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SIEME-NS

CRmCALllY SAFETY ANALYSIS

FOR THE PALISADES SPENT FUEL

STORAGE POOL NUS RACKS

OCTOBER 1991

NOTICE

EMF-91-174(P) Issue Date: l 0/04/91

This document contains·information proprietary to Siemens Nuclear Power Corporation: 11 1s submitted in confidence and 1s to be used solely for the purpose for which 1t :s furnished and returned upon. request. This document and such 1nforma11on 1s not to oe reproduced. transmitted. disclosed, or used otherwise 1n whole or in part wtttiout the written authorization of Siemens Nuclear Power Corporation. This document. in whole or in pan. and any contract resullong therefrom. shall not oecome 01no1ng until the respective authorized representatives of both pan1es e•ecute 1ne documenl(s) comprising the contract and an executed copy is promptty returned to 1ne original signatory.

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t.

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PLEASE READ CAREFULLY

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EMF-91-174(P) Issue Date: l 0;04; 91

CRITICALITY SAFETY ANALYSIS



Prepared by: &r~wi C. D. Manning, Critl Uty Safety Specialist Safety, Security, and Licensing

Reviewed by: ~-!;,,.,._, SecondP Reviewer

Approved by: ~~ T. C. Probasco, Supervisor, Safety Safety, Security, and Licensing

Approved by: ~ w.~nager, · satety, Security, and Licensing

Date: 1014/t:tt I

Date: lo/4./ 'i (

Date: /o/yij/

Th18 docu- contaift'I Siemen• Nuc!Mr ~ Cotpotalion procne<aty •nformarion Md I• 1ubj«t lo !lie rHUiclion1 on !lie flm or title pa9e.

.·-.·

TABLE OF CONTENTS

EMF-91-174(P) Page i

Section Page

1.0 SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2.0 MAJOR CONSERVATISM USED IN ANALYSIS ............................. 3

3.0 DESIGN BASES .......................... ·. . . . . . . . . . . . . . . . . . . . . . . . . . 4

4.0 CALCULATIONAL METHODS AND MAJOR ASSUMPTIONS ................... 8

5.0 UNCERTAINTIES ................................................... 9

6.0 NORMALJABNORMAL CONDITIONS AND RESULT . . . . . . . . . . . . . . . . . . . . . . . . . 1 O

7.0 VALIDATION OF CALCULATION METHODS .............................. 14

8.0 REFERENCES .. : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

Th18 l»cull*lt contain• SI_,,. N-Power Corporation proonetaty inlormallon and i1 1ubj9Ct to Ille 1"1rtctione on Ille flrst or title page.

LIST OF TABLES

EMF-91-17 4(P) Page ii

3.1 Palisades Cycle N Fuel Bundle Design Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . 5

6.1 K-Effective for Normal Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

7.1 Benchmark Calculation Results Keno-Va with 16 Group Cross Sections . . . . . . . . . 15

-.'.•.

Thill ~ _..,. SlelMM Nuci.t Pow. Corilcra!icn propn.auy inform.Uon and Is aubj8ct to tlte '"'11ctions on the first ot title ~··

·•

UST OF FIGURES

EMF-91-17 4(P) Page iii

Figure Page

1 .1 Palisades Spent Fuel Pool Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

3.1 Palisades Main Pool Storage Cell Nominal Arrangement . . . . . . . . . . . . . . . . . . . . . . 7

Thltl dOcument contmlntl SMmet11 Nuc:J- Power Corporation pr0on_,., infonnllllon and i1 1ubjec:t to Ill• rwtrlction1 on Ille first or !tt!e J)age.

EMF-91·17 4(P) Page 1

1.0 SUMMARY

Since 1987, the Palisades Nuclear Generating Station spent fuel storage pool has

undergone various modifications. These modifications include the installation of high density

spent fuel storage racks, designed by Westinghouse, at the north end of the fuel storage pool

and on both sides of the NUS storage racks located in the north tilt pit.

This criticality safety reanalysis of the NUS spent fuel storage pool, shown in Figure 1 ,

demonstrates the maximum k-eff of the NUS racks loaded with the 15x15 fuel assemblies at worst

credible conditions to be .9131 at the 95/95 confidence level. Storage of 15x15 fuel assemblies

meets the requirements of NUREG-0800 and ANSI/ANS 57.2-1983 subject to the conditions given

beJow:

•

•

•

Fuel Design- As listed in Table 1 with a maximum bundle average enrichment of 4.4 wt.% U-235. Any significant deviation from this design may require additional analysis.

When stored, the fuel bundles must be fully inserted into the fuel storage boxes .

The B4C plates must contain no significant gaps or missing plates and the B4C

content must remain constant over time.

This analysis does not include spent fuel storage in the Westinghouse spent fuel storage

racks. Misplacement of a 4.4 wt.% U-235 bundle in the Westinghouse spent fuel storage rack

was not addressed.

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EMF-91-174(P) Page 3

2.0 MAJOR CONSERVATISM USED IN ANALYSIS

The conservatisms in the model include the following:

•

•

•

•

•

The fuel enrichment is modeled as the bundle average in all rod locations .

The poison plates are all modeled as having the minimum allowed width and thickness.

The modeled temperature is 20°c. The higher temperatures in the actual system will result in k-eff values lower than those reported here.

The system was modeled as flooded with pure water (zero soluble poisons) .

The NUS storage array reactivity (k-eff) was determined assuming a full array with a fuel bundle average enrichment of 4.4 Wt.% U-235 and no burnable poisons.

Thitl document c:onlaine Slemene Nucl- Power Cotporalion crocnetaty 1nform111ion and i• aubjllC! to the restriction• on the fil"lt or tili• page.

t:

•

3.0 DESIGN BASES

3.1 Fuel Assembly Description and Design Parameters

EMF-91-17 4(P) Page 4

The mechanical design for this assembly is consistent with the mechanical designs of the

recent past and foreseeable Mure. This fuel assembly is a 15x15 arrangement and includes a

single instrument tube and eight guide bars (Reference 1 ). The parameters from the cycle N

design are listed on Table 3.1 .

Thltl doculftMt c:ontaift8 Siemene NuctMr Pa- CofJIOralion procn~ 1nfermalien lllld ia autojKI to the ...utctiena en the flrat er title page.


