblies. It is believed valid to assume that this is not a credible moderator arrange-ment which could result from any accidental addition of water.

I 1he analytical efforts described herein were reviewed by an independent

! second party knowledgeable in the performance of criticality safety evaluations. _ 1his independent assessment of the adequacy of this analysis is discussed in Appendix A.

e.

XN-309 REFERENCES

1) F. D. Lang, G. R. Correll, and K. P. Galbraith, "Final Design Report for Palisades Fuel," XN-74-32, Exxon Nuclear Company, Inc.,

October, 1974.

2) Letter, W. J. Beckius to W. E.. Niemuth, May 13, 197 5.
3) W~ J. Beckius, Personal Communication, Consumers Power Company, Jtme 10, 1975.
4) B. Webb, Personal Communication, Const.nners Power Company, June 10,

(

1975.

l 5) G. E. Whitesides and N. F. Cross, "Keno - A M.iltigroup ~nte Carlo Criticality Program," CTC-5, Union Carbide Cor]?oration Nuclear Division, Septemb~r, 1969.

6) W. W. Porath, "CCELL Users Guide," BNW/JN-86, Pacific Northwest Laboratories, February, 1972 *
  • 7) 8)

C. L. Bennett and W. L. Purcell, BRT-1: Battelle Revised 1HERMJS,"

BNWL-1434, Pacific Northwest Laboratories, June, 1970.

L. L. Carter, C. R. Rich~y, arid L. E. Hus hey, "GAMI'EC..: II :

for Generating Consistent :M.lltigroup Constants Utilized in Dif-A Code fusion and Transport Theory Calculations," BNWL-35, Pacific North-west Laboratories, March, 1965.

9) Letter, W. J. Beckius t~ W. E. Niemuth, June 3, 1975.
10) U. P. Jenquin and D. R. Oden, Verification of Neutronic Design Methods Using Lattice Criticals as Benchmarks," BNW/JN-118, Pacific Northwest Laboratories, December, 1972 .

XN-309

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APPENDIX A

  • .4'~B

~~ '* u aaueue Paciiic Northw<'SI Laboratori<'s DJ!e July 11, 1975 .

To L. E. Hansen From D. R. Oden ,.C.,f!~

Subject Q.A. Review of Criticality Safety Calculations For The Palisades I

Plant New Fuel Storage Rack

\ .

We have completed the Q.A. review of the criticality safety calculations for the Palisades Plant New Fuel Storage Rack, and have found no significant errors or omissions. We concur with your con-clusion that keff of the loaded rack, under the conditions specified,

\'.

is less than 0.98 at a 95% confidence level for all*degrees of neutron moderation.

Infonnatton provided by Exxon Nuclear and reviewed by BNW is summartzed in Attachments I and II. This consists of letters of cor-

  • f res-pondence, a copy of a drawing of the storage rack, and several computer printouts of Exxon Nuclear calculations. All of this infonna-tion is being returned to you at this time.

Items checked in the Q.A. review included the following:

)

  • Code input parameters including n~clei densities for all materials
  • Computer code opttons selected
  • Assumptions made tn problem set up and analysts versus problem definition in correspondence
  • Lattice and array geometry tn CCELL and KENO
  • Method of homogenization to obtatn bundle average parameters
  • Region/material correspondence in CCELL and KENO
  • Results of analysis
  • Adequacy of computer codes used.

Concerning this last item of code validation, there is a degree of uncertainty. The CCELL code has gained acceptance as a reliable method of generating cross sections for uo 2-water lattice through its

  • L. E. Hansen Page 2 July 11, 1975 use at Exxon Nuclear for criticality safety and fuel design work. The KENO code is us*ed throughout the industry for the calculation of arrays of fissile material in criticality safety applications. However, as you are aware, the lack of experimental data on lattices with low density water moderation makes it impossible to completely validate methods of

,- . calculation in this area. With no evidence .to the contrary, one must I

i thus take the position that if a model is validated for the full water density uo2 lattice criticals it will adequately handle the low water density cases, which may not be tenable.

Although it is not a validation of the array keff calculation, the fact that the XMC calculated bundle average k agrees well with that 00 from CCELL for the 20% water density case, lends a degree of credibility i .

to the CCELL cross section generation technique. In the absence of

  • li l .

experimental data this is about all that can be done short of setting up the whole array calculation in XMC. The cost of this approach must however be weighed relative to the benefit res~lttng from having yet another (although more sophisticated) calculated result.

j I ..

  • ATTACH!-~ENT I.

REFERENCE LIST OF CORRESPONDENCE AND DRAWINGS l ..

PROVIDED BY EXXON NUCLEAR FOR PALISADES STORAGE RACK Q.A. REVIEW

\ . 1. Letter, W. J. Beckius to Wayne Niemuth, 5/13/75.