TABLE 3.1 PALISADES CYCLE N FUEL BUNDLE DESIGN PARAMETERS

Fuel Rod Array 15 x 15

Number of Rods/Assembly 216

Number of Nonfueled Positions/Assembly 9; 8 Q!Jide bars, 1 instrument tube

Assembly Maximum Envelope Dimensions 8.3245 inch square

Fuel Rod Pitch 0.550 inch

Cladding OD 0.417 inch

Cladding ID 0.3580 inch

Cladding Thickness 0.0295 inch

Fuel Rod Active Length 131 .8 inches

Fuel Rod Pellet-to-Clad Diametrical Gap 0.0070 inches :.. ···:. ... ·.· .. · ·.· .. <·.;.;.·.·.· .. ·· :·.·· .. ··.-:.; .. ·:·:· :-;. .. ·.· ·.·: .. · .. · · .... : • .. · . ·..... . . -·· . ·'.· ... ·.·::'.;'·'. .. · · .. · .·::·.=::·.:·:.:··: ... · ;:::;:;··:-.····. : ·:-·

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Pellet OD 0.3510 inch

Density 94.5% TD

Dish Volume 1.4%

Thie ~- containt1 Siem- Nuc:lffl Pow.r Cotporalicn propnMaly rnfcrmalicn and i1 1ul:ljec! to ti!• l'M!rlctfon1 en ti!• fim or ti~• s:ia;a ..

___J

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3.2 Storage Racks Description and Design Parameters


The Palisades spent fuel storage pool consists of racks comprising the main pool and tilt

pit pool. The pool layout is shown in Figure 1. The main pool consists of NUS and

Westinghouse storage racks. The NUS racks are at the south end of the main pool in a 24x1 6

array. The NUS racks have a 2-inch water gap separating them from the Westinghouse racks

(Reference 2). The NUS racks in the main storage pool have 8.56 inch square ID storage cells

and a 10.25 inch center to center spacing. The north tilt pit pool consists of a north-to-south row

of three storage racks. The center rack is a 1 Ox5 NUS storage rack. The NUS rack is designed

to store control rods as well as fuel assemblies. This rack has a 9.0 inch storage cell ID and

cells are located on 10.69 x 11.25 inch ·centers. This lay out is also shown in Figure 1.

The storage cells each consist of an inner and outer stainless steel can. The gap

between the inner and outer can is .25 inch. Between the two cans is a neutron absorber plate.

The· plate is composed of B4C bonded in a carbon matrix. The plate is 0.21 inch thick and 8.26

inches wide with a 8-1 O loading of 0.0959 + .00959 g/cm2. Between each storage cell is a water

gap with nominal thickness of 0.69 inches. The nominal arrangement of the Main Pool Storage

Cell is shown in Figure 3.1 .

Thie ~ c:om.IM Siem- Nuc:IMt "-Corporlllion propn-.y informUion and la aubject to !lie ..-ic:Uona on ti!• first or lit!• ,,.ge.

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i ti~ l-'r

Poison Plate: .0 4C in Carbon t1atrix \.lfllth - 8.26" Thicknes~ - 0.21" 100 Ldg. - 0.0959 g/cm2

--t------ ---I

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EMF-91-174(P) .) Page 7

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Outside I ·later 69") l'iap ( o.

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0.125"

Stainles (O. 2!i"

s Steel

thick

s Steel thick)

I Storage Cell: 10.25" Center-to-center Ccin IO-a. 56" Can OD-9.5£i"

· Inside llatur Gap-0.155 11

Outside Water Gap-0.69"

Note: With the poison plate in place a ~ap of 0.04" (total) is allowed in the ~oison plate slot.

FIGURE 3.1 PALISADES MAIN POOL STORAGE CELL NOMINAL ARRANGEMENT

This ~ contalne ~ Nuc;i.r Power Cotporlllion propn~ inform&llon and Is subject ID th• .-iellon• on the flrst or !Ille page.

...

•

4.0 CALCUL.ATIONAL METHODS AND MAJOR ASSUMPTIONS

4.1 Calculational Methods


The methodology used included CASM0-3G (Reference 3) and KENO-Va (Reference 4).

Both codes share wide acceptance for fuel analysis throughout the nuclear industry.

CASM0-3G was used to determine the small changes in reactivity due to uncertainties

in the design tolerances of the fuel assembly and spent fuel storage racks.

KENO-Va was used to model the fuel storage pool. Hansen-Roach 16 group cross

sections prepared by BONAMl/NITAWL are used with the KENO-Va models. KENO-Va, BONAM I,

NITAWL and the cross section library used are all part of the SCALE 3 system and have been

extensively benchmarked against data from critical experiments. Supplemental benchmarking

using data from critical experiments with bundle arrays containing boron-containing plates is

included in section 7.0.

4.2 Major Assumptions

The major assumptions made in this analysis are as follows:

•

•

•

•

•

Fuel enrichment is the bundle average in all rod locations .

Fuel bundles contain no burnable poisons .

No soluble poisons are present in the water .

Spent fuel pool bulk water temperature is 20°c .

B4C content in the neutron absorber plates is constant over time and tie plates contain no significant gaps or missing plates.

The first four assumptions make the analysis conservative. The last is a limiting condition.

As part of Facility Change Package FC-375 each storage cell was weighed after absorber

plate loading and compared to a reference weight. Random sampling of cells with a neutron

source was also done before final approval of FC-375. Had significant gaps been present, they

· would have been found by these two checks (Reference 5).

Thlto d-conwne Siem- Nucl- Power Corporation procnetaty 1ntormaUon and i1 1ubjec1 to the r"9trlction1 on the fll9t or U~• page.

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5.0 UNCERTAINTIES

EMF-91-17 4(P) Page 9

The effect of dimensional and material tolerances on the system k-eff was determined

using CASM0-3G. CASM0-3G is a 2-D transport code used extensively for fuel depletion

calculations. CASMO delta-k values are preferred over those from KENO-Va with a Monte Carlo

uncertainty. The fuel was modeled without burnable poison and with all rods at a bundle average

U-235 enrichment of 4.0 wt.%.

The effect of various tolerances is detailed below.

• Enrichment - When the nominal enrichment was increased by 0.05 wt.% U-235, the k-infinity increased by .00215.