2. Letter, G. R. Correll to L. E. Hansen, Criticality Analysis of Palisades 11

"! New Fuel Storage Racks 5/19/75.

11

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1

  • 3. Letter, Wayne Niemuth to W. J. Beckius, 5/27/75.

\

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4. Letter, W. J. Beckius to W. E. Niemuth, 6/3/75.
5. Figure 2.4.2, Reload E Assembly. *
6. Dra~ing, Palisades New Fuel Storage - Measurements of Installed Rack.

1 7. Sketch, Palisades New Fuel Storage Array (Nominal Spacing)

8. Sketch, Abnonnal Array Arrangement.

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9. Tables, Two Pages of Exxon Calculated Results.
10.
  • Figure, Array Reactivity and Bundle Averaged k Versus Fractional Water Derisity 00

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  • ATTACHMENT II LIST OF COMPUTER PRINTOUTS PROVIDED BY EXXON NUCLEAR FOR PALISADES STORAGE RACK Q.A. REVIEW Reference Number Code Date Title 1 GAMTEC 6/9/75 Palisades Concrete (No Sulfur) 2 THERMOS 6/24/75 Fe thermal data 0% void 25% void 50% void 70% void l' 85% vofd 90% void

(

i 95% void

! 3 THERMOS 6/23/75 Fe thermal data 80% void 4 (5 parts) XMC 6/13/75 Bundle k~, 80% void 5 CC ELL 6/9/75 Lattice Avg. Cell 0% void j

6 7

8 CC ELL CC ELL CC ELL 6/9/75 6/12/75 6/9/75 Bundle Avg. Cell 0% void

  • Bundle Avg. Cell 70% void 50% void 25% void Bundle Avg. Cell 80% void 85% void 90% void 95%I void 9 CC ELL 6/9/75 Lattice Avg. Cell 80% void 85% void 90% void 95% void 10 KENO 9/26/75 Worst Spacing, 80% void 11 KENO 9/25/75 Worst Spacing, 0% void 12 KENO 9/25/75
  • 0% Void Het. Fe 13 KENO 9/25/75 25% void Het. Fe 14 KENO 9/25/75 50% void Het. Fe

' 15 KENO 9/25/75 70% void Het. Fe 16 KENO 9/25/75 80% void Het. Fe KENO 9/25/75 85% void Het. Fe 17 lB KENO 9/25/75 90Z void Het. Fe 19 KENO 9/25/75 95% void Het. Fe

1--

  • Active Fuel Length EXAMPLE CALCULATION OF u235 LINEAR DENSITY Based on January, 1979, Exxon Data 131.8 :!: 25 in Weight of U02 per rod 2117.48 :!: 42.35 grams Average enrichment (Batch H) 3.27 :!: .05 High enrichment 3.43 :!: .05 I. Using Average Enrichment (3.27 :!: .05 = 3.32%)

Weight of Uranium only (u235 + u238)

= (2117.48 + 42.35)g .o33 2 7""""":.0~3~3~2""=~-t--l~.,...,..,.<++~<4--............~~

= (2159,83)g(.8814)

= 1903.67 g Weight of u235 only

= (1903.67)(.0332)g

  • Linear density (g/cm)

= 63.2g

_ (63.2g of u235 rod)(208 rods)

- 131. *- .25 in 2.5 cm in

= 39.34 g/cm u 235 per Bundle II. Using High Enrichment (3.43 + .05 = 3.48)

Weight of Uranium only (u235 + u238)

= (2117 48 + 42 35)g (.0348)(235)+(.9652)(238)

  • * (.0348)(235)+(.9652)(238)+(2)(16)

= (2159.83)g(.8814)

= 1903.67g Weight of u235 only*

= (1903.67g)(.0348)

= 66.25g Linear Density g/cm gU235

- (66.25

- rod* * )(208 rods)

(131.8-.25 in)(2.54 cm/in)

= 41.24 g/cm u235 per bundle Note: Values for active fuel length and weight of U02 per rod was obtained by phone from Exxon Nuclear on 1/8/79, These are to be the final design para.meters.

  • 2-Active Fuel Length CALCULATION OF u235 LINEAR DENSITY Based on October, 1978, Exxon Data 131.8  :!:. .25 in Weight of U02 per rod 2101.2 gms Average Enrichment 3.27  :!:. .05 High Enrichment 3.43  :!:. .05 I. For Average Enrichment Weight of Uranium only (u235 + u238)

(.0332)(235)+(.9668)(238)

= <2101

  • 2g) (.0332)(235)+(.9668)(238)+(2)(16)

= (2101.2g) (.8814)

= 1852.00g Weight of Uranium - 235 only

= (1851.99g)(.0332)

= 61.49g

  • Linear Density ( g/ cm)

(61.49 g u 235 rod)(208 rods)

=. 131. - .25 in 2.5 cm in)

= 38.28 g/cm u235 per bundle II.

= 1852.00 g Weight of u235 only

= (1852.0g) (.0348)

= 64.45g Linear Density g/ cm

= 40.12 g/cm u235 bundle

  • 3-Active Fuel Length (in)

CALCULATION OF u235 LINEAR DENSITY Based on Batch G Fuel Data 131.8 .:!: .25 Weight of uo 2 (ems) 2104.o :!:. 23 Average Enrichment (Batch G) 3.00 :!:. .05 High Enrichment 3.20 :!:. .05 I. Weight of Uranium Only (u235 + u238)

(.0325)(235)+(.9675)(238)

= ( 1

  • 02 )( 2104
  • 0 ) (.0325)(235)+(.9675)(258)+(2)(16)

= (2146.08)(.8814)

= 1891.56g Weight of u235 only

= (1891.56g)(.0325)

- 6l.48g

-* Linear Density (g/cm)

= (61.48g (131.

= 38'.27 u235 rod)(208 rods)

- .25 in 2.5 cm in) g/cm. u235 per bundle Our present Technical Specifications (for Batch-G Fuel) assume the highest enrichment and limit us to a maximum lineer_density of 38.3 g/cm u235 *

  • 5.3 5.3.2 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) (Contd)

Reactor Core and Control

a. The reactor core shall approximate a right circular cylinder with an equivalent diameter of about 136 inches and an active height of about 132 inches.
b. The reactor core shall consist of approximately 43,000 Zircaloy-4 clad fuel rods containing slightly enriched uranium in the form of sintered uo2 pellets. The fuel rods shall be grouped into 204 assemblies.

A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution.

c. The fully loaded core shall contain approximately 211,000 pounds uo2 and approximately 56,000 pounds of Zircaloy-4.

Poison may be placed in the fuel bundles for long-term reactivity control.

d. The core excess reactivity shall be controlled by a combination of boric acid chemical shim, cruciform control rods, and mechanically fixed boron rods where required. Forty-five control rods shall be distributed throughout the core as shown in Figure 3-5 of the FSAR.

Four of these control rods may consist of part-length absorbers.

5.3.3 Emergency Core Cooling System An emergency core cooling system shall be installed consisting of various subsystems each with internal redundancy. These subsystems shall include four safety injection tanks, three high-pressure and two low-pressure safety injection pumps, a safety injection and refueling water storage tank, and interconnecting piping as shown in Section 6 of the FSAR.

5.4 FUEL STORAGE 5.4.l New Fuel Storage

a. Unirradiated fuel may be stored in the new fuel storage rack which is designed to ensure an effective multiplication factor of less than O. 98 under the worst credible conditions for fuel enriched to 3. 30 weight percent U-235.

5-3

  • 5.4 FUEL STORAGE (Contd)
b. New fuel may be stored in shipping containers.
c. New fuel enriched to 3.27 weight percent U-235 may be stored in the poi-soned high capacity racks which are designed to ensure an effective multi-plication factor of less than 0.95 when flooded with unborated water.
d. The new fuel storage racks are designed as a Class I structure.

5.4.2 Spent Fuel Storage

a. Irradiated fuel bundles will be stored, prior to off-site shipment in the stainless steel-lined spent fuel pool.
b. The spent fuel racks are designed to maintain fuel in a geometry which insures an effective multiplication factor of 0.95 or less with new fuel flooded with unborated water.
c. The spent fuel pool water boron concentration shall be verified at least once monthly to be e~ual to or greater than 1720 ppm .
  • d.

e.

The spent fuel racks are designed as a Class I structure.

The fuel placed in the spent fuel pool and stored in the poisoned high c~pacity storage racks shall not contain more than 41.24 grams of U-235 per axial centimeter of active fuel assembly subject to a maximum as-sembly average loading of 3.27 weight percent U-235. The fuel placed in the spent fuel pool and stored in the unpoisoned lower capacity racks shall not contain more than 38.3 grams of U-235 per axial centimeter of active fuel ass.embly, subject to a maximum assembly average loading of 3*. 05 weight percent U-235.

f. Spent fuel shipping casks shall not be moved in the fuel storage building until such time as the NRC has reviewed and approved the spent fuel cask drop evaluation.
g. Fuel stored in the higher capacity storage racks as described in the SER supporting Amendment No. 28, shall have decayed for a minimum of 12 months if the storage racks are not supported by similarly designed, adjacent rac~s and the spent fuel pool wall or the cask anti-tipping device.tl)

References (l)Until needed for fuel storage, two A-type racks in the northeast corner of the spent fuel pool will be removed and replaced with the cask anti-tipping device to provide necessary seismic restraint.

FSAR, Appendix B.

5-4

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