• Pellet Density - The nominal smear density is 93.177% TO based on a 1 .4 volume % dish and a 94.5% TD pellet density. K-inf increased by .00182 with a 96.0% TD pellet density and a 1 .4 volume % dish.

• Pellet Diameter - Increasing all pellet diameters by .0005 inch produced a .00028 rise in k-inf.

• Cladding 0. D. - Decreasing cladding 0. D. by .002 inch produced a .00186 rise ink-inf.

• Temperature - Temperature of the fuel and moderator was decreased to 4°C which resulted in an .00206 increase in k-inf.

• Inner Can Wall Thickness - The inner can wall thickness was decreased by 0.01 o inch with a resulting decrease in k-inf of .00101. The inner wall thickness was increased by 0.01 O inch with an increase in k-inf of .00085.

• Outer Can Wall Thickness - The value of .00085 from the inner can thickness is used again for the outer can thickness.

• Cell Pitch - The storage rack cell pitch was decreased by .04 inch to 10.21 inches. The resulting increase in k-inf is .00476.

The RMS sum of the nine uncertainties due to tolerances is .00631. The tolerance

uncertainty is used with the bias uncertainty and the KENO standard deviation to prepare the

95/95 upper limit of k-eff._ Section 7.0 reports the bias to be 0.0035 + .00368. The 95/95 upper

limit for k-eff is calculated using the methodology presented in Reference 9 .

Thltl docwnent contalne S'-9 NuclMt p....., Corpol'!l!lon crccnewy 1nformalion 1111d i• 1ubject to Ille .-!cticn• on Ille Rrwt or UHe page.

•

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6.0 NORMAL/ABNORMAL CONDmONS AND RESULT

EMF-91-17 4(P) Page 10

All conditions were modeled using KENO-Va and were found to have an acceptable

reactivity (k-eff) of < .95 after accounting for all uncertainties and bias. The maximum k-effective

at the 95/95 confidence level for any credible abnormal condition is .9131 . The assumption that

the entire NUS spent fuel storage rack contains 15x15 fuel with 4.4 wt.% U-235 conservatively

accounts for the presence and potentiaJ intermixing of older. and lower enriched fuel bundles.

6.1 Normal conditions

The main spent fuel storage pool was shown to bt;t more reactive than the north tilt pit

pool in Reference 6. The 24x16 NUS spent fuel storage rack was modeled as both a finite and ·

an infinite system. The resulting k-eff and a description of each model used is listed in Table 2.

·-·=-:~-;.~~-

Thi. ~ conlaffte ~ Nudew Power Corpotallon propnewy infcmullion and I• 1ut>jec:t to Ille rftl7ic:tion1 on Ill• flrwt or !!Ue pa;e.

TABLE 6.1 K-EFFECTIVE FOR NORMAL CONDmONS

Case ID Description

4.4bndlR A 15x15 fuel assembly modeled as infinitely long and reflected on four sides with 30 cm of water. Model is used for comparisons and to 'show a single dropped bundle has acceptable reactivity.

4.4infR1 An inf. x inf. array of infinitely long fuel assemblies in infinitely long NUS fuel storage racks.

4.4inf-zR An inf. x inf. array of 335 cm long fuel assemblies in fuel storage racks. The neutron absorber plate length is 339 cm. The top and bottom of the array is reflected with 30 cm of water.

4.4fin2R2 A 24x16 array of 335 cm long fuel assemblies in fuel storage boxes. The neutron absorber plate length is 339 cm. All sides of the array are reflected by 30 cm of water.

EMF-91-174(P) • Page 11

k-eff .95 .UL

.8985

.9006

.9053

.9091

•

•

•

6.2 Interaction

EMF-91-174{P) Page 12

The NUS fuel storage racks and the Westinghouse fuel. storage racks are separated by

a 2.00 inch water gap. Therefore, potential neutron interaction between the two racks was

analyzed. This analysis included modeling the entire spent fuel storage pool. The NUS racks

were modeled as described in case 4.4fin2A2. The Westinghouse racks were modeled at

nominal conditions containing 15x15 fuel bundles. All positions in the 15x15 fuel bundles were

assumed to contain fuel with 1.5 wt.% U-235. The entire fuel pool with a 2.0 inch water gap

between the two racks was shown to have k-eff = .. 9053 + .0037. This same model with a 6.0

inch water gap which essentially isolates the two racks has k-eff = .9046 + .0036. The difference

· between the two is 0.0007 and demonstrates that interaction between the two racks is negligible.

6.3 Abnormal Conditions

6.3.1 Spacing

Given the arrangement of the spent fuel storage rack and the size of the fuel bundles it

is not possible for fuel bundles to be accidently placed in the water. channels between the fuel

storage racks or between the racks and the concrete pool walls.

The only abnormal spacing condition results from eccentric positioning of the fuel

assemblies in the fuel storage boxes. In order to eccentrically space the fuel assemblies in the -

fuel boxes, the four aligning pins on the bottom of the lower tie plate must be placed on the steel

floor of the fuel storage box. Normally, these aligning pins fit into four 0.88 inch diameter holes

in the bottom of the fuel storage boxes. This type of misplacement of the fuel assembly causes

the assembly to be elevated 1.47 inches. Although the length of the poison plates only exceeds

the length of the active rod length by 2.2 inches, elevating the fuel assembly by 1.47 inches does

not give the a• portion of the fuel assemblies a line of sight path to other fuel assemblies ..

without passing through a neutron absorber plate.

The worst case model for eccentric spacing was an Infinite x Infinite array of eccentrically

spaced fuel assemblies. The fuel assembly in this model was placed at the lo~er left corner of

the fuel storage box with specular reflection placed around the fuel assembly. K-eff for this

arrangement is .9131 at the 95/95 Upper Umit.

Thi. document conlaint1 SiemltM NucJur Power Corporalion propn~ 1nformalion and i1 sub/eel to th• ml1Jfctlon1 on th• first or till• page.

·6.3.2 Shifted Poison Plates

EMF-91-17 4(P) Page 13

The poison plates in the NUS rack could be shifted to one side instead of being centered

in the poison plate channel. This condition was modeled by creating an inifinite array of fuel

storage boxes. The poison plates for each storage box were shifted to provide the maximum

possible gap between the plates at two diagonaJ corners. For this arrangement, the 95/95 upper limit of k-eff is .9080.

6.3.3 Dropped Fuel Bundle

A dropped fuel bundle during fuel handling operations is also considered a credible

abnormal event. If a fuel bundle were dropped into the spent fuel .rack and lay horizontally

across the top of the fuel storage rack, this bundle would be approximately seven inches above

the active portion of the fuel assemblies in storage. Seven inches of water separation will

decouple the fuel in storage and the fuel in the dropped bundle. A single.fuel bundle reflected

on all sides by 30 cm of water has a 95/95 upper limit k-eff = .8985. Rack design prevents a

vertlcaJ fuel bundle from being placed outside of the fuel storage racks.

Thltl docu- confllina Slememl NudMr PCoW9r ~on propn91a1y informmton and Is 1ubjsct Ill th• -c:tiona on th• firlt or tit!• peg•.

•

•

• 7.0 VALIDATION OF CALCULATIONAL METHODS

EMF-91-17 4(P) Page 14

Supplemental benchmarking of the methods employed .in the analysis was performed

using experimental data with boron poison sheets in arrays of bundles. The experiments are

described in References 7 and 8. The benchmark data are in Table 3.

Using the methods in Reference 9, the weighted average k-eff and the standard deviation

of the bias were calculated:

Weighted average k-eff: 1.0035

Bias Standard O.eviation: 0.00368

The weight of each k-eff value is proportional to the reciprocal of its variance.

--~··

This ~ ~ si--N~ Pew. Coiiiorallon propn-.y informmlion and la aubject Ill the fftll1djona on Ill• fll'lt or title page,

·.·.

·· .. ·.··.·.

I .... · ... . . : .. :··:_ '.: .. .

. ·:-.:

EMF-91-17 4(P} • Page 15

TABLE 7.1 BENCHMARK CALCULATION RESULTS KENO-Va WITH 16 GROUP CROSS SECTIONS

CASE. NO • CALCULATED K-EFF

. · ..• f. :· .:·· :. ..·.•.••·•· ••••·>••··.·· /.//REFERENcE:::t:expeR1MENj$:.:!.::::.-._:·;,:-.•.••. :-. • •.•• :_=········-·.···-· .. _.·· ••··.·• .. ··· ··•· · · ·

2378 1.00395 + 0.00376

2384 1.00037 + 0.00306

2388 0.99886 ± 0.00341

2420 1 .00038 + 0.00367

2396 0.99443 + 0.00360

2402 1.00694 + 0.00283

2411 1.01223 + 0.00286

2407 1.00647 + 0.00332

2414 1 .00967 + 0.00327

9 1.00092 + 0.00487

10 1.00181 + 0.00412

11 0.99786 ± 0.00413

12.: .. 0.99885 + 0.00487

31 1.00442 + 0.00421

Th19 document ccntal"" Si...,_ NuclMt Power Catpormian propn..wy 1ntanna!ion and i1 1ubjKI ta Ille restriction• an th• flnrt ar ti~• page.

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•

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8.0

1.

2.

3.

4.

5.

6.

7.

8.

REFERENCES

EMF-91-174(P) Page 16

Principal Reload Fuel Design Parameters Palisades Reload M. ANF-89-063(P).

Palisades Facility Change Package #FC-860.

"CASM0-3: A Fuel Assembly Burnup Program (Methodology)," Studsvik/NFA-86/8, Studsvik Energiteknik AB, Nykoping, Sweden, November 1986.

"SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200.

Palisades Facility Change Package #FC-375.

"Palisades Nuclear Generating Station Spent Fuel Storage Pool Criticality Safety REANALYSIS." XN-NF-542, May 1980.

Baldwin, M. N., et.al., "Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, "BAW-1484-7, July 1979 .

Bierman, S. R., Durst, B. M., and Clayton, E. D., "Critical Separation between Subcritical Clusters of 4.31 % Enriched U02 Rods in Water with Fixed Neutron Poisons.," NUREG/CR-0073, May 1978.

9. W. Marshall, P. D. Clemson, G. Walker, "Criticality Safety Criteria," ANS Trans, 35, 278 (1980) .

Thltl aocum.m containe Siem- NuclNI POW9t Corporation proonewy informlllion and i1 subject to the rw11ic:tian1 on the fim or title page.

•

•

•

CRmCALITY SAFETY ANALYSIS



DISTRIBUTION

Consumers Power (5)

J. W. Hulsman

C. D. Manning

T. C. Probasco

Document Control (5)

EMF-91-17 4(P) Issue Date: l 0104191

ATTACHMENT 4

Consumers Power Company Pa 1 i sades Pl ant Docket 50-255

TECHNICAL SPECIFICATION CHANGE REQUEST ENRICHMENT IN NEW AND SPENT FUEL STORAGE

SIEMENS NUCLEAR POWER PROPRIETARY REPORT EMF-91-142l(P)

October 28, 1991

Siemens Nuclear Power Corporation

EMF-91-1421 (P) Issue Date: 8130191


FOR THE PALISADES NEW FUEL STORAGE ARRAY

August 1991

NOTICE Tht. clocwMnt contain• inlctmalion propriM&ty to. Siemen1 Nuclear Power Co._.aon: ~ i1 aullmitt8d in confidence lllCI ie to be UMct solely for Ill• purl)OM lot wnicn rt i1 tum11nec1 &nd returned upon request This document &nd sucn infonnlzlloft is not ro be 191'roducad. !ransmitl8d, diacle>MCI, or uMd otll- in - et in 'l)alt witllout Ille -... IUllloriDllon of Siemens Nuclear P..- Corporation.

, ~- -- ·::1 1T:ro~;;-0To2 -9 rtn:.::::.o:-

0 PDR ADO~~ ~~01=1r1~ee p - r. - ·- - 1::F·.J'/' SNP-SR.4121.012. 1A (Nla!/91)

CUSTOMER DISCLAIMER.

IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT

PLEASE READ CAREFULLY

Siemens Nuclear Power Corporation's warranties and representations concerning the subject mat!Br of this document are those sat for1h in the Agreement between Siemens Nuclear Power Corporation and the Customer pursuant to which this document is issued. Accorcingty, except as otherwise expressly provided in such Agreement, neither Siemens Nuclear Power Corporation nor any person acting on its behalf mal<es any warranty or representation, expressed or implied, with respect to the accuracy, completaness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disdosed in this document will not infringe privately owned rights: or assumes any liabilities with respect to the use of any information, apparatus, method or process disclosed in this document.

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

In order to avoid impairment of rights of Siemens Nuclear Power Corporation in patents or inventions which may be included in the information contained in this document, the 1'8Cipient. by its aa:eptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Siemens Nuclear Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any pai.nts ar9 implied by the furnishing of this document

•

EMF-91-1421 (P)

Issue Date: 8/30/91

CRmCALITY SAFETY ANALYSIS


Approved By: ~e~ C. D. Manning, Criticality Safety Specialist Safety, Security, and Licensing

Approved By: ~ ~ T.C:PrObisco, Safety Supervisor Safety, Security, and Licensing

Approved By: 1'2"~~ W. E. Stavig, M nager Safety, Security, and Licensing

Date: 'J?) 'Jf'"J J,,,1 I 4

This document was prepared for Siemens Nuclear Power Corporation by Battelle Northwest under contract number W16662.

•

•

EMF-91-1421 (P)

Issue Date: 8/30/91



Prepared by: R. A. Jensen( Senior Research Engineer

•

Approved by: G)G) ~ . &'4 S. W. Heaberlin, Manager Reactor Systems Analysis Section

11/,c.J"'I ::i1. JC/I'//

Date

TABLE OF CONTENTS

EMF-91-1421 (P) .Page i

Section . Page No.

1.0 INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2.0 SUMMARY ..................... : .................................. 2

3.0 FUEL ASSEMBLY DESCRIPTION ........................................ 2

4.0 STORAGE ARRAY DESCRIPTION ....................................... 5

5.0 CALCULATIONAL METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

6.0 . METHODS VALIDATION .............................................. 8

7.0 MAJOR CONSERVATISMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 O

8.0 RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

9.0 ABNORMAL CONDITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

10.0 CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

11 .0 REFERENCES 0 0 0 0 0 0 0 0 0 0 I I I 0 I I 0 I I 0 0 I 0 I 0 O IO 0 I IO O 0 IO I 0 0 O O I I I 0 I I• I I I Io 16 .

I

This document contain• Siem- Nuclear Power Corpcl'Ulon pn>11n91a1y infom1alion and i• •ubject to the reslric!Jon• on th• Hm or !Ill• page.

•

•

LIST OF TABLES

EMF-91-1421 (P) Page ii

Table Page No.

Design Base Palisades Batch N Fuel Assembly Parameters . . . . . . . . . . . . . . . . . . . . 4

2 Critical Experiment Benchmark Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . g

3 Palisades New Fuel Storage Arra~ Reactivity (K9ff) .

Calculation Results for 4.25 w/o 35u Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 3

Thi'I docu'"*" eonlaint1 Slem9M Nucl- Power ~on prognewy infOf!Tl&llon and i• tMbject to Ille rftt'lction1 on Ille first or ti~• paiie.

•

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LIST OF FIGURES

Figure

EMF-91-1421 (P) Page iii

Page No.

1 Reload N1 Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

2 Palisades New Fuel Storage (Measurements of Installed Rack) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

3 Palisades New Fuel Storage Array (Worst Case Geometry) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2

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1.0 INTRODUCTION

EMF-91-1421 (P) Page 1

In July 1975 a criticality safety analysis(1) was performed for the Palisades New Fuel

Storage Array to allow storage (checkerboard arrangement) of fuel assemblies enriched to 3.2 .

wt.% 235u. This analysis included a parametric search to define the maximum array reactivity

(k9ff) as a function of uniform interspersed moderation within the array. The maximum k9ff

was shown to occur when the array is fully.flooded with water. s·ubsequent to this evaluation,

another criticality safety analysis(2) was performed in 1979, which demonstrated th~ storage

array to be adequately subcritical at a maximum average fuel assembly enrichment of 3.317

wt.% 235U.

It is the intent of this report to demonstrate that Palisades fuel (using Batch N fuel

assembly parameters) enriched to 4.20 wt.% 235u can be stored in the new fuel storage array

and continue to meet the reactivity limit criterion (k9ff s 0.95) stipulated for the storage of new

fuel by the Technical Specifications for the Palisades Nuclear Power Station.

This criticality analysis was performed in accordance with NUREG-0800 and

ANSl/ANS-57.3-1983 (including Section 6.2.4 and all applicable subsections).

Thhl document comaiM ~ Nucl.., Power Cotporalion propnewy 1nfonnuon and i11ubject to tl"I• -Clion1 on tr1• flrst or une pege.

2.0 SUMMARY

EMF-91-1421 (P) Page 2

The following criticality safety analysis of the new fuel storage array, containing stee!

box beams in alternate storage locations of the 3 x 24 array, demonstrates the array to be

adequately subcritical for Batch N type fuel having a maximum average fuel assembly

enrichment of 4.2 weight % 235u. The analysis demonstrates the storage array to have an

effective multiplication factor of <0.95 for assumed worst credible array conditions of fuel

assembly spacing and uniformly interspersed moderation.

The subject storage array meets the applicable criticality safety criteria (NUREG-0800

and ANSl-57.3-1983) subject to the following limitations:

1. Fuel Design: As specified in Section 3.0. Any significant deviation from this may

require additional analysis.

2. Array Design: As described in Section 4.0.

3. If fuel assemblies are stored with plastic wrapping, the bottom of wrapping shall be

open to assure drainage.

3.0 FUEL ASSEMBLY DESCRIPTION

The fuel assembly design (Batch N) assumed for the analysis is depicted in Figure

1 .1 As indicated, the 15 x 15 lattice arrangement includes a

Figure 1 actually depicts a Palisades Batch N1 assembly which is the most reactive fuel assembly design of batch N. All other assembly designs of this batch have lower average planar enrichments than the N1 design and include various loadings of godolinia bearing fuel pins.

~ WG I

~WG-+-I . LL! L L

L L M

L M M

L MJH

G M H

L M H

L M H·

LI GI LI L

MjMIMjM

HI HI HI H

HI HI H I H

HI HI HI H

HIH!HIH

HI HI HI H

L L . LI GI L L I L I LL

M M M/MIMIM!LiL

H H HIH!HIM!MIL

H H HjHJHjHIM L

H H HIHIH HIM G

H HIHIHIHJHIM L

H H HI HI H HiM L . '

L M H HI HI HI H I I H HI HI HJ HIM L

L M H HI HI HI H H H HjH\HJHjM!L

L M H H!HiHlH H H HI HI H Hi MI L . .

GjM H HjHIHIHIH H H I H i H HlMiG I .

L M H Hi Hi HI H H H HIHIH H ! MI L

L M M HIHiHJHJH H HIHIH MIM!L

L L M ·IM jM i M IM M M MjMjM M I L I L

EMF-91-1421 (P) • Page 3

~NG I

LL L L L I G i L IL L L L I GIL LIL )L.L

:~U.. 2.40 L • 2.68 w/o U-235 Fuel M • 3.1 O w/o U-235 Fuel H • 3.90 w/o U-235 Fuel G Guide Bar I Instrument Tube ~ Identifiers: WG • Wide lnterassembly Gae

NG :a Narrow I nterassembty Water Gap

FIGURE 1. Reload Nl Assembly( 3 l

~Q108067.1

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_;

•

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•

EMF-91-1421 (P) Page 5

Note that since the manufacturing enrichment uncertainty/tolerance is 0.05 wt% 235u

the analysis was performed using 4.25 w/o 235u fuel so that the rack could be qualified to

store 4.20 w/o 235u fuel. C4)

4.0 STORAGE 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 Company. (S) Subsequent information(?)

established that the dimensions indicated are actually measured center-to-center distances

between adjacent top bands. Such dimensions, therefore, represent nominal center-to-center

spacings between adjacent storage locations. It should 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 inches exists o"n the

design value. (Therefore, minimum nominal center-to-center separation of 9-3/8 inches was

assumed for this analysis.)

Concrete walls are adjacent to three sides of the storage array and are separated from

the fuel by 0.5 to 1.5 inches. For the purpose at this analysis, a 16 inch thick concrete

reflector was assumed to be touching three sides and bottom of the storage rack. The fourth

side and top of the array was assumed to be reflected by 6 inches of water, which is

effectively an infinitely thick water reflector .

EMF-91-1421 (P) Page 4

, single zirconium instrument guide tube located in the center of the assembly and eight

zirconium guide bars positioned on the exterior of the assembly. The remaining positions

. within the assembly are occupied by 216 U02 fuel rods.

The fuel assembly specifications and lattice cell parameters assum~d in this evaluation

are given in Table 1. The parameters such as pellet size, pellet density, clad thickness, etc.,

have been set to conservative values within the expected manufacturing tolerances.

TABLE 1. Design Base Palisades Batch N Fuel Assembly Parameters(5)

Parameter Nominal Model

Lattice Pitch, in. 0.550 0.550

Clad OD, in. 0.417 0.417

Clad Material · Zircaloy-4. · Zircaloy-4

Clad Thickness, in. 0.0295 0.0275

Pellet OD, in. 0.3510 0.3515

Pellet Density, % TD 94.5 96.0

Dish Volume, % . 1.4 0.9

Active Fuel Length, in. 131.800 132.026 ··-·

Fuel Rod Array . 15 x 15 15 x 15

Fuel Rods 216 216 .

Guide Bars 8 8

Instrument Tubes . 1

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•

EMF-91-1421 (P) Page 6

The 3 x 24 array contains 36 fuel assemblies with alternate positions occupied by 8 x 8

inch structural steel box beams having a nominal wall thickness of 5/16 inch. (2) A minimum

wall thickness of 0.25 inches was established for this analysis. (S)

Thie document comain• ~ NucJeat Power Cotporalion propnetary infotmallon and i• 1ubj«:t to Ill• rntncllon1 on Ill• flrst or till• pa;•.

l

6.0"

v' • 9.50" :CJi:CJi.:; '.11~5 oUILTl

.r = 9.<~· cuw· rn ~ .. ,:~u -113 • 3.35" 1113" DJ s:·:.:.Lu

•

SLi~SC:tiPf a !)f.:CiS. ::OHI~/~O:.:TH ?E~ CELL Ai aor1a:~ 1'1F ~~G: .

,/ B = 0.K. ::~Ill A:; S:!OULJ rE

-11a s = c:i.;;;- - ro.J s:·:;.Lu

FIGURE 2. Palisades New Fuel Storage Measurements of Installed Rack

• EMF-91-1421 (P)

Page 7

.)

•

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5.0 CALCULATIONAL METHODOLOGY

EMF-91-1421 (P) Page 8

The KENO-Va Monte carlo computer code(10) was used to mockup the Palisades New

Fuel Storage Array. The pin cell composed of fuel rod, gap, clad and moderator regions was

explicitly modelled. The Palisades N1 assembly with 8 guide bars and a single instrument

tube was also explicitly modelled with the exception of using a constant average planar

enrichment for the 216 fuel pins within the assembly (this is a conservative assumption, see

below).

In conjunction with KENO-Va, the SCALE 27 Group Library(11) was used. Both the 235u

and 238u cross sections were corrected for the effects of resonance self-shielding using the

NITAWL code. (12) NITAWL uses the Nordheim Integral Treatment to compute a neutron

spectrum which is used to obtain "effective" resonance cross sections. Dancoff correction

factors are used to account for shielding effects due to the close proximity of lattice pins.

To demonstrate that assigning fuel pins an average planar enrichment through out the

assembly, rather than using zoned enrichments, is conservative the 2-D transport code,

CASM0-2E(13) was used. CASMO calculations indicate that the actual "zoned" N1 assembly

design is approximately 9 mK less reactive than the corresponding constant enrichment

assembly.

6.0 METHODS VALIDATION

The calculational methodology used in this analysis was benchmarked against six

critical experiments.(14• 15) This set of criticals simulated conditions associated with light

EMF-91-1421 (P) Page 9

water fuel storage pools and were conducted at both the Battelle Critical Mass Laboratory

and the Babcock and Wilcox CX-1 O critical facility.

The benchmark results are given in Table 2. The corresponding bias (mean difference)

and benchmark uncertainty (difference standard deviation) of these benchmarks are 0.00755

and 0.00414, respectively.

TABLE 2. Critical Experiment Benchmark Results

Calculated Measured Bench No. K-Effective. C K-Effective. M

4 0.99552 +/- 0.00308 1.00000

7 . 0.99109 +/- 0.00306 1.00000

29 0.99870 +/- 0.00379 1.00000

2245 0.98983 +/- 0.00255 1.00000

2266 0.99244 +/- 0.00267 1.00000

2321 0.98710 +/- 0.00300 1.00000

Mean Difference = 0.00755 (Bias) Standard Deviation = 0.00414 (Benchmark Uncertainty)

A Kolmogorov-Smirnov test was conducted and indicated that the above difference distribution agrees well with a theoretical normal distribution.

Difference M-C

0.00448

0.00891

0.00130

0.01017

0.00756

0.01290

Thill docu- contain• Slemen1 Nuclear F'- Corpora!ion propn-.y informuon and i1 1ubject to tlle rellliction1 on ti!• fim or title page.

')

• 7.0 MAJOR CONSERVATISMS

The major conservatisms of this analysis were:

EMF-91-1421 (P) Page 1 o

• An average planar enrichment was used for all pin locations, rather than using "zoned"

enrichments.

• The most reactive Batch N fuel assembly design (i.e., highest enrichment sub batch

with no burnable poisons) was used in determining storage array reactivity (k-eff).

• Fuel design parameters such as pellet size, pellet density, clad thickness, etc., were

collectively set to conservative values within the expected manufacturing tolerances.

• The system was modeled as flooded with pure water (zero soluble poisons).

Thitl docu"*" con1ai"9 Slemen• Nuct-P-Cotiio..non prapnewy 1nformaaon and i9 auDjec! to Ille rntllc!ion1 on Ill• flm or li!le ga;e.

"

•

8.0 RESULTS

EMF-91-1421 (P) Page 11

In determining the maximum array reactivity (K9ff) for 4.25 w/o 235u fuel this analysis

evaluated the effects of varying moderator density and fuel assembly spacing within the rack.

All rack conditions were found to have an acceptable reactivity of K9ff <0.95 after accounting

for all uncertainties and bias. The maximum K-effective at the 95% confidence level was

0.945.2

The highest calculated reactivity occurs when the array is fully flooded with water

and the fuel assemblies are located in the rack as depicted in Figure 3. A summary of the

results is given in Table 3 .

2 95 upper limit K9ff = Cale. K9ff + Bias + Kg5 .fBenchmark Variance +·KENO Variance

= 0.92682 + 0.00755 + 2.015 ./0.004142 + 0.003482 = 0.945

Where Kg5 = 2.015 is based on a one-sided normal Students t Distribution at . the 0.05 significance level for 5 degrees of freedom.

Thi1 document contmino Siem- Nucleat p,,_, ~n progn-.y 1nforma11on Ind 11 aubject ID- Ille reotnct1on1 on Ill• Rrot 0t title Pata•·

";} .

. f ( .. I. 6°W111cr .

~ ~ --·--·-·----------------·-------·------·- -----------·--1

I I o~o~o~o~o~o~o~oao~o~o~LJ~' I . f 16" Concwle [f:J [=] [~J (] (~] [] [{~] [J IJ~d [] [~d [] f~!I [] bJ] [~] b~ D bU [] fJJ l~J QJ [] . 16• Concrule

f ! D~D~D~D~D~D~D~D~D~D~D~D~j ~ . 0 ·--------·--·----·--------·-·-·------·--·-·-·------------..-----------------!: ii) . 10· Con.::rr;le

J ~ . t ~ t ~ !l ~ 0. ii' 0 ~

.... CJ) a ....

l 0 0 O> g CJ)

• (I>

I '° (I> 0 g 3 ..

0 (I> :> g -• -< "' -i !! g • ! ~

•

Bollom 10· Concrolti fop 6" Walur Ccrolt:r lo-Genier Spucing

ol c .. 11~. 9.37~·

•

10 fuol Assembly Locullun ·

1--1 S1ool Box Loc111io11

(IJ" x a· 0.0., .25" wall

1l1ic.knu~~. c .. 11101ud)

.• --·,I • r

•

EMF-91-1421 (P) Page 13

TABLE 3. Palisades New Fuel Storage Array Reactivity (K9ff)· Calculation.

Results for 4.25 w/o 235u Fuel)

Array Rack Moderator Calculated

Geometry Void Fraction K-effective

Nominal 0.0 0.91399 +/- 0.00343

Adverse 0.0 0.92682 +/- 0.00348

Adverse 0.05 0.91123 +/- 0.00310

Adverse 0.10 0.89581 +/- 0.00309

Adverse 0.30 0.85648 +/- 0.00288

Adverse 0.50 0.81299 +/- 0.00329

Adverse 0.70 0.80100 +/- 0.00323

Adverse 0.75 0.79879 +/- 0.00305

Adverse 0.80 0.79041 +/- 0.00303

Adverse 0.90 0.71333 +/- 0~00283

Adverse 0.95 0.63454 +/- 0.00275

• Adverse Rack Geometry is depicted in Figure 3. Three sides and bottom reflected by 16 in. of concrete; open side and top reflected by 6 in. of water at full density.

-

• Fuel/moderator temperature was 20°c.

• Other than the steel box beams no rack structual components were modelled.

Thill document COlllaiM si.m- Nuclat Power Cot!loral!on propn.wy infotmaJion and ii 1ubjec! IO !lie rnlric:1Jon1 on !II• first or ttU• page.

·J

• r

9.0 ABNORMAL CONDITIONS

EMF-91-1421 (P) Page 14

In accordance with ANSl/ANS-57.3-1983 (6.2.4.1) this analysis evaluated the potential

adverse effects of fuel handling accidents. Due to the fact that the Palisades New Fuel Array

is a dr}' storage rack fuel handling accidents do not adversely impact nuclear safety. The

infinite multiplication factor (KJ for dry 5 w/o 235u uranium oxide systems is less than 0.8. (16)

Specific design features(9) of the rack which preclude flooding include:

1) All cells and spaces between cells have openings at the bottom to facilitate draining.

2) The rack is situated 3 ft above a course steel grading floor. The floor below the

• grading is approximately another 12 ft .

• Thie doc:ullleftt contmiM Slementl NuclMt Power Co~on propn.wy inletmUion and i1 1uDjec1 to me ,_ct1on1 on me flm or !ltl• page."

·J

•

10.0 CONCLUSIONS

EMF-91-1421 (P) Page 15

This a!'lalysis conservatively demonstrates the reactivity of the Palisades new fuel

storage array as described for a Batch N fuel. assembly with an enrichment of 4.20 w/o 235u

to be less than 0.95 under the worst credible storage array conditions. 3

r

3 Although this analysis was performed using an average planar enrichment of 4.25 wt% 235u it is necessary to lower the allowable enrichment by the mam.~facturing enrichment uncertainty/tolerance which is 0.05 wt% 235u. (4)

< .~ l •

•

•

11.0 REFERENCES

EMF-91-1421 (P) Page 16

(1) L. E. Hansen, "Palisades New Fuel Storage Array Criticality Safety Analysis", XN-NF-309,

Exxon Nuclear Co., (July 1975).

(2) C. 0. Brown, Palisades New Fuel Storage Array Criticality Safety Reanalysis, XN-NF-

508, Exxon Nuclear Co., (February 1979).

(3) Advanced Nuclear Fuels Drawing, "Fuel Assembly Load Map Type N1 ," ANF-307, 309,

Revision O.

(4) "Palisades Characteristics of Reload N," ANF-CS-386, Revision 1.

(5) "Principal Reload Fuel Design Parameters Palisades Reload M," ANF-89-063(~) .

(6) Letter, W. J. Beckius (CPCO) to W. E. Niemuth (ENC), May 13, 1975.

(7) W. J. Beckius, Personal.Communication, Consumers Power Company, June 10, 1975.

(8) B. Webb, Personal. Communication, Consumers Power Company, June 1 O, 1975.

(9) T. Hollowell, Personal Communication, Consumers Power Company,

August 26, 1991 .

(1 O) L. M. Petrie and N. F. Landers, "KENO V.A An Improved Monte Carlo Criticality Program

with Supergrouping." NUREG/CA-0200, December 1984.

(11) "SCALE: A Modular Code System for Performing Standardized Computer Analyses for

Ucensing Evaluation," NUREG/CR-0200.

(12) R. M. Westfall, "NITAWL-S," ORNL./NUREG/CR-0200, October 1981.

(13) "CASM0-2E: A Fuel Assembly Burnup Program (Methodology)," Studsvik/NFA-86/8,

Studsvik Energitekmik AB, Nykoping, Sweden .

Thie clocl.lment contailltl SJ_..,, Nucleat Pow.r Corporanon propn-.y 1ntorme11on and i1 1ubject to Ill• rftlriclion1 on Ill• firwt or tine pa<;ie.

EMF-91-1421 (P) Page 17

(14) M. N. Baldwin, et at., "Critical Experiments Supporting Close Proximity Water Storage of

Power Reactor Fuel, "BAW-1484-7, July 1979.

(15) S. R. Bierman, B. M. Durst, and E. D. Clayton, "Critical Separation between Subcritical

Clusters of 4.29% Enriched U02 Rods in Water with Fixed Neutron Poisons," PNL-2615,

' March 1 978.

(16) S. R. Bierman and E. D. Clayton, "Gmelin Handbook of Inorganic Chemistry,"

Supplement Volume A6, 8th Edition, 1983.

Thi1 document contain1 Slem9"1 Nuclear Power Corporation propnef41y •nfcrma11cn Md io subject to me rntnc:11cn1 en me first er UH• page.

,

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I " • . "


EMF-91-1421 (P) Issue Date: 8/30/91


DISTRIBUTION

Consumers Power (5)

J. W. Hulsman

C. D. Manning

T. C. Probasco

Document Control (5)

Thill document comaiM si-- Nucl.., Power Coriiorauon grocnewy 1nfonna11on and 11 1ubject ta !II• /'fttriction1 on !lie flm or tit!• g&Q•.

. ., f ..• ,· ...

•

consumers Power ·

POW ERIN& NllCHl&AN'S PRD&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043

October 28, 1991

Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

DOCKET 50-255 - LICENSE DPR-20 - PALISADES PLANT'-

GB Slade General Manager

TECHNICAL SPECIFICATIONS CHANGE REQUEST - FUEL ASSEMBLY ENRICHMENT ALLOWED IN NEW FUEL AND SPENT FUEL RACKS

Enclosed is a proposed change to the Palisades Technical Specifications which will allow the storage of fuel assemblies enriched to 4.20 weight percent (w/o) U235 in the new fuel racks and enriched to 4.40 w/o U235 in the Region I (NUS) racks in,the spent fuel pool. This change is the result of changes to the enrichment of the Palisades fuel in order to design a low leakage core to minimize the effect of fluence on the reactor vessel and to maintain a reasonable length fuel cycle. Plant operation with the higher U23? enrichment will be the subject of a separate change request to be submitted 1n the near future.

Storage of .fuel assemblies enriched as described above in the new fuel racks and spent fuel racks has been analyzed and evaluated by Siemens Nuclear Power Corporation, the Palisades nuclear fuel supplier. Siemens Nuclear Power reports, "Criticality Safety Analysis for the Palisades New Fuel Storage Array" (EMF-91-142l(P)) and "Criticality Safety Analysis for the Palisades Spent Fuel Storage Pool NUS Racks" (EMF-91-174(P)), find that the fuel 'tored as proposed by this change request meets the reactivity criteria of Sections 9.1.l and 9.1.2 of NUREG 0800, ANSI/ANS 57.2-1983 and 57.3-1983, and the Palisades Technical Specifications (Keff ~ .95).

Attachment 1 contains proposed changed Technical Specifications pages. Attachment 2 contains marked up Technical Specifications pages. Attachment 3 is Siemens report EMF-91-174(P) and Attachment 4 is Siemens report EMF-91-142l(P).

Siemens Nuclear Power Corporation considers the information contained in EMF-91-174(P) and EMF-91-142l(P) to be proprietary. In accordance with the requirements of IO CFR 2.790(b), two affidavits are enclosed to support the withholding of this proprietary information from public disclosure.

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